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

Review of the magnetic fusion program by the 1986 ERAB Fusion Panel

NASA Astrophysics Data System (ADS)

Davidson, Ronald C.

1987-09-01

The 1986 ERAB Fusion Panel finds that fusion energy continues to be an attractive energy source with great potential for the future, and that the magnetic fusion program continues to make substantial technical progress. In addition, fusion research advances plasma physics, a sophisticated and useful branch of applied science, as well as technologies important to industry and defense. These factors fully justify the substantial expenditures by the Department of Energy in fusion research and development (R&D). The Panel endorses the overall program direction, strategy, and plans, and recognizes the importance and timeliness of proceeding with a burning plasma experiment, such as the proposed Compact Ignition Tokamak (CIT) experiment.

Final Report Advanced Quasioptical Launcher System

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

Jeffrey Neilson

2010-04-30

This program developed an analytical design tool for designing antenna and mirror systems to convert whispering gallery RF modes to Gaussian or HE11 modes. Whispering gallery modes are generated by gyrotrons used for electron cyclotron heating of fusion plasmas in tokamaks. These modes cannot be easily transmitted and must be converted to free space or waveguide modes compatible with transmission line systems.This program improved the capability of SURF3D/LOT, which was initially developed in a previous SBIR program. This suite of codes revolutionized quasi-optical launcher design, and this code, or equivalent codes, are now used worldwide. This program added functionality tomore » SURF3D/LOT to allow creating of more compact launcher and mirror systems and provide direct coupling to corrugated waveguide within the vacuum envelope of the gyrotron. Analysis was also extended to include full-wave analysis of mirror transmission line systems. The code includes a graphical user interface and is available for advanced design of launcher systems.« less

What`s fair is fair

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

Nachtrieb, R.; Freidberg, J.P.

The newly elucidated strategy for the magnetic fusion program set forth by the Department of Energy calls for increased emphasis on alternate concepts. This strategy is motivated by the recognition that in spite of its many attractive features, a tokamak tends to be a low power density device, ultimately translating into large and corresponding expensive reactor. ITER, as it is currently envisaged, is a good example of a large, expensive, plain vanilla tokamak. In its defense, ITER rightly claims that its base design is very conservative in order to minimize the risk of failure. In order to increase power densitymore » and reduce cost there are two qualitatively different approaches that one can follow: discover advanced modes of tokamak operation or develop near alternate concepts. To decide which path to follow is a difficult task because of the uncertainties involved in making accurate comparisons between different concepts at different stages of development. One area, however, that most would agree is meaningful is ideal MHD stability. For any given concept to be credible as a reactor, it must at least be stable against macroscopic ideal MHD modes. The TPX design, for instance, goes to considerable trouble to obtain stability against external kinks: a close fitting metallic cage, rotation to stabilize the resistive wall version of the external kink, and, if all else fails, feedback. For credibility any other advanced tokamak or alternate concept should be held to the same standards of ideal MHD stability. As a first step in addressing this requirement we have investigated the stability of the RFP since it can be simply and accurately modeled as a straight cylinder. The RFP is well known to have good stability at high P against internal modes but is very unstable to external modes. We have developed a linear stability code which treats the plasma as an ideal compressible fluid, and includes longitudinal flow and a resistive wall.« less

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

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

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.

Halo current diagnostic system of experimental advanced superconducting tokamak

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

Chen, D. L.; Shen, B.; Sun, Y.

2015-10-15

The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

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.

Overview of MST Research

NASA Astrophysics Data System (ADS)

Chapman, B. E.

2017-10-01

MST progress in advancing the RFP for (1) fusion plasma confinement with ohmic heating and minimal external magnetization, (2) predictive capability in toroidal confinement physics, and (3) basic plasma physics is summarized. Validation of key plasma models is a program priority, which is enhanced by programmable power supplies (PPS) to maximize inductive capability. The existing PPS enables access to very low plasma current, down to Ip =0.02 MA. This greatly expands the Lundquist number range S =104 -108 and allows nonlinear, 3D MHD computation using NIMROD and DEBS with dimensionless parameters that overlap those of MST plasmas. A new, second PPS will allow simultaneous PPS control of the Bp and Bt circuits. The PPS also enables MST tokamak operation, thus far focused on disruptions and RMP suppression of runaway electrons. Gyrokinetic modeling with GENE predicts unstable TEM in improved-confinement RFP plasmas. Measured fluctuations have TEM properties including a density-gradient threshold larger than for tokamak plasmas. Turbulent energization of an electron tail occurs during sawtooth reconnection. Probe measurements hint that drift waves are also excited via the turbulent cascade in standard RFP plasmas. Exploration of basic plasma science frontiers in MST RFP and tokamak plasmas is proposed as part of WiPPL, a basic science user facility. Work supported by USDoE.

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

Kim, Dong-Hwan; Hong, Suk-Ho; National Fusion Research Institute

Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliabilitymore » of the method.« less

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

Advancing the understanding of plasma transport in mid-size stellarators

NASA Astrophysics Data System (ADS)

Hidalgo, Carlos; Talmadge, Joseph; Ramisch, Mirko; TJ-II, the; HXS; TJ-K Teams

2017-01-01

The tokamak and the stellarator are the two main candidate concepts for magnetically confining fusion plasmas. The flexibility of the mid-size stellarator devices together with their unique diagnostic capabilities make them ideally suited to study the relation between magnetic topology, electric fields and transport. This paper addresses advances in the understanding of plasma transport in mid-size stellarators with an emphasis on the physics of flows, transport control, impurity and particle transport and fast particles. The results described here emphasize an improved physics understanding of phenomena in stellarators that complements the empirical approach. Experiments in mid-size stellarators support the development of advanced plasma scenarios in Wendelstein 7-X (W7-X) and, in concert with better physics understanding in tokamaks, may ultimately lead to an advance in the prediction of burning plasma behaviour.

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.

Neutron field measurement at the Experimental Advanced Superconducting Tokamak using a Bonner sphere spectrometer

NASA Astrophysics Data System (ADS)

Hu, Zhimeng; Zhong, Guoqiang; Ge, Lijian; Du, Tengfei; Peng, Xingyu; Chen, Zhongjing; Xie, Xufei; Yuan, Xi; Zhang, Yimo; Sun, Jiaqi; Fan, Tieshuan; Zhou, Ruijie; Xiao, Min; Li, Kai; Hu, Liqun; Chen, Jun; Zhang, Hui; Gorini, Giuseppe; Nocente, Massimo; Tardocchi, Marco; Li, Xiangqing; Chen, Jinxiang; Zhang, Guohui

2018-07-01

The neutron field measurement was performed in the Experimental Advanced Superconducting Tokamak (EAST) experimental hall using a Bonner sphere spectrometer (BSS) based on a 3He thermal neutron counter. The measured spectra and the corresponding integrated neutron fluence and dose values deduced from the spectra at two exposed positions were compared to the calculated results obtained by a general Monte Carlo code MCNP5, and good agreements were found. The applicability of a homemade dose survey meter installed at EAST was also verified with the comparison of the ambient dose equivalent H*(10) values measured by the meter and BSS.

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.

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

Overview of the TCV tokamak program: scientific progress and facility upgrades

NASA Astrophysics Data System (ADS)

Coda, S.; Ahn, J.; Albanese, R.; Alberti, S.; Alessi, E.; Allan, S.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Ariola, M.; Bernert, M.; Beurskens, M.; Bin, W.; Blanchard, P.; Blanken, T. C.; Boedo, J. A.; Bolzonella, T.; Bouquey, F.; Braunmüller, F. H.; Bufferand, H.; Buratti, P.; Calabró, G.; Camenen, Y.; Carnevale, D.; Carpanese, F.; Causa, F.; Cesario, R.; Chapman, I. T.; Chellai, O.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Costea, S.; Crisanti, F.; Cruz, N.; Czarnecka, A.; Decker, J.; De Masi, G.; De Tommasi, G.; Douai, D.; Dunne, M.; Duval, B. P.; Eich, T.; Elmore, S.; Esposito, B.; Faitsch, M.; Fasoli, A.; Fedorczak, N.; Felici, F.; Février, O.; Ficker, O.; Fietz, S.; Fontana, M.; Frassinetti, L.; Furno, I.; Galeani, S.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garrido, I.; Geiger, B.; Giovannozzi, E.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Graves, J. P.; Guirlet, R.; Hakola, A.; Ham, C.; Harrison, J.; Hawke, J.; Hennequin, P.; Hnat, B.; Hogeweij, D.; Hogge, J.-Ph.; Honoré, C.; Hopf, C.; Horáček, J.; Huang, Z.; Igochine, V.; Innocente, P.; Ionita Schrittwieser, C.; Isliker, H.; Jacquier, R.; Jardin, A.; Kamleitner, J.; Karpushov, A.; Keeling, D. L.; Kirneva, N.; Kong, M.; Koubiti, M.; Kovacic, J.; Krämer-Flecken, A.; Krawczyk, N.; Kudlacek, O.; Labit, B.; Lazzaro, E.; Le, H. B.; Lipschultz, B.; Llobet, X.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Maget, P.; Maljaars, E.; Malygin, A.; Maraschek, M.; Marini, C.; Martin, P.; Martin, Y.; Mastrostefano, S.; Maurizio, R.; Mavridis, M.; Mazon, D.; McAdams, R.; McDermott, R.; Merle, A.; Meyer, H.; Militello, F.; Miron, I. G.; Molina Cabrera, P. A.; Moret, J.-M.; Moro, A.; Moulton, D.; Naulin, V.; Nespoli, F.; Nielsen, A. H.; Nocente, M.; Nouailletas, R.; Nowak, S.; Odstrčil, T.; Papp, G.; Papřok, R.; Pau, A.; Pautasso, G.; Pericoli Ridolfini, V.; Piovesan, P.; Piron, C.; Pisokas, T.; Porte, L.; Preynas, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Reich, M.; Reimerdes, H.; Reux, C.; Ricci, P.; Rittich, D.; Riva, F.; Robinson, T.; Saarelma, S.; Saint-Laurent, F.; Sauter, O.; Scannell, R.; Schlatter, Ch.; Schneider, B.; Schneider, P.; Schrittwieser, R.; Sciortino, F.; Sertoli, M.; Sheikh, U.; Sieglin, B.; Silva, M.; Sinha, J.; Sozzi, C.; Spolaore, M.; Stange, T.; Stoltzfus-Dueck, T.; Tamain, P.; Teplukhina, A.; Testa, D.; Theiler, C.; Thornton, A.; Tophøj, L.; Tran, M. Q.; Tsironis, C.; Tsui, C.; Uccello, A.; Vartanian, S.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vijvers, W. A. J.; Vlahos, L.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weisen, H.; Wischmeier, M.; Zestanakis, P.; Zuin, M.; the EUROfusion MST1 Team

2017-10-01

The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

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

Shaing, K. C.

It is shown that potato orbits in the near-axis region of a high beta tokamak are squeezed in a magnetic well. The squeezing factor is the same as that for the banana orbits derived in an earlier work [Phys. Plasmas 3, 2843 (1996)]. It depends on the energy of the particle. For high-energy particles, the size of the squeezed orbits is independent of their energy. This implies improved confinement for high-energy particles and for high beta tokamaks with advanced fuels.

Long-distance delivery of multi-channel polarization signals in nuclear fusion research

NASA Astrophysics Data System (ADS)

Ko, Jinseok; Chung, Jinil; Lee, Kyuhang

2017-04-01

A polarization-preserving optical system that includes a dual photoelastic modulator (PEM) has been designed and fabricated for the motional Stark effect (MSE) diagnostic system which measures internal magnetic field structures inside the tokamak for the Korea Superconducting Tokamak Advanced Research. The collection optics located outside the vacuum window is composed of four lenses, a dielectric coated mirror, and a dichroic beam splitter in addition to the PEM and a polarizer. The fiber dissector is designed based on the focal plane that aligns 25 lines of sight, each of which constitutes a bundle of 19 600-μm fibers. The fibers run about 40 m from the front optics in the tokamak vacuum vessel to the detector in the diagnostic area remote from the tokamak hall. This takes the advantage of the fact that the polarization information is intensity-modulated once going through the PEM and the polarizer. The polarization signals measured by the MSE diagnostic successfully demonstrates its proof-of-principle physics that is critical in the stable and steady-state operation of the tokamak plasmas.

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

Kuprat, A.P.; Glasser, A.H.

The authors discuss unstructured grids for application to transport in the tokamak edge SOL. They have developed a new metric with which to judge element elongation and resolution requirements. Using this method, the authors apply a standard moving finite element technique to advance the SOL equations while inserting/deleting dynamically nodes that violate an elongation criterion. In a tokamak plasma, this method achieves a more uniform accuracy, and results in highly stretched triangular finite elements, except near separatrix X-point where transport is more isotropic.

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.

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.

Energetic particle-driven compressional Alfvén eigenmodes and prospects for ion cyclotron emission studies in fusion plasmas

DOE PAGES

Gorelenkov, N. N.

2016-10-01

As a fundamental plasma oscillation the compressional Alfvén waves (CAW) are interesting for plasma scientists both academically and in applications for fusion plasmas. They are believed to be responsible for the ion cyclotron emission (ICE) observed in many tokamaks. The theory of CAW and ICE was significantly advanced at the end of 20th century in particular motivated by first DT experiments on TFTR and subsequent JET DT experimental studies. More recently, ICE theory was advanced by ST (or spherical torus) experiments with the detailed theoretical and experimental studies of the properties of each instability signal. There the instability responsible formore » ICE signals previously indistinguishable in high aspect ratio tokamaks became the subjects of experimental studies. We discuss further the prospects of ICE theory and its applications for future burning plasma (BP) experiments such as the ITER tokamak-reactor prototype being build in France where neutrons and gamma rays escaping the plasma create extremely challenging conditions for fusion alpha particle diagnostics.a« 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

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.

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.

Small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

2017-04-01

SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

Data Acquisition Systems

NASA Technical Reports Server (NTRS)

1994-01-01

In the mid-1980s, Kinetic Systems and Langley Research Center determined that high speed CAMAC (Computer Automated Measurement and Control) data acquisition systems could significantly improve Langley's ARTS (Advanced Real Time Simulation) system. The ARTS system supports flight simulation R&D, and the CAMAC equipment allowed 32 high performance simulators to be controlled by centrally located host computers. This technology broadened Kinetic Systems' capabilities and led to several commercial applications. One of them is General Atomics' fusion research program. Kinetic Systems equipment allows tokamak data to be acquired four to 15 times more rapidly. Ford Motor company uses the same technology to control and monitor transmission testing facilities.

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.

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.

Fusion plasma theory project summaries

NASA Astrophysics Data System (ADS)

1993-10-01

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

Observation of runaway electron beams by visible color camera in the Experimental Advanced Superconducting Tokamak

NASA Astrophysics Data System (ADS)

Shi, Yuejiang; Fu, Jia; Li, Jiahong; Yang, Yu; Wang, Fudi; Li, Yingying; Zhang, Wei; Wan, Baonian; Chen, Zhongyong

2010-03-01

The synchrotron radiation originated from the energetic runaway electrons has been measured by a visible complementary metal oxide semiconductor camera working in the wavelength ranges of 380-750 nm in the Experimental Advanced Superconducting Tokamak [H. Q. Liu et al., Plasma Phys. Contr. Fusion 49, 995 (2007)]. With a tangential viewing into the plasma in the direction of electron approach on the equatorial plane, the synchrotron radiation from the energetic runaway electrons was measured in full poloidal cross section. The synchrotron radiation diagnostics provides a direct pattern of the runaway beam inside the plasma. The energy and pitch angle of runaway electrons have been obtained according to the synchrotron radiation pattern. A stable shell shape of synchrotron radiation has been observed in a few runaway discharges.

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

Advanced X-ray Imaging Crystal Spectrometer for Magnetic Fusion Tokamak Devices

NASA Astrophysics Data System (ADS)

Lee, S. G.; Bak, J. G.; Bog, M. G.; Nam, U. W.; Moon, M. K.; Cheon, J. K.

2008-03-01

An advanced X-ray imaging crystal spectrometer is currently under development using a segmented position sensitive detector and time-to-digital converter (TDC) based delay-line readout electronics for burning plasma diagnostics. The proposed advanced XICS utilizes an eight-segmented position sensitive multi-wire proportional counter and supporting electronics to increase the spectrometer performance includes the photon count-rate capability and spatial resolution.

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

A. Brooks; A.H. Reiman; G.H. Neilson

High-beta, low-aspect-ratio (compact) stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady-state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks (2-4). They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A beta = 4% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment (NCSX). It has amore » substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at beta = 4% have been evaluated for the Quasi-Omnigeneous Stellarator (QOS) experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described.« less

Overview of MST Research

NASA Astrophysics Data System (ADS)

Sarff, J. S.

2016-10-01

MST progress in advancing the RFP for (1) fusion plasma confinement with ohmic heating and minimal external magnetization, (2) predictive capability in toroidal confinement physics, and (3) basic plasma physics is summarized. Validation of key plasma models is a program priority. Programmable power supplies (PPS) are being developed to maximize inductive capability. Well-controlled flattops with current as low as 0.02 MA are produced with an existing PPS, and Ip <= 0.8 MA is anticipated with a second PPS under construction. The Lundquist number spans S =10(4 - 9) for 0.02-0.8 MA, allowing nonlinear MHD validation using NIMROD and DEBS at low S to be connected to highest S experiments. The PPS also enables MST tokamak operation for studying transients and runaway electron suppression with RMPs. Gyrokinetic modeling with GENE predicts unstable TEM in improved-confinement plasmas. Fluctuations are measured with TEM properties including a density-gradient threshold larger than for tokamak plasmas. Probe measurements hint that drift waves are also excited via the turbulent cascade in standard RFP plasmas. Turbulent energization of an electron tail occurs during sawtooth reconnection. New diagnostics are being developed to measure the energetic ion profile and transport from EP instabilities with NBI. Supported by US DoE and NSF.

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 (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

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

Recent Progress on Spherical Torus Research

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

Ono, Masayuki; Kaita, Robert

2014-01-01

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ~ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ~ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configurationmore » can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of attractive fusion energy power source. Since the start of the two megaampere class ST facilities in 2000, National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all of fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.« less

Apollo - An advanced fuel fusion power reactor for the 21st century

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

Kulcinski, G.L.; Emmert, G.A.; Blanchard, J.P.

1989-03-01

A preconceptual design of a tokamak reactor fueled by a D-He-3 plasma is presented. A low aspect ratio (A=2-4) device is studied here but high aspect ratio devices (A > 6) may also be quite attractive. The Apollo D-He-3 tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The overall efficiency ranges from 37 to 52% depending on whether the bremsstrahlung energy is utilized. The low neutron wall loading (0.1 MW/m/sup 2/) allows a permanent first wall to be designed and the low nuclear decay heat enablesmore » the reactor to be classed as inherently safe. The cost of electricity from Apollo is > 40% lower than electricity from a similar sized DT reactor.« less

Establishment and Assessment of Plasma Disruption and Warning Databases from EAST

NASA Astrophysics Data System (ADS)

Wang, Bo; Robert, Granetz; Xiao, Bingjia; Li, Jiangang; Yang, Fei; Li, Junjun; Chen, Dalong

2016-12-01

Disruption database and disruption warning database of the EAST tokamak had been established by a disruption research group. The disruption database, based on Structured Query Language (SQL), comprises 41 disruption parameters, which include current quench characteristics, EFIT equilibrium characteristics, kinetic parameters, halo currents, and vertical motion. Presently most disruption databases are based on plasma experiments of non-superconducting tokamak devices. The purposes of the EAST database are to find disruption characteristics and disruption statistics to the fully superconducting tokamak EAST, to elucidate the physics underlying tokamak disruptions, to explore the influence of disruption on superconducting magnets and to extrapolate toward future burning plasma devices. In order to quantitatively assess the usefulness of various plasma parameters for predicting disruptions, a similar SQL database to Alcator C-Mod for EAST has been created by compiling values for a number of proposed disruption-relevant parameters sampled from all plasma discharges in the 2015 campaign. The detailed statistic results and analysis of two databases on the EAST tokamak are presented. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2014GB103000)

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.

Analysis of higher harmonics on bidirectional heat pulse propagation experiment in helical and tokamak plasmas

NASA Astrophysics Data System (ADS)

Kobayashi, T.; Ida, K.; Inagaki, S.; Tsuchiya, H.; Tamura, N.; Choe, G. H.; Yun, G. S.; Park, H. K.; Ko, W. H.; Evans, T. E.; Austin, M. E.; Shafer, M. W.; Ono, M.; López-bruna, D.; Ochando, M. A.; Estrada, T.; Hidalgo, C.; Moon, C.; Igami, H.; Yoshimura, Y.; Tsujimura, T. Ii.; Itoh, S.-I.; Itoh, K.

2017-07-01

In this contribution we analyze modulation electron cyclotron resonance heating (MECH) experiment and discuss higher harmonic frequency dependence of transport coefficients. We use the bidirectional heat pulse propagation method, in which both inward propagating heat pulse and outward propagating heat pulse are analyzed at a radial range, in order to distinguish frequency dependence of transport coefficients due to hysteresis from that due to other reasons, such as radially dependent transport coefficients, a finite damping term, or boundary effects. The method is applied to MECH experiments performed in various helical and tokamak devices, i.e. Large Helical Device (LHD), TJ-II, Korea Superconducting Tokamak Advanced Research (KSTAR), and Doublet III-D (DIII-D) with different plasma conditions. The frequency dependence of transport coefficients are clearly observed, showing a possibility of existence of transport hysteresis in flux-gradient relation.

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.

Alternate fusion fuels workshop

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

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

The study of heat flux for disruption on experimental advanced superconducting tokamak

NASA Astrophysics Data System (ADS)

Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen

2016-05-01

Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.

MHD Instabilities and Toroidal Field Effects on Plasma Column Behavior in Tokamak

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

Khorshid, Pejman; Plasma Physics Research Center, Islamic Azad University, 14665-678, Tehran; Wang, L.

2006-12-04

In the edge plasma of the CT-6B and IRAN-T1 tokamaks the shape of plasma column based on MHD behavior has been studied. The bulk of plasma behavior during plasma column rotation as non-rigid body plasma has been investigated. We found that mode number and rotation frequency of plasma column are different in angle position, so that the mode number detected from Mirnov coils array located in poloidal angle on the inner side of chamber is more than outer side which it can be because of toroidal magnetic field effects. The results of IR-T1 and CT-6B tokamaks compared with each other,more » so that in the CT-6B because of its coils number must be less, but because of its Iron core the effect of toroidal magnetic field became more effective with respect to IR-T1. In addition, it is shown that the plasma column behaves as non-Rigid body plasma so that the poloidal rotation velocity variation in CT-6B is more than IR-T1. A relative correction for island rotation frequency has been suggested in connection with IRAN-T1 and CT-6B tokamak results, which can be considered for optical measurement purposes and also for future advanced tokamak control design.« less

Filterscope diagnostic system on the Experimental Advanced Superconducting Tokamak (EAST)

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

Xu, Z.; Wu, C. R.; Yao, X. J.

2016-11-15

A filterscope diagnostic system has been mounted to observe the line emission and visible bremsstrahlung emission from plasma on the experimental advanced superconducting tokamak during the 2014 campaign. By this diagnostic system, multiple wavelengths including D{sub α} (656.1 nm), D{sub γ} (433.9 nm), He II (468.5 nm), Li I (670.8 nm), Li II (548.3 nm), C III (465.0 nm), O II (441.5 nm), Mo I (386.4 nm), W I (400.9 nm), and visible bremsstrahlung radiation (538.0 nm) are monitored with corresponding wavelength filters. All these multi-channel signals are digitized at up to 200 kHz simultaneously. This diagnostic plays a crucialmore » role in studying edge localized modes and H-mode plasmas, due to the high temporal resolution and spatial resolution that have been designed into it.« less

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.

Some Aspects of Advanced Tokamak Modeling in DIII-D

NASA Astrophysics Data System (ADS)

St John, H. E.; Petty, C. C.; Murakami, M.; Kinsey, J. E.

2000-10-01

We extend previous work(M. Murakami, et al., General Atomics Report GA-A23310 (1999).) done on time dependent DIII-D advanced tokamak simulations by introducing theoretical confinement models rather than relying on power balance derived transport coefficients. We explore using NBCD and off axis ECCD together with a self-consistent aligned bootstrap current, driven by the internal transport barrier dynamics generated with the GLF23 confinement model, to shape the hollow current profile and to maintain MHD stable conditions. Our theoretical modeling approach uses measured DIII-D initial conditions to start off the simulations in a smooth consistent manner. This mitigates the troublesome long lived perturbations in the ohmic current profile that is normally caused by inconsistent initial data. To achieve this goal our simulation uses a sequence of time dependent eqdsks generated autonomously by the EFIT MHD equilibrium code in analyzing experimental data to supply the history for the simulation.

Fast valve based on double-layer eddy-current repulsion for disruption mitigation in Experimental Advanced Superconducting Tokamak.

PubMed

Zhuang, H D; Zhang, X D

2015-05-01

A fast valve based on the double-layer eddy-current repulsion mechanism has been developed on Experimental Advanced Superconducting Tokamak (EAST). In addition to a double-layer eddy-current coil, a preload system was added to improve the security of the valve, whereby the valve opens more quickly and the open-valve time becomes shorter, making it much safer than before. In this contribution, testing platforms, open-valve characteristics, and throughput of the fast valve are discussed. Tests revealed that by choosing appropriate parameters the valve opened within 0.15 ms, and open-valve times were no longer than 2 ms. By adjusting working parameter values, the maximum number of particles injected during this open-valve time was estimated at 7 × 10(22). The fast valve will become a useful tool to further explore disruption mitigation experiments on EAST in 2015.

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

Stencel, J.R.; Finley, V.L.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory for CY90. The report is prepared to provide the US Department of Energy (DOE) and the public with information on the level of radioactive and nonradioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs. The objective of the Annual Site Environmental Report is to document evidence that DOE facility environmental protection programs adequately protect the environment and the public health. The PPPL has engaged in fusion energy research sincemore » 1951 and in 1990 had one of its two large tokamak devices in operation: namely, the Tokamak Fusion Test Reactor. The Princeton Beta Experiment-Modification is undergoing new modifications and upgrades for future operation. A new machine, the Burning Plasma Experiment -- formerly called the Compact Ignition Tokamak -- is under conceptual design, and it is awaiting the approval of its draft Environmental Assessment report by DOE Headquarters. This report is required under the National Environmental Policy Act. The long-range goal of the US Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. 59 refs., 39 figs., 45 tabs.« less

Advances in stellarator gyrokinetics

NASA Astrophysics Data System (ADS)

Helander, P.; Bird, T.; Jenko, F.; Kleiber, R.; Plunk, G. G.; Proll, J. H. E.; Riemann, J.; Xanthopoulos, P.

2015-05-01

Recent progress in the gyrokinetic theory of stellarator microinstabilities and turbulence simulations is summarized. The simulations have been carried out using two different gyrokinetic codes, the global particle-in-cell code EUTERPE and the continuum code GENE, which operates in the geometry of a flux tube or a flux surface but is local in the radial direction. Ion-temperature-gradient (ITG) and trapped-electron modes are studied and compared with their counterparts in axisymmetric tokamak geometry. Several interesting differences emerge. Because of the more complicated structure of the magnetic field, the fluctuations are much less evenly distributed over each flux surface in stellarators than in tokamaks. Instead of covering the entire outboard side of the torus, ITG turbulence is localized to narrow bands along the magnetic field in regions of unfavourable curvature, and the resulting transport depends on the normalized gyroradius ρ* even in radially local simulations. Trapped-electron modes can be significantly more stable than in typical tokamaks, because of the spatial separation of regions with trapped particles from those with bad magnetic curvature. Preliminary non-linear simulations in flux-tube geometry suggest differences in the turbulence levels in Wendelstein 7-X and a typical tokamak.

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

Understanding L-H transition in tokamak fusion plasmas

NASA Astrophysics Data System (ADS)

Xu, Guosheng; Wu, Xingquan

2017-03-01

This paper reviews the current state of understanding of the L-H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L-H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L-H transition. We uncover a comprehensive physical picture of the L-H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L-H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.

Development of Data Acquisition Card Driver for ICRH System on EAST

NASA Astrophysics Data System (ADS)

Liu, Daming; Luo, Jiarong; Zhao, Yanping; Qin, Chengming

2008-04-01

Presented in this paper is the development of the driver for the data acquisition card with a peripheral component interconnection (PCI) local bus on the ion cyclotron range of frequency heating (ICRH) system. The driver is mainly aimed at the embedded VxWorks system (real-time operating system) which is widely used in various fields of real-time systems. An efficient way is employed to develop this driver, which will advance the real-time control of the ICRH system on the experimental advanced superconductor tokamak (EAST). The driver is designed using the TORNADO integrated development environment (IDE), and implemented in C plus language. The details include the hardware configuration, analogue/digital (A/D) and digital/analogue (D/A) conversion, input and output (I/O) operation of the driver to support over five cards. The data acquisition card can be manipulated in a low-level program and meet the requirements of A/D conversion and D/A outputs.

Relaunch of the Interactive Plasma Physics Educational Experience (IPPEX)

NASA Astrophysics Data System (ADS)

Dominguez, A.; Rusaitis, L.; Zwicker, A.; Stotler, D. P.

2015-11-01

In the late 1990's PPPL's Science Education Department developed an innovative online site called the Interactive Plasma Physics Educational Experience (IPPEX). It featured (among other modules) two Java based applications which simulated tokamak physics: A steady state tokamak (SST) and a time dependent tokamak (TDT). The physics underlying the SST and the TDT are based on the ASPECT code which is a global power balance code developed to evaluate the performance of fusion reactor designs. We have relaunched the IPPEX site with updated modules and functionalities: The site itself is now dynamic on all platforms. The graphic design of the site has been modified to current standards. The virtual tokamak programming has been redone in Javascript, taking advantage of the speed and compactness of the code. The GUI of the tokamak has been completely redesigned, including more intuitive representations of changes in the plasma, e.g., particles moving along magnetic field lines. The use of GPU accelerated computation provides accurate and smooth visual representations of the plasma. We will present the current version of IPPEX as well near term plans of incorporating real time NSTX-U data into the simulation.

Recent Progress on Spherical Torus Research and Implications for Fusion Energy Development Path

NASA Astrophysics Data System (ADS)

Ono, Masayuki

2014-10-01

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A =R0 / a) reduced to A near 1.5, well below the normal tokamak operating range of A equal to 2.5 or greater. As the aspect ratio is reduced, the ideal tokamak beta (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural plasma elongation which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to the longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in the UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all areas of fusion research, including fundamental fusion energy science as well as technological innovation. These results suggest exciting future prospects for ST research in both the near and longer term. The talk will summarize the key physics results from worldwide ST experiments, and describe ST community plans to provide the database for FNSF design while improving predictive capabilities for ITER and beyond. This work supported by DoE Contract No. DE-AC02-09CH11466.

Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

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

Yang, T.F.; Lee, A.Y.; Ruck, G.W.

1979-01-25

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm/sup 2/) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm/sup 2/) and for ISX-B/sup 2/ (11 kA/cm/sup 2/). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure.

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.

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

Fusion energy division annual progress report, period ending December 31, 1980

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

Not Available

1981-11-01

The ORNL Program encompasses most aspects of magnetic fusion research including research on two magnetic confinement programs (tokamaks and ELMO bumpy tori); the development of the essential technologies for plasma heating, fueling, superconducting magnets, and materials; the development of diagnostics; the development of atomic physics and radiation effect data bases; the assessment of the environmental impact of magnetic fusion; the physics and engineering of present-generation devices; and the design of future devices. The integration of all of these activities into one program is a major factor in the success of each activity. An excellent example of this integration is themore » extremely successful application of neutral injection heating systems developed at ORNL to tokamaks both in the Fusion Energy Division and at Princeton Plasma Physics Laboratory (PPPL). The goal of the ORNL Fusion Program is to maintain this balance between plasma confinement, technology, and engineering activities.« less

SOVRaD - A Digest of Recent Soviet R and D Articles. Volume 2, Number 6, 1976

DTIC Science & Technology

1976-06-01

6 Laser- Powered Rocket Model 1 High- Power CO2 Laser Radiation Effect in SF6 1 Tests With 9-Beam Laser Fusion Systems 1 Focusing Optics For...Boundary Layer 6 Deformation Theory of Artif.cial Muscles . 6 Dolphin Swimming Stereophotogrammetry 7 Stable Spark Gap for High- Power Pulsers 7...8 Resume of Soviet Tokamak Program .............. 9 First Measurements of Tokamak-10 Plasma , . . 10 Electrochemical Power Generation 11

The study of heat flux for disruption on experimental advanced superconducting tokamak

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

Yang, Zhendong, E-mail: dongyz@ipp.ac.cn, E-mail: jafang@dhu.edu.cn; Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031; Fang, Jianan, E-mail: dongyz@ipp.ac.cn, E-mail: jafang@dhu.edu.cn

Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptionsmore » have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dR{sub sep} = −2 cm, while it changes to upper single null (dR{sub sep} = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m{sup 2}.« less

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

Tierney, Brian; Dart, Eli; Tierney, Brian

The Energy Sciences Network (ESnet) is the primary provider of network connectivity for the U.S. Department of Energy Office of Science, the single largest supporter of basic research in the physical sciences in the United States of America. In support of the Office of Science programs, ESnet regularly updates and refreshes its understanding of the networking requirements of the instruments, facilities, scientists, and science programs that it serves. This focus has helped ESnet to be a highly successful enabler of scientific discovery for over 20 years. In March 2008, ESnet and the Fusion Energy Sciences (FES) Program Office of themore » DOE Office of Science organized a workshop to characterize the networking requirements of the science programs funded by the FES Program Office. Most sites that conduct data-intensive activities (the Tokamaks at GA and MIT, the supercomputer centers at NERSC and ORNL) show a need for on the order of 10 Gbps of network bandwidth for FES-related work within 5 years. PPPL reported a need for 8 times that (80 Gbps) in that time frame. Estimates for the 5-10 year time period are up to 160 Mbps for large simulations. Bandwidth requirements for ITER range from 10 to 80 Gbps. In terms of science process and collaboration structure, it is clear that the proposed Fusion Simulation Project (FSP) has the potential to significantly impact the data movement patterns and therefore the network requirements for U.S. fusion science. As the FSP is defined over the next two years, these changes will become clearer. Also, there is a clear and present unmet need for better network connectivity between U.S. FES sites and two Asian fusion experiments--the EAST Tokamak in China and the KSTAR Tokamak in South Korea. In addition to achieving its goal of collecting and characterizing the network requirements of the science endeavors funded by the FES Program Office, the workshop emphasized that there is a need for research into better ways of conducting remote collaboration with the control room of a Tokamak running an experiment. This is especially important since the current plans for ITER assume that this problem will be solved.« less

High beta and second stability region transport and stability analysis. Final report

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

Hughes, M.H.; Phillips, M.W.

1996-01-01

This report describes MHD equilibrium and stability studies carried out at Northrop Grumman`s Advanced Technology and Development Center during the period March 1 to December 31, 1995. Significant progress is reported in both ideal and resistive MHD modeling of TFTR plasmas. Specifically, attention is concentrated on analysis of Advanced Tokamak experiments at TFTR involving plasmas in which the q-profiles were non-monotonic.

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

NASA Astrophysics Data System (ADS)

Wang, Y.; Tobias, B.; Chang, Y.-T.; Yu, J.-H.; Li, M.; Hu, F.; Chen, M.; Mamidanna, M.; Phan, T.; Pham, A.-V.; Gu, J.; Liu, X.; Zhu, Y.; Domier, C. W.; Shi, L.; Valeo, E.; Kramer, G. J.; Kuwahara, D.; Nagayama, Y.; Mase, A.; Luhmann, N. C., Jr.

2017-07-01

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. Microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These have the potential to greatly advance microwave fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfvén eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today’s most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

DOE PAGES

Wang, Y.; Tobias, B.; Chang, Y. -T.; ...

2017-03-14

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. The microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These also have the potential to greatly advance microwavemore » fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfven eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today's most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.« less

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

ICRF fast wave current drive and mode conversion current drive in EAST tokamak

NASA Astrophysics Data System (ADS)

Yin, L.; Yang, C.; Gong, X. Y.; Lu, X. Q.; Du, D.; Chen, Y.

2017-10-01

Fast wave in the ion-cyclotron resonance frequency (ICRF) range is a promising candidate for non-inductive current drive (CD), which is essential for long pulse and high performance operation of tokamaks. A numerical study on the ICRF fast wave current drive (FWCD) and mode-conversion current drive (MCCD) in the Experimental Advanced Superconducting Tokamak (EAST) is carried out by means of the coupled full wave and Ehst-Karney parameterization methods. The results show that FWCD efficiency is notable in two frequency regimes, i.e., f ≥ 85 MHz and f = 50-65 MHz, where ion cyclotron absorption is effectively avoided, and the maximum on-axis driven current per unit power can reach 120 kA/MW. The sensitivity of the CD efficiency to the minority ion concentration is confirmed, owing to fast wave mode conversion, and the peak MCCD efficiency is reached for 22% minority-ion concentration. The effects of the wave-launch position and the toroidal wavenumber on the efficiency of current drive are also investigated.

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

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

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 LTX- β Research Program

NASA Astrophysics Data System (ADS)

Majeski, R.; Bell, R. E.; Boyle, D. P.; Hughes, P. E.; Kaita, R.; Kozub, T.; Merino, E.; Zhang, X.; Biewer, T. M.; Canik, J. M.; Elliott, D. B.; Reinke, M. L.; Bialek, J.; Hansen, C.; Jarboe, T.; Kubota, S.; Rhodes, T.; Dorf, M. A.; Rognlien, T.; Scotti, F.; Soukhanovskii, V. A.; Koel, B. E.; Donovan, D.; Maan, A.

2017-10-01

LTX- β, the upgrade to the Lithium Tokamak Experiment, approximately doubles the toroidal field (to 3.4 kG) and plasma current (to 150 - 175 kA) of LTX. Neutral beam injection at 20 kV, 30 A will be added in February 2018, with systems provided by Tri-Alpha Energy. A 9.3 GHz, 100 kW, short-pulse (5-10 msec) source will be available in summer 2018 for electron Bernstein wave heating. New lithium evaporation sources will allow between-shots recoating of the walls. Upgrades to the diagnostic set are intended to strengthen the research program in the critical areas of equilibrium, core transport, scrape-off layer physics, and plasma-material interactions. The LTX- β research program will combine the capability for gradient-free temperature profiles, to stabilize ion and electron temperature gradient-driven modes, with approaches to stabilization of ∇n-driven modes, such as the trapped electron mode (TEM). Candidate stabilization mechanisms for the TEM include sheared flow stabilization, which can be tested on LTX- β. The goal will be to minimize anomalous transport in a low aspect ratio tokamak, which would lead to a very compact, tokamak-based fusion core. This work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

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.

CXSFIT Code Application to Process Charge-Exchange Recombination Spectroscopy Data at the T-10 Tokamak

NASA Astrophysics Data System (ADS)

Serov, S. V.; Tugarinov, S. N.; Klyuchnikov, L. A.; Krupin, V. A.; von Hellermann, M.

2017-12-01

The applicability of the CXSFIT code to process experimental data from Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics at the T-10 tokamak is studied with a view to its further use for processing experimental data at the ITER facility. The design and operating principle of the CXRS diagnostics are described. The main methods for processing the CXRS spectra of the 5291-Å line of C5+ ions at the T-10 tokamak (with and without subtraction of parasitic emission from the edge plasma) are analyzed. The method of averaging the CXRS spectra over several shots, which is used at the T-10 tokamak to increase the signal-to-noise ratio, is described. The approximation of the spectrum by a set of Gaussian components is used to identify the active CXRS line in the measured spectrum. Using the CXSFIT code, the ion temperature in ohmic discharges and discharges with auxiliary electron cyclotron resonance heating (ECRH) at the T-10 tokamak is calculated from the CXRS spectra of the 5291-Å line. The time behavior of the ion temperature profile in different ohmic heating modes is studied. The temperature profile dependence on the ECRH power is measured, and the dynamics of ECR removal of carbon nuclei from the T-10 plasma is described. Experimental data from the CXRS diagnostics at T-10 substantially contribute to the implementation of physical programs of studies on heat and particle transport in tokamak plasmas and investigation of geodesic acoustic mode properties.

Intermittent bursts induced by double tearing mode reconnection

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

Wei, Lai; Wang, Zheng-Xiong, E-mail: zxwang@dlut.edu.cn

Reversed magnetic shear (RMS) configuration is assumed to be the steady-state operation scenario for the future advanced tokamaks like International Thermonuclear Experimental Reactor. In this work, we numerically discover a phenomenon of violent intermittent bursts induced by self-organized double tearing mode (DTM) reconnection in the RMS configuration during the very long evolution, which may continuously lead to annular sawtooth crashes and thus badly impact the desired steady-state operation of the future advanced RMS tokamaks. The key process of the intermittent bursts in the off-axis region is similar to that of the typical sawtooth relaxation oscillation in the positive magnetic shearmore » configuration. It is interestingly found that in the decay phase of the DTM reconnection, the zonal field significantly counteracts equilibrium field to make the magnetic shear between the two rational surfaces so weak that the residual self-generated vortices of the previous DTM burst are able to trigger a reverse DTM reconnection by curling the field lines.« less

Hydrocarbon deposition in gaps of tungsten and graphite tiles in Experimental Advanced Superconducting Tokamak edge plasma parameters

NASA Astrophysics Data System (ADS)

Xu, Qian; Yang, Zhongshi; Luo, Guang-Nan

2015-09-01

The three-dimensional (3D) Monte Carlo code PIC-EDDY has been utilized to investigate the mechanism of hydrocarbon deposition in gaps of tungsten tiles in the Experimental Advanced Superconducting Tokamak (EAST), where the sheath potential is calculated by the 2D in space and 3D in velocity particle-in-cell method. The calculated results for graphite tiles using the same method are also presented for comparison. Calculation results show that the amount of carbon deposited in the gaps of carbon tiles is three times larger than that in the gaps of tungsten tiles when the carbon particles from re-erosion on the top surface of monoblocks are taken into account. However, the deposition amount is found to be larger in the gaps of tungsten tiles at the same CH4 flux. When chemical sputtering becomes significant as carbon coverage on tungsten increases with exposure time, the deposition inside the gaps of tungsten tiles would be considerable.

Design of a collective scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research

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

Lee, W., E-mail: woochanglee@unist.ac.kr; Lee, D. J.; Park, H. K.

The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm{sup −1}.more » The upper limit corresponds to the normalized wavenumber k{sub θ}ρ{sub e} of ∼0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.« less

Modeling of Steady-state Scenarios for the Fusion Nuclear Science Facility, Advanced Tokamak Approach

NASA Astrophysics Data System (ADS)

Garofalo, A. M.; Chan, V. S.; Prater, R.; Smith, S. P.; St. John, H. E.; Meneghini, O.

2013-10-01

A Fusion National Science Facility (FNSF) would complement ITER in addressing the community identified science and technology gaps to a commercially attractive DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. Steady-state plasma operation is highly desirable to address the requirements for fusion nuclear technology testing [1]. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues with a more compact device, and the direct relevance to an attractive target power plant. Key features of AT are fully noninductive current drive, strong plasma cross section shaping, internal profiles consistent with high bootstrap fraction, and operation at high beta, typically above the free boundary limit, βN > 3 . Work supported by GA IR&D funding, DE-FC02-04ER54698, and DE-FG02-95ER43309.

Modeling of the control of the driven current profile in ICRF MCCD on EAST plasma

NASA Astrophysics Data System (ADS)

Yin, L.; Yang, C.; Gong, X. Y.; Lu, X. Q.; Cao, J. J.; Wu, Z. Y.; Chen, Y.; Du, D.

2018-05-01

Control of the current profile is a crucial issue for improved confinement and the inhibition of instability in advanced tokamak operation. Using typical discharge data for the Experimental Advanced Superconducting Tokamak, numerical simulations of driven-current profile control in mode conversion current drive (MCCD) in the ion cyclotron range of frequencies were performed employing a full-wave method and Ehst-Karney efficiency formula. Results indicate that the driven current profile in MCCD can be effectively modified by shifting the mode conversion layer. The peak of the driven current can be located at an aimed position in the normalized minor radius range (-0.60 ≤r/a≤0) by changing the radiofrequency and the minority-ion concentration. The efficiency of the off-axis MCCD can reach 233 kA/MW through optimization, and the mode converted ion cyclotron wave plays an important role in such scenarios. The effects of electron temperature and plasma density on the driven current profile are also investigated.

Intermittent bursts induced by double tearing mode reconnection

NASA Astrophysics Data System (ADS)

Wei, Lai; Wang, Zheng-Xiong

2014-06-01

Reversed magnetic shear (RMS) configuration is assumed to be the steady-state operation scenario for the future advanced tokamaks like International Thermonuclear Experimental Reactor. In this work, we numerically discover a phenomenon of violent intermittent bursts induced by self-organized double tearing mode (DTM) reconnection in the RMS configuration during the very long evolution, which may continuously lead to annular sawtooth crashes and thus badly impact the desired steady-state operation of the future advanced RMS tokamaks. The key process of the intermittent bursts in the off-axis region is similar to that of the typical sawtooth relaxation oscillation in the positive magnetic shear configuration. It is interestingly found that in the decay phase of the DTM reconnection, the zonal field significantly counteracts equilibrium field to make the magnetic shear between the two rational surfaces so weak that the residual self-generated vortices of the previous DTM burst are able to trigger a reverse DTM reconnection by curling the field lines.

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.

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

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

Wang, Y.; Tobias, B.; Chang, Y. -T.

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. The microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These also have the potential to greatly advance microwavemore » fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfven eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today's most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.« less

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.

Optical layout and mechanical structure of polarimeter-interferometer system for Experimental Advanced Superconducting Tokamak

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

Zou, Z. Y.; Liu, H. Q., E-mail: hqliu@ipp.ac.cn; Jie, Y. X.

A Far-InfaRed (FIR) three-wave POlarimeter-INTerferometer (POINT) system for measurement current density profile and electron density profile is under development for the EAST tokamak. The FIR beams are transmitted from the laser room to the optical tower adjacent to EAST via ∼20 m overmoded dielectric waveguide and then divided into 5 horizontal chords. The optical arrangement was designed using ZEMAX, which provides information on the beam spot size and energy distribution throughout the optical system. ZEMAX calculations used to optimize the optical layout design are combined with the mechanical design from CATIA, providing a 3D visualization of the entire POINT system.

Optical layout and mechanical structure of polarimeter-interferometer system for Experimental Advanced Superconducting Tokamak.

PubMed

Zou, Z Y; Liu, H Q; Jie, Y X; Ding, W X; Brower, D L; Wang, Z X; Shen, J S; An, Z H; Yang, Y; Zeng, L; Wei, X C; Li, G S; Zhu, X; Lan, T

2014-11-01

A Far-InfaRed (FIR) three-wave POlarimeter-INTerferometer (POINT) system for measurement current density profile and electron density profile is under development for the EAST tokamak. The FIR beams are transmitted from the laser room to the optical tower adjacent to EAST via ∼20 m overmoded dielectric waveguide and then divided into 5 horizontal chords. The optical arrangement was designed using ZEMAX, which provides information on the beam spot size and energy distribution throughout the optical system. ZEMAX calculations used to optimize the optical layout design are combined with the mechanical design from CATIA, providing a 3D visualization of the entire POINT system.

The engineering design of the Tokamak Physics Experiment

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

Schmidt, J.A.

A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design.more » If adequately funded the construction project should be completed in the year 2000.« less

Physical properties of Ce-TZP at cryogenic temperature

NASA Astrophysics Data System (ADS)

Han, Y. M.; Chen, Z.; Zhou, M.; Huang, R. J.; Huang, C. J.; Li, L. F.

2014-01-01

Electrical insulators, which are used to insulate cryogenic supply lines and conductor windings, are critical units in superconducting TOKAMAK magnets. Electrical insulators used in superconducting magnets fall into axial and radial insulators. These insulators can be made from glass ribbon epoxy densification and have been used in the Experiment Advanced Superconducting Tokamak (EAST). The properties of Ce-TZP can satisfy the requirement of electrical insulators. In this paper, thermal conductivity, mechanical properties and coefficient of thermal expansion of Ce-TZP have been investigated at cryogenic temperatures. Results indicate that the Ce-TZP shows better properties than epoxy and it demonstrates that the Ce-TZP can be used as insulation material in superconducting magnets.

Prospects for Advanced Tokamak Operation of ITER

NASA Astrophysics Data System (ADS)

Neilson, George H.

1996-11-01

Previous studies have identified steady-state (or "advanced") modes for ITER, based on reverse-shear profiles and significant bootstrap current. A typical example has 12 MA of plasma current, 1,500 MW of fusion power, and 100 MW of heating and current-drive power. The implementation of these and other steady-state operating scenarios in the ITER device is examined in order to identify key design modifications that can enhance the prospects for successfully achieving advanced tokamak operating modes in ITER compatible with a single null divertor design. In particular, we examine plasma configurations that can be achieved by the ITER poloidal field system with either a monolithic central solenoid (as in the ITER Interim Design), or an alternate "hybrid" central solenoid design which provides for greater flexibility in the plasma shape. The increased control capability and expanded operating space provided by the hybrid central solenoid allows operation at high triangularity (beneficial for improving divertor performance through control of edge-localized modes and for increasing beta limits), and will make it much easier for ITER operators to establish an optimum startup trajectory leading to a high-performance, steady-state scenario. Vertical position control is examined because plasmas made accessible by the hybrid central solenoid can be more elongated and/or less well coupled to the conducting structure. Control of vertical-displacements using the external PF coils remains feasible over much of the expanded operating space. Further work is required to define the full spectrum of axisymmetric plasma disturbances requiring active control In addition to active axisymmetric control, advanced tokamak modes in ITER may require active control of kink modes on the resistive time scale of the conducting structure. This might be accomplished in ITER through the use of active control coils external to the vacuum vessel which are actuated by magnetic sensors near the first wall. The enhanced shaping and positioning flexibility provides a range of options for reducing the ripple-induced losses of fast alpha particles--a major limitation on ITER steady-state modes. An alternate approach that we are pursuing in parallel is the inclusion of ferromagnetic inserts to reduce the toroidal field ripple within the plasma chamber. The inclusion of modest design changes such as the hybrid central solenoid, active control coils for kink modes, and ferromagnetic inserts for TF ripple reduction show can greatly increase the flexibility to accommodate advance tokamak operation in ITER. Increased flexibility is important because the optimum operating scenario for ITER cannot be predicted with certainty. While low-inductance, reverse shear modes appear attractive for steady-state operation, high-inductance, high-beta modes are also viable candidates, and it is important that ITER have the flexibility to explore both these, and other, operating regimes.

Plasma confinement theory and transport simulation

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

Ross, D.W.

The objectives are: (1) to advance the transport studies of tokamaks, including development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for TEXT-Upgrade. Recent reports, publications, and conference presentations of the Fusion Research Center are listed.

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

Sen, Amiya K.

The goal of this grant has been to study the basic physics of various sources of anomalous transport in tokamaks. Anomalous transport in tokamaks continues to be one of the major problems in magnetic fusion research. As a tokamak is not a physics device by design, direct experimental observation and identification of the instabilities responsible for transport, as well as physics studies of the transport in tokamaks, have been difficult and of limited value. It is noted that direct experimental observation, identification and physics study of microinstabilities including ITG, ETG, and trapped electron/ion modes in tokamaks has been very difficultmore » and nearly impossible. The primary reasons are co-existence of many instabilities, their broadband fluctuation spectra, lack of flexibility for parameter scans and absence of good local diagnostics. This has motivated us to study the suspected tokamak instabilities and their transport consequences in a simpler, steady state Columbia Linear Machine (CLM) with collisionless plasma and the flexibility of wide parameter variations. Earlier work as part of this grant was focused on both ITG turbulence, widely believed to be a primary source of ion thermal transport in tokamaks, and the effects of isotope scaling on transport levels. Prior work from our research team has produced and definitively identified both the slab and toroidal branches of this instability and determined the physics criteria for their existence. All the experimentally observed linear physics corroborate well with theoretical predictions. However, one of the large areas of research dealt with turbulent transport results that indicate some significant differences between our experimental results and most theoretical predictions. Latter years of this proposal were focused on anomalous electron transport with a special focus on ETG. There are several advanced tokamak scenarios with internal transport barriers (ITB), when the ion transport is reduced to neoclassical values by combined mechanisms of ExB and diamagnetic flow shear suppression of the ion temperature gradient (ITG) instabilities. However, even when the ion transport is strongly suppressed, the electron transport remains highly anomalous. The most plausible physics scenario for the anomalous electron transport is based on electron temperature gradient (ETG) instabilities. This instability is an electron analog of and nearly isomorphic to the ITG instability, which we had studied before extensively. However, this isomorphism is broken nonlinearily. It is noted that as the typical ETG mode growth rates are larger (in contrast to ITG modes) than ExB shearing rates in usual tokamaks, the flow shear suppression of ETG modes is highly unlikely. This motivated a broader range of investigations of other physics scenarios of nonlinear saturation and transport scaling of ETG modes.« less

The Spherical Tokamak MEDUSA for Mexico

NASA Astrophysics Data System (ADS)

Ribeiro, C.; Salvador, M.; Gonzalez, J.; Munoz, O.; Tapia, A.; Arredondo, V.; Chavez, R.; Nieto, A.; Gonzalez, J.; Garza, A.; Estrada, I.; Jasso, E.; Acosta, C.; Briones, C.; Cavazos, G.; Martinez, J.; Morones, J.; Almaguer, J.; Fonck, R.

2011-10-01

The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14m, a < 0.10m, BT < 0.5T, Ip < 40kA, 3ms pulse) is currently being recomissioned at the Universidad Autónoma de Nuevo León, Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

A Real-Time Data Acquisition and Processing Framework Based on FlexRIO FPGA and ITER Fast Plant System Controller

NASA Astrophysics Data System (ADS)

Yang, C.; Zheng, W.; Zhang, M.; Yuan, T.; Zhuang, G.; Pan, Y.

2016-06-01

Measurement and control of the plasma in real-time are critical for advanced Tokamak operation. It requires high speed real-time data acquisition and processing. ITER has designed the Fast Plant System Controllers (FPSC) for these purposes. At J-TEXT Tokamak, a real-time data acquisition and processing framework has been designed and implemented using standard ITER FPSC technologies. The main hardware components of this framework are an Industrial Personal Computer (IPC) with a real-time system and FlexRIO devices based on FPGA. With FlexRIO devices, data can be processed by FPGA in real-time before they are passed to the CPU. The software elements are based on a real-time framework which runs under Red Hat Enterprise Linux MRG-R and uses Experimental Physics and Industrial Control System (EPICS) for monitoring and configuring. That makes the framework accord with ITER FPSC standard technology. With this framework, any kind of data acquisition and processing FlexRIO FPGA program can be configured with a FPSC. An application using the framework has been implemented for the polarimeter-interferometer diagnostic system on J-TEXT. The application is able to extract phase-shift information from the intermediate frequency signal produced by the polarimeter-interferometer diagnostic system and calculate plasma density profile in real-time. Different algorithms implementations on the FlexRIO FPGA are compared in the paper.

Energetic particle-driven compressional Alfvén eigenmodes and prospects for ion cyclotron emission studies in fusion plasmas

NASA Astrophysics Data System (ADS)

Gorelenkov, N. N.

2016-10-01

As a fundamental plasma oscillation the compressional Alfvén waves (CAWs) are interesting for plasma scientists both academically and in applications for fusion plasmas. They are believed to be responsible for the ion cyclotron emission (ICE) observed in many tokamaks. The theory of CAW and ICE was significantly advanced at the end of 20th century in particular motivated by first DT experiments on TFTR and subsequent JET DT experimental studies. More recently, ICE theory was advanced by ST (or spherical torus) experiments with the detailed theoretical and experimental studies of the properties of each instability signal. There the instability responsible for ICE signals previously indistinguishable in high aspect ratio tokamaks became the subjects of experimental studies. We discuss further the prospects of ICE theory and its applications for future burning plasma experiments such as the ITER tokamak-reactor prototype being build in France where neutrons and gamma rays escaping the plasma create extremely challenging conditions for fusion alpha particle diagnostics. This manuscript has been authored by Princeton University under Contract Number DE-AC02-09CH11466 with the US Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

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.

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.

Non-Solenoidal Startup Research Directions on the Pegasus Toroidal Experiment

NASA Astrophysics Data System (ADS)

Fonck, R. J.; Bongard, M. W.; Lewicki, B. T.; Reusch, J. A.; Winz, G. R.

2017-10-01

The Pegasus research program has been focused on developing a physical understanding and predictive models for non-solenoidal tokamak plasma startup using Local Helicity Injection (LHI). LHI employs strong localized electron currents injected along magnetic field lines in the plasma edge that relax through magnetic turbulence to form a tokamak-like plasma. Pending approval, the Pegasus program will address a broader, more comprehensive examination of non-solenoidal tokamak startup techniques. New capabilities may include: increasing the toroidal field to 0.6 T to support critical scaling tests to near-NSTX-U field levels; deploying internal plasma diagnostics; installing a coaxial helicity injection (CHI) capability in the upper divertor region; and deploying a modest (200-400 kW) electron cyclotron RF capability. These efforts will address scaling of relevant physics to higher BT, separate and comparative studies of helicity injection techniques, efficiency of handoff to consequent current sustainment techniques, and the use of ECH to synergistically improve the target plasma for consequent bootstrap and neutral beam current drive sustainment. This has an ultimate goal of validating techniques to produce a 1 MA target plasma in NSTX-U and beyond. Work supported by US DOE Grant DE-FG02-96ER54375.

Ultrafast two-dimensional lithium beam emission spectroscopy diagnostic on the EAST tokamak

NASA Astrophysics Data System (ADS)

Zoletnik, S.; Hu, G. H.; Tál, B.; Dunai, D.; Anda, G.; Asztalos, O.; Pokol, G. I.; Kálvin, S.; Németh, J.; Krizsanóczi, T.

2018-06-01

A diagnostic instrument is described for the Experimental Advanced Superconducting Tokamak (EAST) for the measurement of the edge plasma electron density profile and plasma turbulence properties. An accelerated neutral lithium beam is injected into the tokamak and the Doppler shifted 670.8 nm light emission of the Li2p-2s transition is detected. A novel compact setup is used, where the beam injection and observation take place from the same equatorial diagnostic port and radial-poloidal resolution is achieved with microsecond time resolution. The observation direction is optimized in order to achieve a sufficient Doppler shift of the beam light to be able to separate from the strong edge lithium line emission on this lithium coated device. A 250 kHz beam chopping technique is also demonstrated for the removal of background light. First results show the capability of measuring turbulence and its poloidal flow velocity in the scrape-off layer and edge region and the resolution of details of transient phenomena like edge localized modes with few microsecond time resolution.

Tokamak magneto-hydrodynamics and reference magnetic coordinates for simulations of plasma disruptions

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

Zakharov, Leonid E.; Li, Xujing

This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97–104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasmamore » electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.« less

The High Field Path to Practical Fusion Energy

NASA Astrophysics Data System (ADS)

Mumgaard, Robert; Whyte, D.; Greenwald, M.; Hartwig, Z.; Brunner, D.; Sorbom, B.; Marmar, E.; Minervini, J.; Bonoli, P.; Irby, J.; Labombard, B.; Terry, J.; Vieira, R.; Wukitch, S.

2017-10-01

We propose a faster, lower cost development path for fusion energy enabled by high temperature superconductors, devices at high magnetic field, innovative technologies and modern approaches to technology development. Timeliness, scale, and economic-viability are the drivers for fusion energy to combat climate change and aid economic development. The opportunities provided by high-temperature superconductors, innovative engineering and physics, and new organizational structures identified over the last few years open new possibilities for realizing practical fusion energy that could meet mid-century de-carbonization needs. We discuss re-factoring the fusion energy development path with an emphasis on concrete risk retirement strategies utilizing a modular approach based on the high-field tokamak that leverages the broader tokamak physics understanding of confinement, stability, and operational limits. Elements of this plan include development of high-temperature superconductor magnets, simplified immersion blankets, advanced long-leg divertors, a compact divertor test tokamak, efficient current drive, modular construction, and demountable magnet joints. An R&D plan culminating in the construction of an integrated pilot plant and test facility modeled on the ARC concept is presented.

Overview of RWM Stabilization and Other Experiments With New Internal Coils in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Jackson, G. L.; Evans, T. E.; La Haye, R. J.; Kellman, A. G.; Schaffer, M. J.; Scoville, J. T.; Strait, E. J.; Szymanski, D. D.; Bialek, J.; Garofalo, A. M.; Navratil, G. A.; Reimerdes, H.; Edgell, D. H.; Okabayashi, M.; Hatcher, R.

2003-10-01

A set of 12 single-turn internal coils (I-coils) has been installed and operated in the DIII-D tokamak. The primary purpose of these coils (A_coil = 1.1 m^2, I ≤,7 kA, d_wall = 1.47 cm) is to improve stabilization of the n=1 resistive wall mode (RWM), compared to the existing external C-coil set, especially for high βN advanced tokamak discharges in low toroidal rotation plasmas. The versatility of the I-coil set and its associated power systems allow for a variety of experiments: fast feedback stabilization of RWMs, dc error field correction, edge stochastic fields, n=1,2, or 3 toroidal magnetic braking, and MHD spectroscopy (0-60 Hz). The resonant field amplification from an applied n=1 field was found to be completely suppressed, demonstrating successfully the controllability with the new system. With the I-coils, the high βN regime (above the no wall limit) has been explored both with RWM feedback and with dynamic error field correction. Experiments on edge ergodization will also be discussed.

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.

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.

The Compact Ignition Tokamak and electron cyclotron heating: Description of need; assessment of prospects

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

Ignat, D.W.; Cohn, D.R.; Woskov, P.P.

1989-01-01

The CIT will benefit from auxiliary heating of 10 to 40 MW. The schedules of both the CIT construction project and the operating plan contain adequate time to develop and implement ECH systems based on the gyrotron and the induction free electron laser (IFEL). Each approach has advantages and is the object of R and D at the level of many millions of dollars per year. While the gyrotron is further advanced in terms of power and pulse length achieved, rapid progress is scheduled for the IFEL, including experiments on tokamaks. Plans of CIT, gyrotron, and IFEL make 1992 anmore » appropriate time frame to commit to one or both systems. 12 refs., 8 figs., 2 tabs.« less

Design and Manufacturing of the Kstar Tokamak Helium Refrigeration System

NASA Astrophysics Data System (ADS)

Dauguet, P.; Briend, P.; Abe, I.; Fauve, E.; Bernhardt, J. M.; Andrieu, F.; Beauvisage, J.

2008-03-01

The KSTAR (Korean Superconducting Tokamak Advanced Research) project makes intensive use of superconducting (SC) magnets operated at 4.4 K. The cold components of KSTAR require a forced flow of supercritical helium for magnets and structure, boiling liquid helium for current leads, and gaseous helium for thermal shields. A helium refrigeration system has been custom-designed for this project. The purpose of this paper is to give a brief overview of the proposed cryogenic system. The specified thermal loads for the different operating modes are presented. This specification results in the definition of a design mode for the refrigerator. The design and construction of the resulting 9 kW at 4.5-K Helium Refrigeration System (HSR) are presented.

Design of the radiation shielding for the time of flight enhanced diagnostics neutron spectrometer at Experimental Advanced Superconducting Tokamak

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

Du, T. F.; Chen, Z. J.; Peng, X. Y.

A radiation shielding has been designed to reduce scattered neutrons and background gamma-rays for the new double-ring Time Of Flight Enhanced Diagnostics (TOFED). The shielding was designed based on simulation with the Monte Carlo code MCNP5. Dedicated model of the EAST tokamak has been developed together with the emission neutron source profile and spectrum; the latter were simulated with the Nubeam and GENESIS codes. Significant reduction of background radiation at the detector can be achieved and this satisfies the requirement of TOFED. The intensities of the scattered and direct neutrons in the line of sight of the TOFED neutron spectrometermore » at EAST are studied for future data interpretation.« less

First Results of ELM Triggering With a Multichamber Lithium Granule Injector Into EAST Discharges

DOE PAGES

Sun, Z.; Lunsford, R.; Maingi, R.; ...

2017-12-12

A critical challenge facing the basic long-pulse H-mode for ITER is to control edge-localized modes (ELMs). A new method using a multichamber lithium (Li) granule injector (LGI) for ELM triggering experiments has been developed in Experimental Advanced Superconducting Tokamak (EAST). First experimental results of the control of ELMs are obtained in EAST with a tungsten divertor. It is found that the injector has good capacities, i.e., allowing good flexibilities in granule size selection, injection rate, and injection velocity. In conclusion, LGI has successfully triggered ELMs during the H-mode. These results indicate the LGI would be a promising method to controlmore » ELMs in long-pulse steady-state tokamaks.« less

Interactions of toroidally coupled tearing modes in the KSTAR tokamak

NASA Astrophysics Data System (ADS)

Kim, Gnan; Yun, Gunsu S.; Woo, Minho; Park, Hyeon K.; KSTAR Team2, the

2018-03-01

The evolutions of toroidally coupled radially-distant and radially-adjacent tearing modes are visualized in 2D in detail on the Korea superconducting tokamak for advanced research. The coupled tearing modes are in-phase on the out-board mid-plane and become destabilized or compete with each other depending on their spatial separation. When two coupled tearing modes are far apart, both are increasingly destabilized. On the other hand, when they become close to each other, one becomes stabilized while the other becomes destabilized. In such cases, an additional tearing mode is often formed on outer rational flux surface and the three tearing modes compete. The competitions suggest that spatial overlap (merging) of coupled magnetic islands is difficult.

First Results of ELM Triggering With a Multichamber Lithium Granule Injector Into EAST Discharges

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

Sun, Z.; Lunsford, R.; Maingi, R.

A critical challenge facing the basic long-pulse H-mode for ITER is to control edge-localized modes (ELMs). A new method using a multichamber lithium (Li) granule injector (LGI) for ELM triggering experiments has been developed in Experimental Advanced Superconducting Tokamak (EAST). First experimental results of the control of ELMs are obtained in EAST with a tungsten divertor. It is found that the injector has good capacities, i.e., allowing good flexibilities in granule size selection, injection rate, and injection velocity. In conclusion, LGI has successfully triggered ELMs during the H-mode. These results indicate the LGI would be a promising method to controlmore » ELMs in long-pulse steady-state tokamaks.« less

Physics and operation oriented activities in preparation of the JT-60SA tokamak exploitation

NASA Astrophysics Data System (ADS)

Giruzzi, G.; Yoshida, M.; Artaud, J. F.; Asztalos, Ö.; Barbato, E.; Bettini, P.; Bierwage, A.; Boboc, A.; Bolzonella, T.; Clement-Lorenzo, S.; Coda, S.; Cruz, N.; Day, Chr.; De Tommasi, G.; Dibon, M.; Douai, D.; Dunai, D.; Enoeda, M.; Farina, D.; Figini, L.; Fukumoto, M.; Galazka, K.; Galdon, J.; Garcia, J.; Garcia-Muñoz, M.; Garzotti, L.; Gil, C.; Gleason-Gonzalez, C.; Goodman, T.; Granucci, G.; Hayashi, N.; Hoshino, K.; Ide, S.; Imazawa, R.; Innocente, P.; Isayama, A.; Itami, K.; Joffrin, E.; Kamada, Y.; Kamiya, K.; Kawano, Y.; Kawashima, H.; Kobayashi, T.; Kojima, A.; Kubo, H.; Lang, P.; Lauber, Ph.; de la Luna, E.; Maget, P.; Marchiori, G.; Mastrostefano, S.; Matsunaga, G.; Mattei, M.; McDonald, D. C.; Mele, A.; Miyata, Y.; Moriyama, S.; Moro, A.; Nakano, T.; Neu, R.; Nowak, S.; Orsitto, F. P.; Pautasso, G.; Pégourié, B.; Pigatto, L.; Pironti, A.; Platania, P.; Pokol, G. I.; Ricci, D.; Romanelli, M.; Saarelma, S.; Sakurai, S.; Sartori, F.; Sasao, H.; Scannapiego, M.; Shimizu, K.; Shinohara, K.; Shiraishi, J.; Soare, S.; Sozzi, C.; Stępniewski, W.; Suzuki, T.; Suzuki, Y.; Szepesi, T.; Takechi, M.; Tanaka, K.; Terranova, D.; Toma, M.; Urano, H.; Vega, J.; Villone, F.; Vitale, V.; Wakatsuki, T.; Wischmeier, M.; Zagórski, R.

2017-08-01

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

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.

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.

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.

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

Finley, V.L.; Wiezcorek, M.A.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory (PPPL) for CY93. The report is prepared to provide the U.S. Department of Energy (DOE) and the public with information on the level of radioactive and non-radioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs that were undertaken in 1993. The objective of the Annual Site Environmental Report is to document evidence that DOE facility environmental protection programs adequately protect the environment and the public health. The Princeton Plasmamore » Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. In 1993, PPPL had both of its two large tokamak devices in operation; the Tokamak Fusion Test Reactor (TFTR) and the Princeton Beta Experiment-Modification (PBX-M). PBX-M completed its modifications and upgrades and resumed operation in November 1991. TFTR began the deuterium-tritium (D-T) experiments in December 1993 and set new records by producing over six million watts of energy. The engineering design phase of the Tokamak Physics Experiment (TPX), which replaced the cancelled Burning Plasma Experiment in 1992 as PPPL`s next machine, began in 1993 with the planned start up set for the year 2001. In 1993, the Environmental Assessment (EA) for the TFRR Shutdown and Removal (S&R) and TPX was prepared for submittal to the regulatory agencies.« less

$$\\mathscr{H}_2$$ optimal control techniques for resistive wall mode feedback in tokamaks

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

Clement, Mitchell; Hanson, Jeremy; Bialek, Jim

DIII-D experiments show that a new, advanced algorithm improves resistive wall mode (RWM) stability control in high performance discharges using external coils. DIII-D can excite strong, locked or nearly locked external kink modes whose rotation frequencies and growth rates are on the order of the magnetic ux di usion time of the vacuum vessel wall. The VALEN RWM model has been used to gauge the e ectiveness of RWM control algorithms in tokamaks. Simulations and experiments have shown that modern control techniques like Linear Quadratic Gaussian (LQG) control will perform better, using 77% less current, than classical techniques when usingmore » control coils external to DIII-D's vacuum vessel. Experiments were conducted to develop control of a rotating n = 1 perturbation using an LQG controller derived from VALEN and external coils. Feedback using this LQG algorithm outperformed a proportional gain only controller in these perturbation experiments over a range of frequencies. Results from high N experiments also show that advanced feedback techniques using external control coils may be as e ective as internal control coil feedback using classical control techniques.« less

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

Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

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

Li, Y. L.; Xu, G. S.; Wan, B. N.

A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structuresmore » were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.« less

$$\\mathscr{H}_2$$ optimal control techniques for resistive wall mode feedback in tokamaks

DOE PAGES

Clement, Mitchell; Hanson, Jeremy; Bialek, Jim; ...

2018-02-28

DIII-D experiments show that a new, advanced algorithm improves resistive wall mode (RWM) stability control in high performance discharges using external coils. DIII-D can excite strong, locked or nearly locked external kink modes whose rotation frequencies and growth rates are on the order of the magnetic ux di usion time of the vacuum vessel wall. The VALEN RWM model has been used to gauge the e ectiveness of RWM control algorithms in tokamaks. Simulations and experiments have shown that modern control techniques like Linear Quadratic Gaussian (LQG) control will perform better, using 77% less current, than classical techniques when usingmore » control coils external to DIII-D's vacuum vessel. Experiments were conducted to develop control of a rotating n = 1 perturbation using an LQG controller derived from VALEN and external coils. Feedback using this LQG algorithm outperformed a proportional gain only controller in these perturbation experiments over a range of frequencies. Results from high N experiments also show that advanced feedback techniques using external control coils may be as e ective as internal control coil feedback using classical control techniques.« less

NIMROD: A computational laboratory for studying nonlinear fusion magnetohydrodynamics

NASA Astrophysics Data System (ADS)

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

2003-05-01

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

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

Edge localized mode characteristics during edge localized mode mitigation by supersonic molecular beam injection in Korea Superconducting Tokamak Advanced Research

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

Lee, H. Y.; Hong, J. H.; Jang, J. H.

It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the D{sub α} amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in themore » electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase.« less

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

Lazarus, E; Peng, Yueng Kay Martin

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

ADX: a high field, high power density, Advanced Divertor test eXperiment

NASA Astrophysics Data System (ADS)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

2014-10-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

Observation of internal transport barrier in ELMy H-mode plasmas on the EAST tokamak

NASA Astrophysics Data System (ADS)

Yang, Y.; Gao, X.; Liu, H. Q.; Li, G. Q.; Zhang, T.; Zeng, L.; Liu, Y. K.; Wu, M. Q.; Kong, D. F.; Ming, T. F.; Han, X.; Wang, Y. M.; Zang, Q.; Lyu, B.; Li, Y. Y.; Duan, Y. M.; Zhong, F. B.; Li, K.; Xu, L. Q.; Gong, X. Z.; Sun, Y. W.; Qian, J. P.; Ding, B. J.; Liu, Z. X.; Liu, F. K.; Hu, C. D.; Xiang, N.; Liang, Y. F.; Zhang, X. D.; Wan, B. N.; Li, J. G.; Wan, Y. X.; EAST Team

2017-08-01

The internal transport barrier (ITB) has been obtained in ELMy H-mode plasmas by neutron beam injection and lower hybrid wave heating on the Experimental Advanced Superconducting Tokamak (EAST). The ITB structure has been observed in profiles of ion temperature, electron temperature, and electron density within ρ < 0.5. It was also observed that the ITB formation is stepwise. Due to the ITB formation, the confinement quality H 98y2 increases from 1 to 1.1 and the normalized beta, β N, increases from 1.5 to near 2. The fishbone activity observed during the ITB phase suggests the central safety factor q(0) ˜ 1. Transport coefficients are calculated by particle balance and power balance analysis, showing an obvious reduction after the ITB formation.

Summer Research Experiences with a Laboratory Tokamak

NASA Astrophysics Data System (ADS)

Farley, N.; Mauel, M.; Navratil, G.; Cates, C.; Maurer, D.; Mukherjee, S.; Shilov, M.; Taylor, E.

1998-11-01

Columbia University's Summer Research Program for Secondary School Science Teachers seeks to improve middle and high school student understanding of science. The Program enhances science teachers' understanding of the practice of science by having them participate for two consecutive summers as members of laboratory research teams led by Columbia University faculty. In this poster, we report the research and educational activities of two summer internships with the HBT-EP research tokamak. Research activities have included (1) computer data acquisition and the representation of complex plasma wave phenomena as audible sounds, and (2) the design and construction of pulsed microwave systems to experience the design and testing of special-purpose equipment in order to achieve a specific technical goal. We also present an overview of the positive impact this type of plasma research involvement has had on high school science teaching.

Synchrotron radiation intensity and energy of runaway electrons in EAST tokamak

NASA Astrophysics Data System (ADS)

Zhang, YK; Zhou, RJ; Hu, LQ; Chen, MW; Chao, Y.; EAST team

2018-05-01

Not Available Project supported by the National Natural Science Foundation of China (Grant Nos. 11775263 and 11405219), the JSPS-NRF-NSFC A3 Foresight Program in the Field of Plasma Physics, China (Grant No. 11261140328), and the National Magnetic Confnement Fusion Science Program of China (Grant No. 2015GB102004).

Formation of the internal transport barrier in KSTAR

NASA Astrophysics Data System (ADS)

Chung, J.; Kim, H. S.; Jeon, Y. M.; Kim, J.; Choi, M. J.; Ko, J.; Lee, K. D.; Lee, H. H.; Yi, S.; Kwon, J. M.; Hahn, S.-H.; Ko, W. H.; Lee, J. H.; Yoon, S. W.

2018-01-01

One of key objectives of tokamak experiments is the exploration of enhanced confinement regimes, and the access of the internal transport barrier (ITB) formation is dealt with an important physics issue in the most of major tokamaks. Also, the advanced tokamak scenario with ITB is expected to lead to a continuous reactor with high fusion power density. From that point of view, the formation of the ITB in KSTAR which is designed for long pulse operation capability is very important although its heating and current drive systems are not fully equipped yet. We have therefore assumed that an early injection of the full NBI power (∼5.5 MW) during the current ramp-up would give a chance to form an internal barrier if the plasma could stay in the L-mode. To avoid the H-mode transition, we have produced inboard limited plasmas with detaching from the both upper and lower divertors. Using this approach, an ITB formation during L-mode has been observed which shows improved core confinement. Ion and electron temperature profiles show the barrier clearly in the temperature, and it was sustained for about 7 s in the dedicated experiment. This is the first stationary ITB observed in a full superconducting tokamak. This operation scenario with the ITB could be an alternative way to achieve a high performance regime in KSTAR, and the length of the ITB discharge could be extended even longer. In this paper, we present the formation of the ITB using measured and simulated characteristic profiles.

A folded waveguide ICRF antenna for PBX-M and TFTR

NASA Astrophysics Data System (ADS)

Bigelow, T. S.; Carter, M. D.; Fogelman, C. H.; Yugo, J. J.; Baity, F. W.; Bell, G. L.; Gardner, W. L.; Goulding, R. H.; Hoffman, D. J.; Ryan, P. M.; Swain, D. W.; Taylor, D. J.; Wilson, R.; Bernabei, S.; Kugel, H.; Ono, M.

1996-02-01

The folded waveguide (FWG) antenna is an advanced ICRF launcher under development at ORNL that offers many significant advantages over current-strap type antennas. These features are particularly beneficial for reactor-relevant applications such as ITER and TPX. Previous tests of a development folded waveguide with a low density plasma load have shown a factor of 5 increase in power capability over loop antennas into similar plasma conditions. The performance and reliability of a FWG with an actual tokamak plasma load must now be verified for further acceptance of this concept. A 58 MHz, 4 MW folded waveguide is being designed and built for the PBX-M and TFTR tokamaks at Princeton Plasma Physics Laboratory. This design has a square cross-section that can be installed as either a fast wave (FW) or ion-Bernstein wave (IBW) launcher by 90° rotation. Two new features of the design are: a shorter quarter-wavelength resonator configuration and a rear-feed input power coupling loop. Loading calculations with a standard shorting plate indicate that a launched power level of 4 MW is possible on either machine. Mechanical and disruption force analysis indicates that bolted construction will withstand the disruption loads. An experimental program is planned to characterize the plasma loading, heating effectiveness, power capability, impurity generation and other factors for both FW and IBW cases. High power tests of the new configuration are being performed with a development FWG unit on RFTF at ORNL.

HBT-EP Program: MHD Dynamics and Active Control through 3D Fields and Currents

NASA Astrophysics Data System (ADS)

Navratil, G. A.; Bialek, J.; Brooks, J. W.; Byrne, P. J.; Desanto, S.; Levesque, J. P.; Mauel, M. E.; Stewart, I. G.; Hansen, C. J.

2017-10-01

The HBT-EP active mode control research program aims to: (i) advance understanding of the effects of 3D shaping on advanced tokamak fusion performance, (ii) resolve important MHD issues associated with disruptions, and (iii) measure and mitigate the effects of 3D scrape-off layer (SOL) currents through active and passive control of the plasma edge and conducting boundary structures. Comparison of kink mode structure and RMP response in circular versus diverted plasmas shows good agreement with DCON modeling. SOL current measurements have been used to study SOL current dynamics and current-sharing with the vacuum vessel wall during kink-mode growth and disruptions. A multi-chord extreme UV/soft X-ray array is being installed to provide detailed internal mode structure information. Internal local electrodes were used to apply local bias voltage at two radial locations to study the effect of rotation profile on MHD mode rotation and stability and radial current flow through the SOL. A GPU-based low latency control system using 96 inputs and 64 outputs to apply magnetic perturbations for active control of kink modes is extended to directly control the SOL currents for kink-mode control. An extensive array of SOL current monitors and edge drive electrodes are being installed for pioneering studies of helical edge current control. Supported by U.S. DOE Grant DE-FG02-86ER53222.

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

NASA Astrophysics Data System (ADS)

Zhu, Ping; Huang, Wenlong; Yan, Xingting

2016-10-01

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

Plasma confinement theory and transport simulation. Technical progress report, May 1, 1991--April 30, 1992

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

Ross, D.W.

The objectives are: (1) to advance the transport studies of tokamaks, including development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for TEXT-Upgrade. Recent reports, publications, and conference presentations of the Fusion Research Center are listed.

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

2016-06-02

A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

Outline for a Spheromak Proof of Principle Experiment

NASA Astrophysics Data System (ADS)

Woodruff, Simon; Macnab, Angus

2007-11-01

A possible means for reducing reactor core complexity and size (and hence cost) could lie with research into the Spheromak concept: a plasma ring with no coils linking the plasma. Much progress has been made in the last 20 years, and now tokamak-like confinement is being reported, with work focusing on understanding beta-limits, transport and novel means of generating magnetic fields both in sustained and pulsed scenarios. Spheromak research is maturing, with many experiments integrated into a national program to resolve well defined critical physics issues. This poster summarizes the work from the last 20 years both as a historical overview and an outline of the present status. A natural consequence is to suggest the possibility of a Next-Step Spheromak, or advanced Proof of Principle device that will build on recent success and address many of the remaining critical issues in preparation for a Spheromak BPX.

Improving spatial and spectral resolution of TCV Thomson scattering

NASA Astrophysics Data System (ADS)

Hawke, J.; Andrebe, Y.; Bertizzolo, R.; Blanchard, P.; Chavan, R.; Decker, J.; Duval, B.; Lavanchy, P.; Llobet, X.; Marlétaz, B.; Marmillod, P.; Pochon, G.; Toussaint, M.

2017-12-01

The recently completed MST2 upgrade to the Thomson scattering (TS) system on TCV (Tokamak à Configuration Variable) at the Swiss Plasma Center aims to provide an enhanced spatial and spectral resolution while maintaining the high level of diagnostic flexibility for the study of TCV plasmas. The MST2 (Medium Sized Tokamak) is a work program within the Eurofusion ITER physics department, aimed at exploiting Europe's medium sized tokamak programs for a better understanding of ITER physics. This upgrade to the TCV Thomson scattering system involved the installation of 40 new compact 5-channel spectrometers and modifications to the diagnostics fiber optic design. The complete redesign of the fiber optic backplane incorporates fewer larger diameter fibers, allowing for a higher resolution in both the core and edge of TCV plasmas along the laser line, with a slight decrease in the signal to noise ratio of Thomson measurements. The 40 new spectrometers added to the system are designed to cover the full range of temperatures expected in TCV, able to measure electron temperatures (Te) with high precision between (6 eV and 20 keV) . The design of these compact spectrometers stems originally from the design utilized in the MAST (Mega Amp Spherical Tokamak) TS system located in Oxfordshire, United Kingdom. This design was implemented on TCV with an overall layout of optical fibers and spectrometers to achieve an overall increase in the spatial resolution, specifically a resolution of approximately 1% of the minor radius within the plasma pedestal region. These spectrometers also enhance the diagnostic spectral resolution, especially within the plasma edge, due to the low Te measurement capabilities. These additional spectrometers allow for a much greater diagnostic flexibility, allowing for quality full Thomson profiles in 75% of TCV plasma configurations.

Ion temperature gradient driven transport in tokamaks with square shaping

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

Joiner, N.; Dorland, W.

2010-06-15

Advanced tokamak schemes which may offer significant improvement to plasma confinement on the usual large aspect ratio Dee-shaped flux surface configuration are of great interest to the fusion community. One possibility is to introduce square shaping to the flux surfaces. The gyrokinetic code GS2[Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1996)] is used to study linear stability and the resulting nonlinear thermal transport of the ion temperature gradient driven (ITG) mode in tokamak equilibria with square shaping. The maximum linear growth rate of ITG modes is increased by negative squareness (diamond shaping) and reduced by positive values (square shaping).more » The dependence of thermal transport produced by saturated ITG instabilities on squareness is not as clear. The overall trend follows that of the linear instability, heat and particle fluxes increase with negative squareness and decrease with positive squareness. This is contradictory to recent experimental results [Holcomb et al., Phys. Plasmas 16, 056116 (2009)] which show a reduction in transport with negative squareness. This may be reconciled as a reduction in transport (consistent with the experiment) is observed at small negative values of the squareness parameter.« less

Computational Study of Anomalous Transport in High Beta DIII-D Discharges with ITBs

NASA Astrophysics Data System (ADS)

Pankin, Alexei; Garofalo, Andrea; Grierson, Brian; Kritz, Arnold; Rafiq, Tariq

2015-11-01

The advanced tokamak scenarios require a large bootstrap current fraction and high β. These large values are often outside the range that occurs in ``conventional'' tokamak discharges. The GLF23, TGLF, and MMM transport models have been previously validated for discharges with parameters associated with ``conventional'' tokamak discharges. It has been demonstrated that the TGLF model under-predicts anomalous transport in high β DIII-D discharges [A.M. Garofalo et al. 2015 TTF Workshop]. In this research, the validity of MMM7.1 model [T. Rafiq et al. Phys. Plasmas 20 032506 (2013)] is tested for high β DIII-D discharges with low and high torque. In addition, the sensitivity of the anomalous transport to β is examined. It is shown that the MMM7.1 model over-predicts the anomalous transport in the DIII-D discharge 154406. In particular, a significant level of anomalous transport is found just outside the internal transport barrier. Differences in the anomalous transport predicted using TGLF and MMM7.1 are reviewed. Mechanisms for quenching of anomalous transport in the ITB regions of high-beta discharges are investigated. This research is supported by US Department of Energy.

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

Finley, V.L.; Wieczorek, M.A.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory (PPPL) for CY94. The report is prepared to provide the US Department of Energy (DOE) and the public with information on the level of radioactive and nonradioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs that were undertaken in 1994. The objective of the Annual Site Environmental Report is to document evidence that PPPL`s environmental protection programs adequately protect the environment and the public health. The Princeton Plasma Physicsmore » Laboratory has engaged in fusion energy research since 195 1. The long-range goal of the US Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. In 1994, PPPL had one of its two large tokamak devices in operation-the Tokamak Fusion Test Reactor (TFTR). The Princeton Beta Experiment-Modification or PBX-M completed its modifications and upgrades and resumed operation in November 1991 and operated periodically during 1992 and 1993; it did not operate in 1994 for funding reasons. In December 1993, TFTR began conducting the deuterium-tritium (D-T) experiments and set new records by producing over ten @on watts of energy in 1994. The engineering design phase of the Tokamak Physics Experiment (T?X), which replaced the cancelled Burning Plasma Experiment in 1992 as PPPL`s next machine, began in 1993 with the planned start up set for the year 2001. In December 1994, the Environmental Assessment (EA) for the TFTR Shutdown and Removal (S&R) and TPX was submitted to the regulatory agencies, and a finding of no significant impact (FONSI) was issued by DOE for these projects.« less

Pellet injector development at ORNL (Oak Ridge National Laboratory)

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Argo, B. E.; Baylor, L. R.; Combs, S. K.; Fehling, D. T.; Fisher, P. W.; Foster, C. A.; Foust, C. R.; Milora, S. L.; Qualls, A. L.

1990-09-01

Advanced plasma fueling systems for magnetic confinement experiments are under development at Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets to speeds in the kilometer-per-second range by either pneumatic (light-gas gun) or mechanical (centrifugal force) techniques. ORNL has recently provided a centrifugal pellet injector for the Tore Supra tokamak and a new, simplified, eight-shot pneumatic injector for the Advanced Toroidal Facility stellarator at ORNL. Hundreds of tritium and DT pellets were accelerated at the Tritium Systems Test Assembly facility at Los Alamos in 1988 to 1989. These experiments, done in a single-shot pipe-gun system, demonstrated the feasibility of forming and accelerating tritium pellets at low (sup 3)He levels. A new, tritium-compatible extruder mechanism is being designed for longer-pulse DT applications. Two-stage light-gas guns and electron beam rocket accelerators for speeds of the order of 2 to 10 km/s are also under development. Recently, a repeating, two-stage light-gas gun accelerated 10 surrogate pellets at a 1-Hz repetition rate to speeds in the range of 2 to 3 km/s; and the electron beam rocket accelerator completed initial feasibility and scaling experiments. ORNL has also developed conceptual designs of advanced plasma fueling systems for the Compact Ignition Tokamak and the International Thermonuclear Experimental Reactor.

Enhanced erosion of tungsten plasma-facing components subject to simultaneous heat pulses and deuterium plasma

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Doerner, R.; Tynan, G.

2009-04-01

When an ELM occurs in tokamaks, up to 30% of the pedestal energy can be deposited on the wall of the tokamak causing heating and material loss due to sublimation/evaporation and melt layer splashing of plasma-facing components (PFCs) and expansion of the ejected material into the plasma. A short-pulse laser system capable of reproducing the thermal load of an ELM heat pulse has been integrated into the existing PFC research program in PISCES, a laboratory facility capable of reproducing plasma-materials interactions expected during normal operation of large tokamaks. An Nd:YAG laser capable of delivering up to 1 J of energy over a 7 ns pulsewidth is used for the experiments. Laser heat pulse only, H +/D + plasma only, and laser plus plasma experiments were conducted and initial results indicate enhanced erosion of tungsten exposed to simultaneous plasma and heat pulses, as compared to exposure to separate plasma-only or heat pulse-only conditions.

Optimal control of a coupled partial and ordinary differential equations system for the assimilation of polarimetry Stokes vector measurements in tokamak free-boundary equilibrium reconstruction with application to ITER

NASA Astrophysics Data System (ADS)

Faugeras, Blaise; Blum, Jacques; Heumann, Holger; Boulbe, Cédric

2017-08-01

The modelization of polarimetry Faraday rotation measurements commonly used in tokamak plasma equilibrium reconstruction codes is an approximation to the Stokes model. This approximation is not valid for the foreseen ITER scenarios where high current and electron density plasma regimes are expected. In this work a method enabling the consistent resolution of the inverse equilibrium reconstruction problem in the framework of non-linear free-boundary equilibrium coupled to the Stokes model equation for polarimetry is provided. Using optimal control theory we derive the optimality system for this inverse problem. A sequential quadratic programming (SQP) method is proposed for its numerical resolution. Numerical experiments with noisy synthetic measurements in the ITER tokamak configuration for two test cases, the second of which is an H-mode plasma, show that the method is efficient and that the accuracy of the identification of the unknown profile functions is improved compared to the use of classical Faraday measurements.

Maximum entropy reconstruction of poloidal magnetic field and radial electric field profiles in tokamaks

NASA Astrophysics Data System (ADS)

Chen, Yihang; Xiao, Chijie; Yang, Xiaoyi; Wang, Tianbo; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

2017-10-01

The Laser-driven Ion beam trace probe (LITP) is a new diagnostic method for measuring poloidal magnetic field (Bp) and radial electric field (Er) in tokamaks. LITP injects a laser-driven ion beam into the tokamak, and Bp and Er profiles can be reconstructed using tomography methods. A reconstruction code has been developed to validate the LITP theory, and both 2D reconstruction of Bp and simultaneous reconstruction of Bp and Er have been attained. To reconstruct from experimental data with noise, Maximum Entropy and Gaussian-Bayesian tomography methods were applied and improved according to the characteristics of the LITP problem. With these improved methods, a reconstruction error level below 15% has been attained with a data noise level of 10%. These methods will be further tested and applied in the following LITP experiments. Supported by the ITER-CHINA program 2015GB120001, CHINA MOST under 2012YQ030142 and National Natural Science Foundation Abstract of China under 11575014 and 11375053.

Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2015-11-01

The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.

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.

Two-dimensional vacuum ultraviolet images in different MHD events on the EAST tokamak

NASA Astrophysics Data System (ADS)

Zhijun, WANG; Xiang, GAO; Tingfeng, MING; Yumin, WANG; Fan, ZHOU; Feifei, LONG; Qing, ZHUANG; EAST Team

2018-02-01

A high-speed vacuum ultraviolet (VUV) imaging telescope system has been developed to measure the edge plasma emission (including the pedestal region) in the Experimental Advanced Superconducting Tokamak (EAST). The key optics of the high-speed VUV imaging system consists of three parts: an inverse Schwarzschild-type telescope, a micro-channel plate (MCP) and a visible imaging high-speed camera. The VUV imaging system has been operated routinely in the 2016 EAST experiment campaign. The dynamics of the two-dimensional (2D) images of magnetohydrodynamic (MHD) instabilities, such as edge localized modes (ELMs), tearing-like modes and disruptions, have been observed using this system. The related VUV images are presented in this paper, and it indicates the VUV imaging system is a potential tool which can be applied successfully in various plasma conditions.

Effect of high Z impurities on the ignition and Lawson conditions for a thermonuclear reactor

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

Meade, D.M.

1973-06-01

The recent advances in plasma heating and confinement using Tokamak devices have produced plasmas which approach thermonuclear conditions. Substantial amounts (0.1 to 1%) of partially stripped high Z impurities have been observed in these discharges. These high Z impurities (Fe,Mo,W) are presumably due to plasma bombardment of the limiter and vacuum chamber wall. Since the plasma energy will be increasing sharply in the next sequence of experiments from approx. =1kJ in ST tokamak to approx. =3MJ in PLT and up to approx. =100MJ in a feasibility experiment, the bombardment of the wall and limiter will become increasingly important. In thismore » paper, the effects of high Z impurities on the ignition and Lawson conditions for a DT reactor are calculated. 7 refs., 2 figs.« less

Real-time MSE measurements for current profile control on KSTAR.

PubMed

De Bock, M F M; Aussems, D; Huijgen, R; Scheffer, M; Chung, J

2012-10-01

To step up from current day fusion experiments to power producing fusion reactors, it is necessary to control long pulse, burning plasmas. Stability and confinement properties of tokamak fusion reactors are determined by the current or q profile. In order to control the q profile, it is necessary to measure it in real-time. A real-time motional Stark effect diagnostic is being developed at Korean Superconducting Tokamak for Advanced Research for this purpose. This paper focuses on 3 topics important for real-time measurements: minimize the use of ad hoc parameters, minimize external influences and a robust and fast analysis algorithm. Specifically, we have looked into extracting the retardance of the photo-elastic modulators from the signal itself, minimizing the influence of overlapping beam spectra by optimizing the optical filter design and a multi-channel, multiharmonic phase locking algorithm.

Present Status of the KSTAR Superconducting Magnet System Development

NASA Astrophysics Data System (ADS)

Kim, Keeman; H, K. Park; K, R. Park; B, S. Lim; S, I. Lee; M, K. Kim; Y, Chu; W, H. Chung; S, H. Baek; J Y, Choi; H, Yonekawa; A, Chertovskikh; Y, B. Chang; J, S. Kim; C, S. Kim; D, J. Kim; N, H. Song; K, P. Kim; Y, J. Song; I, S. Woo; W, S. Han; D, K. Lee; Y, K. Oh; K, W. Cho; J, S. Park; G, S. Lee; H, J. Lee; T, K. Ko; S, J. Lee

2004-10-01

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC, the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation, the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.

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

Gorelenkov, Nikolai N

The area of energetic particle (EP) physics of fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by W.W. Heidbrink and G.J. Sadler [1]. That review coincided with the start of deuterium-tritium (DT) experiments on Tokamak Fusion Test reactor (TFTR) and full scale fusion alphas physics studies. Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the "sea" of Alfven eigenmodes (AE) in particular by the toroidicityinduced AEsmore » (TAE) modes and reversed shear Alfven (RSAE). In present paper we attempt a broad review of EP physics progress in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus) including helical/stellarator devices. Introductory discussions on basic ingredients of EP physics, i.e. particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others are given to help understanding the advanced topics of EP physics. At the end we cover important and interesting physics issues toward the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).« less

Preliminary investigation of the effects of lower hybrid power on asymmetric behaviors in the scrape-off layer in experimental advanced superconducting tokamak

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

Zhang, L.; Ding, B. J., E-mail: bjding@ipp.ac.cn; Li, M. H.

2014-02-15

The striations in front of the lower hybrid (LH) launcher have been observed during LH injection by a visible video camera in the Experimental Advanced Superconducting Tokamak. Edge density at the top of the LH launcher tends to be much larger in reversed magnetic field (B{sub t}) than that in the normal B{sub t}. To study the mechanisms of the observations, the diffusive-convective model is employed. Simulations show that the LH power makes the density in scrape-off layer asymmetric in poloidal direction with five density peaks. The locations of the striations are approximately in agreement with the locations of themore » density peaks in different directions of B{sub t}. Higher LH power strengths the asymmetry of the density and leads to a bad coupling which is in conflict with the experimental results showing a good coupling with a higher power. Furthermore, an ionization term is introduced into this model and the increase of edge density with LH power can be qualitatively explained. The simulations also show that the density peaks in front of the waveguides become clearer when taking into account gas puffing.« less

Physics of Intense Electron Current Sources for Helicity Injection

NASA Astrophysics Data System (ADS)

Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Lewicki, B. T.; Perry, J. M.; Redd, A. J.; Winz, G. R.

2014-10-01

DC helicity injection (HI) for non-solenoidal ST startup requires sources of current at the tokamak edge. Since the rate of HI scales with injection voltage, understanding of the physics setting injector impedance is necessary for a predictive model of the HI rate and subsequent growth of Ip. In Pegasus, arc plasma sources are used for current injection. They operate immersed in tokamak edge plasma, and are biased at ~1-2 kV with respect to the vessel to draw current densities J ~ 1 kA/cm2 from an arc plasma cathode. Prior to tokamak formation, impedance data manifests two regimes, one at low current (< 1 kA) with I ~V 3 / 2 , and a higher current mode where I ~V 1 / 2 holds. The impedance in the I ~V 3 / 2 regime is consistent with an electrostatic double layer. Current in the I ~V 1 / 2 regime is linear in arc gas fueling rate, suggesting a space-charge limit set by nedge. In the presence of tokamak plasmas, voltage oscillations of the order 100s of volts are measured during MHD relaxation activity. These fluctuations occur at the characteristic frequencies of the n = 1 and n = 0 MHD activity observed on magnetic probes, and are suggestive of dynamic activity found in LHI simulations in NIMROD. Advanced injector design techniques have allowed higher voltage operation. These include staged shielding to prevent external arcing, and shaped cathodes, which minimize the onset and material damage due to cathode spot formation. Work supported by US DOE Grant DE-FG02-96ER54375.

Profile control simulations and experiments on TCV: a controller test environment and results using a model-based predictive controller

NASA Astrophysics Data System (ADS)

Maljaars, E.; Felici, F.; Blanken, T. C.; Galperti, C.; Sauter, O.; de Baar, M. R.; Carpanese, F.; Goodman, T. P.; Kim, D.; Kim, S. H.; Kong, M.; Mavkov, B.; Merle, A.; Moret, J. M.; Nouailletas, R.; Scheffer, M.; Teplukhina, A. A.; Vu, N. M. T.; The EUROfusion MST1-team; The TCV-team

2017-12-01

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety factor profile (q-profile) and kinetic plasma parameters such as the plasma beta. This demands to establish reliable profile control routines in presently operational tokamaks. We present a model predictive profile controller that controls the q-profile and plasma beta using power requests to two clusters of gyrotrons and the plasma current request. The performance of the controller is analyzed in both simulation and TCV L-mode discharges where successful tracking of the estimated inverse q-profile as well as plasma beta is demonstrated under uncertain plasma conditions and the presence of disturbances. The controller exploits the knowledge of the time-varying actuator limits in the actuator input calculation itself such that fast transitions between targets are achieved without overshoot. A software environment is employed to prepare and test this and three other profile controllers in parallel in simulations and experiments on TCV. This set of tools includes the rapid plasma transport simulator RAPTOR and various algorithms to reconstruct the plasma equilibrium and plasma profiles by merging the available measurements with model-based predictions. In this work the estimated q-profile is merely based on RAPTOR model predictions due to the absence of internal current density measurements in TCV. These results encourage to further exploit model predictive profile control in experiments on TCV and other (future) tokamaks.

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

Thermonuclear Power Engineering: 60 Years of Research. What Comes Next?

NASA Astrophysics Data System (ADS)

Strelkov, V. S.

2017-12-01

This paper summarizes results of more than half a century of research of high-temperature plasmas heated to a temperature of more than 100 million degrees (104 eV) and magnetically insulated from the walls. The energy of light-element fusion can be used for electric power generation or as a source of fissionable fuel production (development of a fusion neutron source—FNS). The main results of studies of tokamak plasmas which were obtained in the Soviet Union with the greatest degree of thermal plasma isolation among all other types of devices are presented. As a result, research programs of other countries were redirected to tokamaks. Later, on the basis of the analysis of numerous experiments, the international fusion community gradually came to an opinion that it is possible to build a tokamak (ITER) with Q > 1 (where Q is the ratio of the fusion power to the external power injected into the plasma). The ITER program objective is to achieve Q = 1-10 for a discharge time of up to 1000 s. The implementation of this goal does not solve the problem of a steadystate operation. The solution to this problem is a reliable first wall and current generation. This is a task of the next fusion power plant construction stage, called DEMO. Comparison of DEMO and FNS parameters shows that, at this development stage, the operating parameters and conditions of these devices are identical.

Overview of the EUROfusion Medium Size Tokamak scientific program

NASA Astrophysics Data System (ADS)

Bernert, Matthias; Bolzonella, Tommaso; Coda, Stefano; Hakola, Antti; Meyer, Hendrik; Eurofusion Mst1 Team; Tcv Team; Mast-U Team; ASDEX Upgrade Team

2017-10-01

Under the EUROfusion MST1 program, coordinated experiments are conducted at three European medium sized tokamaks (ASDEX Upgrade, TCV and MAST-U). It complements the JET program for preparing a safe and efficient operation for ITER and DEMO. Work under MST1 benefits from cross-machine comparisons but also makes use of the unique capabilities of each device. For the 2017/2018 campaign 25 topic areas were defined targeting three main objectives: 1) Development towards an edge and wall compatible H-mode scenario with small or no ELMs. 2) Investigation of disruptions in order to achieve better predictions and improve avoidance or mitigation schemes. 3) Exploring conventional and alternative divertor configurations for future high P/R scenarios. This contribution will give an overview of the work done under MST1 exemplified by the highlight results for each top objective from the last campaigns, such as evaluation of natural small ELM scenarios, runaway mitigation and control, assessment of detachment in alternative divertor configurations and highly radiative scenarios. See author list of ``H. Meyer et al. 2017 Nucl. Fusion 57, 102014''.

The ITER project construction status

NASA Astrophysics Data System (ADS)

Motojima, O.

2015-10-01

The pace of the ITER project in St Paul-lez-Durance, France is accelerating rapidly into its peak construction phase. With the completion of the B2 slab in August 2014, which will support about 400 000 metric tons of the tokamak complex structures and components, the construction is advancing on a daily basis. Magnet, vacuum vessel, cryostat, thermal shield, first wall and divertor structures are under construction or in prototype phase in the ITER member states of China, Europe, India, Japan, Korea, Russia, and the United States. Each of these member states has its own domestic agency (DA) to manage their procurements of components for ITER. Plant systems engineering is being transformed to fully integrate the tokamak and its auxiliary systems in preparation for the assembly and operations phase. CODAC, diagnostics, and the three main heating and current drive systems are also progressing, including the construction of the neutral beam test facility building in Padua, Italy. The conceptual design of the Chinese test blanket module system for ITER has been completed and those of the EU are well under way. Significant progress has been made addressing several outstanding physics issues including disruption load characterization, prediction, avoidance, and mitigation, first wall and divertor shaping, edge pedestal and SOL plasma stability, fuelling and plasma behaviour during confinement transients and W impurity transport. Further development of the ITER Research Plan has included a definition of the required plant configuration for 1st plasma and subsequent phases of ITER operation as well as the major plasma commissioning activities and the needs of the accompanying R&D program to ITER construction by the ITER parties.

2015 Nuclear Fusion Prize acceptance speech

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

Goldston, R. J.

This is the 2015 Nuclear Fusion Prize acceptance speech of R.J. Goldston: It is a great pleasure to receive the 2015 Nuclear Fusion award for my work developing a heuristic drift-based model for the power scrape-off width in tokamaks. I was particularly pleased to receive the award from IAEA Director General Yukiya Amano, whose thoughtful leadership has advanced the cause of nuclear non-proliferation mightily.

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

Kang, C. S.; Lee, S. G., E-mail: sglee@nfri.re.kr; National Fusion Research Institute, Daejeon 305-806

The behavior of relativistic runaway electrons during Electron Cyclotron Resonance Heating (ECRH) discharges is investigated in the Korea Superconducting Tokamak Advanced Research device. The effect of the ECRH on the runaway electron population is discussed. Observations on the generation of superthermal electrons during ECRH will be reported, which will be shown to be consistent with existing theory for the development of a superthermal electron avalanche during ECRH [A. Lazaros, Phys. Plasmas 8, 1263 (2001)].

2015 Nuclear Fusion Prize acceptance speech

DOE PAGES

Goldston, R. J.

2016-12-19

This is the 2015 Nuclear Fusion Prize acceptance speech of R.J. Goldston: It is a great pleasure to receive the 2015 Nuclear Fusion award for my work developing a heuristic drift-based model for the power scrape-off width in tokamaks. I was particularly pleased to receive the award from IAEA Director General Yukiya Amano, whose thoughtful leadership has advanced the cause of nuclear non-proliferation mightily.

Magnetic control of magnetohydrodynamic instabilities in tokamaks

DOE PAGES

Strait, Edward J.

2014-11-24

Externally applied, non-axisymmetric magnetic fields form the basis of several relatively simple and direct methods to control magnetohydrodynamic (MHD) instabilities in a tokamak, and most present and planned tokamaks now include a set of non-axisymmetric control coils for application of fields with low toroidal mode numbers. Non-axisymmetric applied fields are routinely used to compensate small asymmetries ( δB/B ~ 10 -3 to 10 -4) of the nominally axisymmetric field, which otherwise can lead to instabilities through braking of plasma rotation and through direct stimulus of tearing modes or kink modes. This compensation may be feedback-controlled, based on the magnetic responsemore » of the plasma to the external fields. Non-axisymmetric fields are used for direct magnetic stabilization of the resistive wall mode — a kink instability with a growth rate slow enough that feedback control is practical. Saturated magnetic islands are also manipulated directly with non-axisymmetric fields, in order to unlock them from the wall and spin them to aid stabilization, or position them for suppression by localized current drive. Several recent scientific advances form the foundation of these developments in the control of instabilities. Most fundamental is the understanding that stable kink modes play a crucial role in the coupling of non-axisymmetric fields to the plasma, determining which field configurations couple most strongly, how the coupling depends on plasma conditions, and whether external asymmetries are amplified by the plasma. A major advance for the physics of high-beta plasmas ( β = plasma pressure/magnetic field pressure) has been the understanding that drift-kinetic resonances can stabilize the resistive wall mode at pressures well above the ideal-MHD stability limit, but also that such discharges can be very sensitive to external asymmetries. The common physics of stable kink modes has brought significant unification to the topics of static error fields at low beta and resistive wall modes at high beta. Furthermore, these and other scientific advances, and their application to control of MHD instabilities, will be reviewed with emphasis on the most recent results and their applicability to ITER.« less

Bench Test of the Vibration Compensation Interferometer for EAST Tokamak

NASA Astrophysics Data System (ADS)

Li, Gongshun; Yang, Yao; Liu, Haiqing; Jie, Yinxian; Zou, Zhiyong; Wang, Zhengxing; Zeng, Long; Wei, Xuechao; Li, Weiming; Lan, Ting; Zhu, Xiang; Liu, Yukai; Gao, Xiang

2016-02-01

A visible laser-based vibration compensation interferometer has recently been designed for the EAST tokamak and the bench test has been finished. The system was optimized for its installation on EAST. The value of the final optical power before the detectors without plasma has been calculated from the component bench test result, which is quite close to the measured value. A nanometer level displacement (of the order of the laser's wavelength) has been clearly measured by a modulation of piezoelectric ceramic unit, proving the system's capability. supported by the National Magnetic Confinement Fusion Program of China (Nos. 2014GB106002, 2014GB106003, 2014GB106004) and National Natural Science Foundation of China (Nos. 11105184, 11375237, 11505238)

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

Conceptual design and proof-of-principle testing of the real-time multispectral imaging system MANTIS

NASA Astrophysics Data System (ADS)

Vijvers, W. A. J.; Mumgaard, R. T.; Andrebe, Y.; Classen, I. G. J.; Duval, B. P.; Lipschultz, B.

2017-12-01

The Multispectral Advanced Narrowband Tokamak Imaging System (MANTIS) is proposed to resolve the steep temperature and density gradients in the scrape-off layer of tokamaks in real-time. The initial design is to deliver two-dimensional distributions of key plasma parameters of the TCV tokamak to a real-time control system in order to enable novel control strategies, while providing new insights into power exhaust physics in the full offline analysis. This paper presents the conceptual system design, the mechanical and optical design of a prototype that was built to assess the optical performance, and the results of the first proof-of-principle tests of the prototype. These demonstrate a central resolving power of 50-46 line pairs per millimeter (CTF50) in the first four channels. For the additional channels, the sharpness is a factor two worse for the odd channels (likely affected by sub-optimal alignment), while the even channels continue the trend observed for the first four channels of 3% degradation per channel. This is explained by the self-cancellation of off-axis aberrations, which is an attractive property of the chosen optical design. The results show that at least a 10-channel real-time multispectral imaging system is feasible.

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

Seo, Dongcheol; Peterson, B. J.; Lee, Seung Hun

The resistive bolometers have been successfully installed in the midplane of L-port in Korea Superconducting Tokamak Advanced Research (KSTAR) device. The spatial and temporal resolutions, 4.5 cm and {approx}1 kHz, respectively, enable us to measure the radial profile of the total radiated power from magnetically confined plasma at a high temperature through radiation and neutral particles. The radiated power was measured at all shots. Even at low plasma current, the bolometer signal was detectable. The electron cyclotron resonance heating (ECH) has been used in tokamak for ECH assisted start-up and plasma control by local heating and current drive. The detectorsmore » of resistive bolometer, near the antenna of ECH, are affected by electron cyclotron wave. The tomographic reconstruction, using the Phillips-Tikhonov regularization method, will be carried out for a major radial profile of the radiation emissivity of the circular cross-section plasma.« less

Preliminary consideration of CFETR ITER-like case diagnostic system.

PubMed

Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

2016-11-01

Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

Preliminary consideration of CFETR ITER-like case diagnostic system

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

Li, G. S.; Liu, Y. K.; Gao, X.

2016-11-15

Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basicmore » control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.« less

Initial operation of a newly developed multichord motional Stark effect diagnostic in KSTAR

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

Chung, J., E-mail: jinil@nfri.re.kr; Ko, J.; Wi, H.

2016-11-15

A photo-elastic modulator based 25-chord motional Stark effect (MSE) diagnostic has been successfully developed and commissioned in Korea Superconducting Tokamak Advanced Research. The diagnostic measures the radial magnetic pitch angle profile of the Stark splitting of a D-alpha line at 656.1 nm by the electric field associated with the neutral deuterium heating beam. A tangential view of the neutral beam provides a good spatial resolution of 1–3 cm for covering the major radius from 1.74 m to 2.28 m, and the time resolution is achieved at 10 ms. An in-vessel calibration before the vacuum closing as well as an inmore » situ calibration during the tokamak operation was performed by means of specially designed polarized lighting sources. In this work, we present the final design of the installed MSE diagnostic and the first results of the commissioning.« less

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

Liu, J. X., E-mail: jsliu9@berkeley.edu; Milbourne, T.; Bitter, M.

The implementation of advanced electron cyclotron emission imaging (ECEI) systems on tokamak experiments has revolutionized the diagnosis of magnetohydrodynamic (MHD) activities and improved our understanding of instabilities, which lead to disruptions. It is therefore desirable to have an ECEI system on the ITER tokamak. However, the large size of optical components in presently used ECEI systems have, up to now, precluded the implementation of an ECEI system on ITER. This paper describes a new optical ECEI concept that employs a single spherical mirror as the only optical component and exploits the astigmatism of such a mirror to produce an imagemore » with one-dimensional spatial resolution on the detector. Since this alternative approach would only require a thin slit as the viewing port to the plasma, it would make the implementation of an ECEI system on ITER feasible. The results obtained from proof-of-principle experiments with a 125 GHz microwave system are presented.« less

High-Gain High-Field Fusion Plasma

PubMed Central

Li, Ge

2015-01-01

A Faraday wheel (FW)—an electric generator of constant electrical polarity that produces huge currents—could be implemented in an existing tokamak to study high-gain high-field (HGHF) fusion plasma, such as the Experimental Advanced Superconducting Tokamak (EAST). HGHF plasma can be realized in EAST by updating its pulsed-power system to compress plasma in two steps by induction fields; high gains of the Lawson trinity parameter and fusion power are both predicted by formulating the HGHF plasma. Both gain rates are faster than the decrease rate of the plasma volume. The formulation is checked by earlier ATC tests. Good agreement between theory and tests indicates that scaling to over 10 T at EAST may be possible by two-step compressions with a compression ratio of the minor radius of up to 3. These results point to a quick new path of fusion plasma study, i.e., simulating the Sun by EAST. PMID:26507314

Burn Control Mechanisms in Tokamaks

NASA Astrophysics Data System (ADS)

Hill, M. A.; Stacey, W. M.

2015-11-01

Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

Development of laser-based technology for the routine first wall diagnostic on the tokamak EAST: LIBS and LIAS

NASA Astrophysics Data System (ADS)

Hu, Z.; Gierse, N.; Li, C.; Liu, P.; Zhao, D.; Sun, L.; Oelmann, J.; Nicolai, D.; Wu, D.; Wu, J.; Mao, H.; Ding, F.; Brezinsek, S.; Liang, Y.; Ding, H.; Luo, G.; Linsmeier, C.; EAST Team

2017-12-01

A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.

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 the vacuum vessel for RWM feedback. RWM feedback based on VALEN outperformed a classical control algorithm using external coils to suppress the normalized plasma response to a rotating n=1 perturbation applied by internal coils over a range of frequencies. This study describes the design, development and testing of the GPU based control hardware and algorithm along with its performance during experiment and simulation.

Two Strategic Decisions Facing Fusion

NASA Astrophysics Data System (ADS)

Baldwin, D. E.

1998-06-01

Two strategic decisions facing the U.S. fusion program are described. The first decision deals with the role and rationale of the tokamak within the U. S. fusion program, and it underlies the debate over our continuing role in the evolving ITER collaboration (mid-1998). The second decision concerns how to include Inertial Fusion Energy (IFE) as a viable part of the national effort to harness fusion energy.

Localized Electron Heating by Strong Guide-Field Magnetic Reconnection

NASA Astrophysics Data System (ADS)

Guo, Xuehan; Sugawara, Takumichi; Inomoto, Michiaki; Yamasaki, Kotaro; Ono, Yasushi; UTST Team

2015-11-01

Localized electron heating of magnetic reconnection was studied under strong guide-field (typically Bt 15Bp) using two merging spherical tokamak plasmas in Univ. Tokyo Spherical Tokamak (UTST) experiment. Our new slide-type two-dimensional Thomson scattering system documented for the first time the electron heating localized around the X-point. The region of high electron temperature, which is perpendicular to the magnetic field, was found to have a round shape with radius of 2 [cm]. Also, it was localized around the X-point and does not agree with that of energy dissipation term Et .jt . When we include a guide-field effect term Bt / (Bp + αBt) for Et .jt where α =√{ (vin2 +vout2) /v∥2 } , the energy dissipation area becomes localized around the X-point, suggesting that the electrons are accelerated by the reconnection electric field parallel to the magnetic field and thermalized around the X-point. This work was supported by JSPS A3 Foresight Program ``Innovative Tokamak Plasma Startup and Current Drive in Spherical Torus,'' a Grant-in-Aid from the Japan Society for the Promotion of Science (JSPS) Fellows 15J03758.

Electron cyclotron heating/current-drive system using high power tubes for QUEST spherical tokamak

NASA Astrophysics Data System (ADS)

Onchi, Takumi; Idei, H.; Hasegawa, M.; Nagata, T.; Kuroda, K.; Hanada, K.; Kariya, T.; Kubo, S.; Tsujimura, T. I.; Kobayashi, S.; Quest Team

2017-10-01

Electron cyclotron heating (ECH) is the primary method to ramp up plasma current non-inductively in QUEST spherical tokamak. A 28 GHz gyrotron is employed for short pulses, where the radio frequency (RF) power is about 300 kW. Current ramp-up efficiency of 0.5 A/W has been obtained with focused beam of the second harmonic X-mode. A quasi-optical polarizer unit has been newly installed to avoid arcing events. For steady-state tokamak operation, 8.56 GHz klystron with power of 200 kW is used as the CW-RF source. The high voltage power supply (54 kV/13 A) for the klystron has been built recently, and initial bench test of the CW-ECH system is starting. The array of insulated-gate bipolar transistor works to quickly cut off the input power for protecting the klystron. This work is supported by JSPS KAKENHI (15H04231), NIFS Collaboration Research program (NIFS13KUTR085, NIFS17KUTR128), and through MEXT funding for young scientists associated with active promotion of national university reforms.

Numerical Solution of the Electron Heat Transport Equation and Physics-Constrained Modeling of the Thermal Conductivity via Sequential Quadratic Programming Optimization in Nuclear Fusion Plasmas

NASA Astrophysics Data System (ADS)

Paloma, Cynthia S.

The plasma electron temperature (Te) plays a critical role in a tokamak nu- clear fusion reactor since temperatures on the order of 108K are required to achieve fusion conditions. Many plasma properties in a tokamak nuclear fusion reactor are modeled by partial differential equations (PDE's) because they depend not only on time but also on space. In particular, the dynamics of the electron temperature is governed by a PDE referred to as the Electron Heat Transport Equation (EHTE). In this work, a numerical method is developed to solve the EHTE based on a custom finite-difference technique. The solution of the EHTE is compared to temperature profiles obtained by using TRANSP, a sophisticated plasma transport code, for specific discharges from the DIII-D tokamak, located at the DIII-D National Fusion Facility in San Diego, CA. The thermal conductivity (also called thermal diffusivity) of the electrons (Xe) is a plasma parameter that plays a critical role in the EHTE since it indicates how the electron temperature diffusion varies across the minor effective radius of the tokamak. TRANSP approximates Xe through a curve-fitting technique to match experimentally measured electron temperature profiles. While complex physics-based model have been proposed for Xe, there is a lack of a simple mathematical model for the thermal diffusivity that could be used for control design. In this work, a model for Xe is proposed based on a scaling law involving key plasma variables such as the electron temperature (Te), the electron density (ne), and the safety factor (q). An optimization algorithm is developed based on the Sequential Quadratic Programming (SQP) technique to optimize the scaling factors appearing in the proposed model so that the predicted electron temperature and magnetic flux profiles match predefined target profiles in the best possible way. A simulation study summarizing the outcomes of the optimization procedure is presented to illustrate the potential of the proposed modeling method.

Measurements of the parallel wavenumber of lower hybrid waves in the scrape-off layer of a high-density tokamak

NASA Astrophysics Data System (ADS)

Baek, S. G.; Wallace, G. M.; Shinya, T.; Parker, R. R.; Shiraiwa, S.; Bonoli, P. T.; Brunner, D.; Faust, I.; LaBombard, B. L.; Takase, Y.; Wukitch, S.

2016-05-01

In lower hybrid current drive (LHCD) experiments on tokamaks, the parallel wavenumber of lower hybrid waves is an important physics parameter that governs the wave propagation and absorption physics. However, this parameter has not been experimentally well-characterized in the present-day high density tokamaks, despite the advances in the wave physics modeling. In this paper, we present the first measurement of the dominant parallel wavenumber of lower hybrid waves in the scrape-off layer (SOL) of the Alcator C-Mod tokamak with an array of magnetic loop probes. The electric field strength measured with the probe in typical C-Mod plasmas is about one-fifth of that of the electric field at the mouth of the grill antenna. The amplitude and phase responses of the measured signals on the applied power spectrum are consistent with the expected wave energy propagation. At higher density, the observed k|| increases for the fixed launched k||, and the wave amplitude decreases rapidly. This decrease is correlated with the loss of LHCD efficiency at high density, suggesting the presence of loss mechanisms. Evidence of the spectral broadening mechanisms is observed in the frequency spectra. However, no clear modifications in the dominant k|| are observed in the spectrally broadened wave components, as compared to the measured k|| at the applied frequency. It could be due to (1) the probe being in the SOL and (2) the limited k|| resolution of the diagnostic. Future experiments are planned to investigate the roles of the observed spectral broadening mechanisms on the LH density limit problem in the strong single pass damping regime.

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.

A Michelson Interferometer for Electron Cyclotron Emission Measurements on EAST

NASA Astrophysics Data System (ADS)

Liu, Yong; Stefan, Schmuck; Zhao, Hailin; John, Fessey; Paul, Trimble; Liu, Xiang; Zhu, Zeying; Zang, Qing; Hu, Liqun

2016-12-01

A Michelson interferometer, on loan from EFDA-JET (Culham, United Kingdom) has recently been commissioned on the experimental advanced superconducting tokamak (EAST, ASIPP, Hefei, China). Following a successful in-situ absolute calibration the instrument is able to measure the electron cyclotron emission (ECE) spectrum, from 80 GHz to 350 GHz in extraordinary mode (X-mode) polarization, with high accuracy. This allows the independent determination of the electron temperature profile from observation of the second harmonic ECE and the possible identification of non-Maxwellian features by comparing higher harmonic emission with numerical simulations. The in-situ calibration results are presented together with the initial measured temperature profiles. These measurements are then discussed and compared with other independent temperature profile measurements. This paper also describes the main hardware features of the diagnostic and the associated commissioning test results. supported by National Natural Science Foundation of China (Nos. 11405211, 11275233), and the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB106002, 2015GB101000), and the RCUK Energy Programme (No. EP/I501045), partly supported by the JSPS-NRF-NSFC A3 Foresight Program in the Field of Plasma Physics (NSFC: No. 11261140328)

Physics basis for an advanced physics and advanced technology tokamak power plant configuration: ARIES-ACT1

DOE PAGES

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

2015-01-01

Here, the advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, and triangularity of 0.63. The broadest pressure cases reached wall-stabilized β N ~ 5.75, limited by n = 3 external kink mode requiring a conducting shell at b/a = 0.3, requiring plasma rotation, feedback, and/or kinetic stabilization. The medium pressure peaking case reaches β N = 5.28 with B T = 6.75, while the peaked pressure case reaches β N < 5.15. Fast particle magnetohydrodynamicmore » stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling shows that 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while >95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring ~1.1 MA of external current drive. This current is supplied with 5 MW of ion cyclotron radio frequency/fast wave and 40 MW of lower hybrid current drive. Electron cyclotron is most effective for safety factor control over ρ~0.2 to 0.6 with 20 MW. The pedestal density is ~0.9×10 20/m 3, and the temperature is ~4.4 keV. The H98 factor is 1.65, n/n Gr = 1.0, and the ratio of net power to threshold power is 2.8 to 3.0 in the flattop.« less

Kinetic-MHD hybrid simulation of fishbone modes excited by fast ions on the experimental advanced superconducting tokamak (EAST)

NASA Astrophysics Data System (ADS)

Pei, Youbin; Xiang, Nong; Hu, Youjun; Todo, Y.; Li, Guoqiang; Shen, Wei; Xu, Liqing

2017-03-01

Kinetic-MagnetoHydroDynamic hybrid simulations are carried out to investigate fishbone modes excited by fast ions on the Experimental Advanced Superconducting Tokamak. The simulations use realistic equilibrium reconstructed from experiment data with the constraint of the q = 1 surface location (q is the safety factor). Anisotropic slowing down distribution is used to model the distribution of the fast ions from neutral beam injection. The resonance condition is used to identify the interaction between the fishbone mode and the fast ions, which shows that the fishbone mode is simultaneously in resonance with the bounce motion of the trapped particles and the transit motion of the passing particles. Both the passing and trapped particles are important in destabilizing the fishbone mode. The simulations show that the mode frequency chirps down as the mode reaches the nonlinear stage, during which there is a substantial flattening of the perpendicular pressure of fast ions, compared with that of the parallel pressure. For passing particles, the resonance remains within the q = 1 surface, while, for trapped particles, the resonant location moves out radially during the nonlinear evolution. In addition, parameter scanning is performed to examine the dependence of the linear frequency and growth rate of fishbones on the pressure and injection velocity of fast ions.

Simulations of toroidal Alfvén eigenmode excited by fast ions on the Experimental Advanced Superconducting Tokamak

NASA Astrophysics Data System (ADS)

Pei, Youbin; Xiang, Nong; Shen, Wei; Hu, Youjun; Todo, Y.; Zhou, Deng; Huang, Juan

2018-05-01

Kinetic-MagnetoHydroDynamic (MHD) hybrid simulations are carried out to study fast ion driven toroidal Alfvén eigenmodes (TAEs) on the Experimental Advanced Superconducting Tokamak (EAST). The first part of this article presents the linear benchmark between two kinetic-MHD codes, namely MEGA and M3D-K, based on a realistic EAST equilibrium. Parameter scans show that the frequency and the growth rate of the TAE given by the two codes agree with each other. The second part of this article discusses the resonance interaction between the TAE and fast ions simulated by the MEGA code. The results show that the TAE exchanges energy with the co-current passing particles with the parallel velocity |v∥ | ≈VA 0/3 or |v∥ | ≈VA 0/5 , where VA 0 is the Alfvén speed on the magnetic axis. The TAE destabilized by the counter-current passing ions is also analyzed and found to have a much smaller growth rate than the co-current ions driven TAE. One of the reasons for this is found to be that the overlapping region of the TAE spatial location and the counter-current ion orbits is narrow, and thus the wave-particle energy exchange is not efficient.

High-beta, steady-state hybrid scenario on DIII-D

DOE PAGES

Petty, C. C.; Kinsey, J. E.; Holcomb, C. T.; ...

2015-12-17

Here, the potential of the hybrid scenario (first developed as an advanced inductive scenario for high fluence) as a regime for high-beta, steady-state plasmas is demonstrated on the DIII-D tokamak. These experiments show that the beneficial characteristics of hybrids, namely safety factor ≥1 with low central magnetic shear, high stability limits and excellent confinement, are maintained when strong central current drive (electron cyclotron and neutral beam) is applied to increase the calculated non-inductive fraction to ≈100% (≈50% bootstrap current). The best discharges achieve normalized beta of 3.4, IPB98(y,2) confinement factor of 1.4, surface loop voltage of 0.01 V, and nearlymore » equal electron and ion temperatures at low collisionality. A zero-dimensional physics model shows that steady-state hybrid operation with Q fus ~ 5 is feasible in FDF and ITER. The advantage of the hybrid scenario as an Advanced Tokamak regime is that the external current drive can be deposited near the plasma axis where the efficiency is high; additionally, good alignment between the current drive and plasma current profiles is not necessary as the poloidal magnetic flux pumping self-organizes the current density profile in hybrids with an m/n=3/2 tearing mode.« less

New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

PubMed

Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q

2015-02-06

A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development.

Experimental investigation of density behaviors in front of the lower hybrid launcher in experimental advanced superconducting tokamak

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

Zhang, L.; Ding, B. J.; Li, M. H.

2013-06-15

A triple Langmuir probe is mounted on the top of the Lower Hybrid (LH) antenna to measure the electron density near the LH grills in Experimental Advanced Superconducting Tokamak. In this work, the LH power density ranges from 2.3 MWm{sup −2} to 10.3 MWm{sup −2} and the rate of puffing gas varies from 1.7 × 10{sup 20} el/s to 14 × 10{sup 20} el/s. The relation between the edge density (from 0.3 × n{sub e-cutoff} to 20 × n{sub e-cutoff}, where n{sub e-cutoff} is the cutoff density, n{sub e-cutoff} = 0.74 × 10{sup 17} m{sup −3} for 2.45 GHz lowermore » hybrid current drive) near the LH grill and the LH power reflection coefficients is investigated. The factors, including the gap between the LH grills and the last closed magnetic flux surface, line-averaged density, LH power, edge safety factor, and gas puffing, are analyzed. The experiments show that injection of LH power is beneficial for increasing edge density. Gas puffing is beneficial for increasing grill density but excess gas puffing is unfavorable for coupling and current drive.« less

Fishbone Mode Excited by Deeply Trapped Energetic Beam Ions in EAST

NASA Astrophysics Data System (ADS)

Zheng, Ting; Wu, Bin; Xu, Liqing; Hu, Chundong; Zang, Qing; Ding, Siye; Li, Yingying; Wu, Xingquan; Wang, Jinfang; Shen, Biao; Zhong, Guoqiang; Li, Hao; Shi, Tonghui; EAST Team

2016-06-01

This paper describes the fishbone mode phenomena during the injection of high-power neutral beams in EAST (Experimental Advanced Superconducting Tokamak). The features of the fishbone mode are presented. The change in frequency of the mode during a fishbone burst is from 1 kHz to 6 kHz. The nonlinear behavior of the fishbone mode is analyzed by using a prey-predator model, which is consistent with the experimental results. This model indicates that the periodic oscillations of the fishbone mode always occur near the critical value of fast ion beta. Furthermore, the neutral beam analysis for the discharge is done by using the NUBEAM module of the TRANSP code. According to the numerical simulation results and theoretical calculation, it can be concluded that the fishbone mode is driven by the deeply trapped energetic beam ions in EAST. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB101001, 2014DFG61950 and 2013GB112003) and National Natural Science Foundation of China (Nos. 11175211 and 11275233)

Control advances for achieving the ITER baseline scenario on KSTAR

NASA Astrophysics Data System (ADS)

Eidietis, N. W.; Barr, J.; Hahn, S. H.; Humphreys, D. A.; in, Y. K.; Jeon, Y. M.; Lanctot, M. J.; Mueller, D.; Walker, M. L.

2017-10-01

Control methodologies developed to enable successful production of ITER baseline scenario (IBS) plasmas on the superconducting KSTAR tokamak are presented: decoupled vertical control (DVC), real-time feedforward (rtFF) calculation, and multi-input multi-output (MIMO) X-point control. DVC provides fast vertical control with the in-vessel control coils (IVCC) while sharing slow vertical control with the poloidal field (PF) coils to avoid IVCC saturation. rtFF compensates for inaccuracies in offline PF current feedforward programming, allowing reduction or removal of integral gain (and its detrimental phase lag) from the shape controller. Finally, MIMO X-point control provides accurate positioning of the X-point despite low controllability due to the large distance between coils and plasma. Combined, these techniques enabled achievement of IBS parameters (q95 = 3.2, βN = 2) with a scaled ITER shape on KSTAR. n =2 RMP response displays a strong dependence upon this shaping. Work supported by the US DOE under Award DE-SC0010685 and the KSTAR project.

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.

Advances in simulation of wave interactions with extended MHD phenomena

NASA Astrophysics Data System (ADS)

Batchelor, D.; Abla, G.; D'Azevedo, E.; Bateman, G.; Bernholdt, D. E.; Berry, L.; Bonoli, P.; Bramley, R.; Breslau, J.; Chance, M.; Chen, J.; Choi, M.; Elwasif, W.; Foley, S.; Fu, G.; Harvey, R.; Jaeger, E.; Jardin, S.; Jenkins, T.; Keyes, D.; Klasky, S.; Kruger, S.; Ku, L.; Lynch, V.; McCune, D.; Ramos, J.; Schissel, D.; Schnack, D.; Wright, J.

2009-07-01

The Integrated Plasma Simulator (IPS) provides a framework within which some of the most advanced, massively-parallel fusion modeling codes can be interoperated to provide a detailed picture of the multi-physics processes involved in fusion experiments. The presentation will cover four topics: 1) recent improvements to the IPS, 2) application of the IPS for very high resolution simulations of ITER scenarios, 3) studies of resistive and ideal MHD stability in tokamk discharges using IPS facilities, and 4) the application of RF power in the electron cyclotron range of frequencies to control slowly growing MHD modes in tokamaks and initial evaluations of optimized location for RF power deposition.

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.

Advances in the FTU collective Thomson scattering system

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

Bin, W., E-mail: wbin@ifp.cnr.it; Bruschi, A.; Grosso, G.

The new collective Thomson scattering diagnostic installed on the Frascati Tokamak Upgrade device started its first operations in 2014. The ongoing experiments investigate the presence of signals synchronous with rotating tearing mode islands, possibly due to parametric decay processes, and phenomena affecting electron cyclotron beam absorption or scattering measurements. The radiometric system, diagnostic layout, and data acquisition system were improved accordingly. The present status and near-term developments of the diagnostic are presented.

Gyrokinetic Particle Simulation of Turbulent Transport in Burning Plasmas (GPS - TTBP) Final Report

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

Chame, Jacqueline

2011-05-27

The goal of this project is the development of the Gyrokinetic Toroidal Code (GTC) Framework and its applications to problems related to the physics of turbulence and turbulent transport in tokamaks,. The project involves physics studies, code development, noise effect mitigation, supporting computer science efforts, diagnostics and advanced visualizations, verification and validation. Its main scientific themes are mesoscale dynamics and non-locality effects on transport, the physics of secondary structures such as zonal flows, and strongly coherent wave-particle interaction phenomena at magnetic precession resonances. Special emphasis is placed on the implications of these themes for rho-star and current scalings and formore » the turbulent transport of momentum. GTC-TTBP also explores applications to electron thermal transport, particle transport; ITB formation and cross-cuts such as edge-core coupling, interaction of energetic particles with turbulence and neoclassical tearing mode trigger dynamics. Code development focuses on major initiatives in the development of full-f formulations and the capacity to simulate flux-driven transport. In addition to the full-f -formulation, the project includes the development of numerical collision models and methods for coarse graining in phase space. Verification is pursued by linear stability study comparisons with the FULL and HD7 codes and by benchmarking with the GKV, GYSELA and other gyrokinetic simulation codes. Validation of gyrokinetic models of ion and electron thermal transport is pursed by systematic stressing comparisons with fluctuation and transport data from the DIII-D and NSTX tokamaks. The physics and code development research programs are supported by complementary efforts in computer sciences, high performance computing, and data management.« less

Development of FullWave : Hot Plasma RF Simulation Tool

NASA Astrophysics Data System (ADS)

Svidzinski, Vladimir; Kim, Jin-Soo; Spencer, J. Andrew; Zhao, Liangji; Galkin, Sergei

2017-10-01

Full wave simulation tool, modeling RF fields in hot inhomogeneous magnetized plasma, is being developed. The wave equations with linearized hot plasma dielectric response are solved in configuration space on adaptive cloud of computational points. The nonlocal hot plasma dielectric response is formulated in configuration space without limiting approximations by calculating the plasma conductivity kernel based on the solution of the linearized Vlasov equation in inhomogeneous magnetic field. This approach allows for better resolution of plasma resonances, antenna structures and complex boundaries. The formulation of FullWave and preliminary results will be presented: construction of the finite differences for approximation of derivatives on adaptive cloud of computational points; model and results of nonlocal conductivity kernel calculation in tokamak geometry; results of 2-D full wave simulations in the cold plasma model in tokamak geometry using the formulated approach; results of self-consistent calculations of hot plasma dielectric response and RF fields in 1-D mirror magnetic field; preliminary results of self-consistent simulations of 2-D RF fields in tokamak using the calculated hot plasma conductivity kernel; development of iterative solver for wave equations. Work is supported by the U.S. DOE SBIR program.

Overview of Current Drive Experiment-Upgrade (CDX-U)

NASA Astrophysics Data System (ADS)

Hwang, Y. S.; Choe, W.; Stutman, D.; Lo, E.; Menard, J.; Ono, M.; Jones, T. G.; Armstrong, R.

1996-11-01

The CDX-U tokamak is a spherical tokamak (ST) facility with R ≈ 32 cm, R/a >= 1.4, and B_TF ≈ 1 kG. With an OH power supply of 60 mV-S capability, experiments were conducted with Ip up to ~ 100 kA and q(a) >= 3.5. The ST plasma performance has been studied along with various MHD-related activities. By appropriate discharge programing, it was possible to obtain MHD-quiescent discharges with a factor of 2 - 3 improvement in the electron energy confinement. Recently, the outer vacuum vessel was replaced with a toroidally continuous stainless steel chamber to accomodate the fast wave antenna. With the newly installed antenna, preliminary heating experiments using high harmonic fast waves have been pursued. The success of fast wave heating is a crucial element for achieving high beta plasmas in ST devices such as NSTX. Also, preliminary electron ripple injection (ERI) experiments were performed in CDX-U to examine the feasibility of this technique for improving ST tokamak confinement. To support the ST physics investigation, various novel plasma profile diagnostics such as the multi-pass Thomson scattering, soft x-ray tomography, and tangential-phase-contrast-imaging systems are under development on CDX-U.

Design of the 2D electron cyclotron emission imaging instrument for the J-TEXT tokamak.

PubMed

Pan, X M; Yang, Z J; Ma, X D; Zhu, Y L; Luhmann, N C; Domier, C W; Ruan, B W; Zhuang, G

2016-11-01

A new 2D Electron Cyclotron Emission Imaging (ECEI) diagnostic is being developed for the J-TEXT tokamak. It will provide the 2D electron temperature information with high spatial, temporal, and temperature resolution. The new ECEI instrument is being designed to support fundamental physics investigations on J-TEXT including MHD, disruption prediction, and energy transport. The diagnostic contains two dual dipole antenna arrays corresponding to F band (90-140 GHz) and W band (75-110 GHz), respectively, and comprises a total of 256 channels. The system can observe the same magnetic surface at both the high field side and low field side simultaneously. An advanced optical system has been designed which permits the two arrays to focus on a wide continuous region or two radially separate regions with high imaging spatial resolution. It also incorporates excellent field curvature correction with field curvature adjustment lenses. An overview of the diagnostic and the technical progress including the new remote control technique are presented.

Study of energetic particle physics with advanced ECEI system on the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Shi, Zhongbing; Jiang, Min; Yu, Liming; Chen, Wei; Shi, Peiwan; Zhong, Wulyu; Yang, Zengchen; Zhang, Boyu; Ji, Xiaoquan; Li, Yonggao; Zhou, Yan; Song, Shaodong; Huang, Mei; Song, Xianming; Li, Jiaxuan; Yuan, Baoshan; Fu, Bingzhong; Liu, Zetian; Ding, Xuantong; Xu, Yuhong; Yang, Qingwei; Duan, Xuru

2017-07-01

Understanding the physics of energetic particles (EP) is crucial for the burning plasmas in next generation fusion devices such as ITER. In this work, three types of internal kink modes (a saturated internal kink mode (SK), a resonant internal kink mode (RK), and a double e-fishbone) excited by energetic particles in the low density discharges during ECRH/ECCD heating have been studied by the newly developed 24(poloidal) × 16(radial) = 384 channel ECEI system on the HL-2A tokamak. The SK and RK rotate in the electron diamagnetic direction poloidally and are destabilized by the energetic trapped electrons. The SK is destabilized in the case of qmin > 1, while the RK is destabilized in the case of qmin < 1. The double e-fishbone, which has two m/n = 1/1 modes propagating in the opposite directions poloidally, has been observed during plasma current ramp-up with counter-ECCD. Strong thermal transfer and mode coupling between the two m/n = 1/1 modes have been studied.

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

NASA Astrophysics Data System (ADS)

Bo, SHI; Jinhong, YANG; Cheng, YANG; Desheng, CHENG; Hui, WANG; Hui, ZHANG; Haifei, DENG; Junli, QI; Xianzu, GONG; Weihua, WANG

2018-07-01

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak. The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916. The real-time feed-back control system for the plasma displacement was employed. Modeling results of the evolution of the poloidal field coil currents, the plasma current, the major radius, the plasma configuration all show agreement with experimental measurements. Results from the simulation show that during disruption, heat flux about 8 MW m‑2 flows to the upper divertor target plate and about 6 MW m‑2 flows to the lower divertor target plate. Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient. This shows that TSC has high flexibility and predictability.

Coating materials for fusion application in China

NASA Astrophysics Data System (ADS)

Luo, G.-N.; Li, Q.; Liu, M.; Zheng, X. B.; Chen, J. L.; Guo, Q. G.; Liu, X.

2011-10-01

Thick SiC coatings of ˜100 μm on graphite tiles, prepared by chemical vapor infiltration of Si into the tiles and the following reactions between Si and C, are used as plasma facing material (PFM) on HT-7 superconducting tokamak and Experimental Advanced Superconducting Tokamak (EAST). With increase in the heating and driving power in EAST, the present plasma facing component (PFC) of the SiC/C tiles bolted to heat sink will be replaced by W coatings on actively cooled Cu heat sink, prepared by vacuum plasma spraying (VPS) adopting different interlayer. The VPS-W/Cu PFC with built-in cooling channels were prepared and mounted into the HT-7 acting as a movable limiter. Behavior of heat load onto the limiter and the material was studied. The Cu coatings on the Inconel 625 tubes were successfully prepared by high velocity air-fuel (HVAF) thermal spraying, being used as the liquid nitrogen (LN2) shields of the in-vessel cryopump for divertor pumping in EAST.

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

PubMed

Xu, X Q

2008-07-01

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

Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST

NASA Astrophysics Data System (ADS)

Xu, X. Q.

2008-07-01

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

Distinct turbulence sources and confinement features in the spherical tokamak plasma regime

DOE PAGES

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

2015-10-30

New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong E x B shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offeringmore » one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. In conclusion, this predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.« less

First measurements of Hiro currents in vertical displacement event in tokamaks

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

Xiong, Hao; Xu, Guosheng; Wang, Huiqian

Specially designed tiles were setup in the 2012 campaign of the Experimental Advanced Superconducting Tokamak (EAST), to directly measure the toroidal surface currents during the disruptions. Hiro currents with direction opposite to the plasma currents have been observed, confirming the sign prediction by the Wall Touching Vertical Mode (WTVM) theory and numerical simulations. During the initial phase of the disruption, when the plasma begins to touch the wall, the surface currents can be excited by WTVM along the plasma facing tile surface, varying with the mode magnitude. The currents are not observed in the cases when the plasma moves awaymore » from the tile surface. This discovery addresses the importance of the plasma motion into the wall in vertical disruptions. WTVM, acting as a current generator, forces the Hiro currents to flow through the gaps between tiles. This effect, being overlooked so far in disruption analysis, may damage the edges of the tiles and is important for the ITER device.« less

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

Zakharov, Leonic E.; Li, Xujing

This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L.E. Zakharov [Plasma Science and Technology, accepted, ID:2013-257 (2013)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electricmore » conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.« less

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

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

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

2013-12-03

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

ALCBEAM - Neutral beam formation and propagation code for beam-based plasma diagnostics

NASA Astrophysics Data System (ADS)

Bespamyatnov, I. O.; Rowan, W. L.; Liao, K. T.

2012-03-01

ALCBEAM is a new three-dimensional neutral beam formation and propagation code. It was developed to support the beam-based diagnostics installed on the Alcator C-Mod tokamak. The purpose of the code is to provide reliable estimates of the local beam equilibrium parameters: such as beam energy fractions, density profiles and excitation populations. The code effectively unifies the ion beam formation, extraction and neutralization processes with beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. This paper describes the physical processes interpreted and utilized by the code, along with exploited computational methods. The description is concluded by an example simulation of beam penetration into plasma of Alcator C-Mod. The code is successfully being used in Alcator C-Mod tokamak and expected to be valuable in the support of beam-based diagnostics in most other tokamak environments. Program summaryProgram title: ALCBEAM Catalogue identifier: AEKU_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKU_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 66 459 No. of bytes in distributed program, including test data, etc.: 7 841 051 Distribution format: tar.gz Programming language: IDL Computer: Workstation, PC Operating system: Linux RAM: 1 GB Classification: 19.2 Nature of problem: Neutral beams are commonly used to heat and/or diagnose high-temperature magnetically-confined laboratory plasmas. An accurate neutral beam characterization is required for beam-based measurements of plasma properties. Beam parameters such as density distribution, energy composition, and atomic excited populations of the beam atoms need to be known. Solution method: A neutral beam is initially formed as an ion beam which is extracted from the ion source by high voltage applied to the extraction and accelerating grids. The current distribution of a single beamlet emitted from a single pore of IOS depends on the shape of the plasma boundary in the emission region. Total beam extracted by IOS is calculated at every point of 3D mesh as sum of all contributions from each grid pore. The code effectively unifies the ion beam formation, extraction and neutralization processes with neutral beam attenuation and excitation in plasma and neutral gas and beam stopping by the beam apertures. Running time: 10 min for a standard run.

Advanced tokamak investigations in full-tungsten ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Bock, A.; Doerk, H.; Fischer, R.; Rittich, D.; Stober, J.; Burckhart, A.; Fable, E.; Geiger, B.; Mlynek, A.; Reich, M.; Zohm, H.; ASDEX Upgrade Team

2018-05-01

The appropriate tailoring of the q-profile is the key to accessing Advanced Tokamak (AT) scenarios, which are of great benefit to future all-metal fusion power plants. Such scenarios depend on low collisionality ν* which permits efficient external current drive and high amounts of intrinsic bootstrap current. At constant pressure, lowering of the electron density ne leads to a strong decrease in the collisionality with increasing electron temperature ν* ˜ Te-3 . Simultaneously, the conditions for low ne also benefit impurity accumulation. This paper reports on how radiative collapses due to central W accumulation were overcome by improved understanding of the changes to recycling and pumping, substantially expanded ECRH capacities for both heating and current drive, and a new solid W divertor capable of withstanding the power loads at low ne. Furthermore, it reports on various improvements to the reliability of the q-profile reconstruction. A candidate steady state scenario for ITER/DEMO (q95 = 5.3, βN = 2.7, fbs > 40%) is presented. The ion temperature profiles are steeper than predicted by TGLF, but nonlinear electromagnetic gyro-kinetic analyses with GENE including fast particle effects matched the experimental heat fluxes. A fully non-inductive scenario at higher q95 = 7.1 for current drive model validation is also discussed. The results show that non-inductive operation is principally compatible with full-metal machines.

In-situ Testing of the EHT High Gain and Frequency Ultra-Stable Integrators

NASA Astrophysics Data System (ADS)

Miller, Kenneth; Ziemba, Timothy; Prager, James; Slobodov, Ilia; Lotz, Dan

2014-10-01

Eagle Harbor Technologies (EHT) has developed a long-pulse integrator that exceeds the ITER specification for integration error and pulse duration. During the Phase I program, EHT improved the RPPL short-pulse integrators, added a fast digital reset, and demonstrated that the new integrators exceed the ITER integration error and pulse duration requirements. In Phase II, EHT developed Field Programmable Gate Array (FPGA) software that allows for integrator control and real-time signal digitization and processing. In the second year of Phase II, the EHT integrator will be tested at a validation platform experiment (HIT-SI) and tokamak (DIII-D). In the Phase IIB program, EHT will continue development of the EHT integrator to reduce overall cost per channel. EHT will test lower cost components, move to surface mount components, and add an onboard Field Programmable Gate Array and data acquisition to produce a stand-alone system with lower cost per channel and increased the channel density. EHT will test the Phase IIB integrator at a validation platform experiment (HIT-SI) and tokamak (DIII-D). Work supported by the DOE under Contract Number (DE-SC0006281).

Effects of Density and Impurity on Edge Localized Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Zhu, Ping

2017-10-01

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

Numerical investigation of non-perturbative kinetic effects of energetic particles on toroidicity-induced Alfvén eigenmodes in tokamaks and stellarators

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

Slaby, Christoph; Könies, Axel; Kleiber, Ralf

2016-09-15

The resonant interaction of shear Alfvén waves with energetic particles is investigated numerically in tokamak and stellarator geometry using a non-perturbative MHD-kinetic hybrid approach. The focus lies on toroidicity-induced Alfvén eigenmodes (TAEs), which are most easily destabilized by a fast-particle population in fusion plasmas. While the background plasma is treated within the framework of an ideal-MHD theory, the drive of the fast particles, as well as Landau damping of the background plasma, is modelled using the drift-kinetic Vlasov equation without collisions. Building on analytical theory, a fast numerical tool, STAE-K, has been developed to solve the resulting eigenvalue problem usingmore » a Riccati shooting method. The code, which can be used for parameter scans, is applied to tokamaks and the stellarator Wendelstein 7-X. High energetic-ion pressure leads to large growth rates of the TAEs and to their conversion into kinetically modified TAEs and kinetic Alfvén waves via continuum interaction. To better understand the physics of this conversion mechanism, the connections between TAEs and the shear Alfvén wave continuum are examined. It is shown that, when energetic particles are present, the continuum deforms substantially and the TAE frequency can leave the continuum gap. The interaction of the TAE with the continuum leads to singularities in the eigenfunctions. To further advance the physical model and also to eliminate the MHD continuum together with the singularities in the eigenfunctions, a fourth-order term connected to radiative damping has been included. The radiative damping term is connected to non-ideal effects of the bulk plasma and introduces higher-order derivatives to the model. Thus, it has the potential to substantially change the nature of the solution. For the first time, the fast-particle drive, Landau damping, continuum damping, and radiative damping have been modelled together in tokamak- as well as in stellarator geometry.« less

Advanced divertor configurations with large flux expansion

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; McLean, A.; Menard, J. E.; Paul, S. F.; Podesta, M.; Raman, R.; Ryutov, D. D.; Scotti, F.; Kaita, R.; Maingi, R.; Mueller, D. M.; Roquemore, A. L.; Reimerdes, H.; Canal, G. P.; Labit, B.; Vijvers, W.; Coda, S.; Duval, B. P.; Morgan, T.; Zielinski, J.; De Temmerman, G.; Tal, B.

2013-07-01

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3-7 MW/m2 to 0.5-1 MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L-H power threshold, enhanced stability of the peeling-ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2-3) Type I ELM frequency and slightly increased (20-30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma-material interface issues.

MHD Effects of a Ferritic Wall on Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Hughes, Paul E.

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

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 collisionality regime, is predicted.

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 advanced ST collisionality regime, is predicted.« less

Ohmic ignition with high engineering beta based on the RFP

NASA Astrophysics Data System (ADS)

Sarff, J. S.; Anderson, J. K.; Chapman, B. E.; McCollam, K. J.

2017-10-01

The RFP configuration allows the possibility of ohmic ignition for fusion energy, eliminating the need for auxiliary heating by rf or neutral beam injection. Complex plasma-facing antennas and NBI sources are therefore not required, simplifying the difficult fusion materials challenge. While all toroidal configurations require a volume-average 〈 B 〉 >= 5 T, the field strength at the magnet in the RFP is only Bcoil 3T since plasma current generates almost all of the field. Engineering beta is therefore maximized. We summarize access to ohmic ignition by examining a Lawson-like power balance for an RFP fusion plasma comparable to the ARIES-AT advanced tokamak, which generates neutron wall loading Pn / A 5 MW/m2. The required energy confinement for ohmic ignition in an RFP is similar to that for a tokamak. Confinement in MST is comparable to a same-size, same-field tokamak plasma, but 〈 B 〉 in MST is only 1/20th that required for fusion. While transport could ultimately be dominated by micro turbulence, extrapolation of stochastic transport using Lundquist number scaling for MHD tearing indicates standard RFP confinement (not enhanced by current profile control) could be sufficient to access ohmic ignition. This bolsters the possibility for steady-state inductive sustainment using oscillating field current drive. The high beta and classical energetic ion confinement measured in MST also bolster the RFP's fusion potential. Work supported by U.S. DoE.

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.

Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

NASA Astrophysics Data System (ADS)

Meyer, H.; Eich, T.; Beurskens, M.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.; Allan, S.; Alves, D.; Ambrosino, R.; Amicucci, L.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Apruzzese, G.; Ariola, M.; Arnichand, H.; Arter, W.; Baciero, A.; Barnes, M.; Barrera, L.; Behn, R.; Bencze, A.; Bernardo, J.; Bernert, M.; Bettini, P.; Bilková, P.; Bin, W.; Birkenmeier, G.; Bizarro, J. P. S.; Blanchard, P.; Blanken, T.; Bluteau, M.; Bobkov, V.; Bogar, O.; Böhm, P.; Bolzonella, T.; Boncagni, L.; Botrugno, A.; Bottereau, C.; Bouquey, F.; Bourdelle, C.; Brémond, S.; Brezinsek, S.; Brida, D.; Brochard, F.; Buchanan, J.; Bufferand, H.; Buratti, P.; Cahyna, P.; Calabrò, G.; Camenen, Y.; Caniello, R.; Cannas, B.; Canton, A.; Cardinali, A.; Carnevale, D.; Carr, M.; Carralero, D.; Carvalho, P.; Casali, L.; Castaldo, C.; Castejón, F.; Castro, R.; Causa, F.; Cavazzana, R.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cesario, R.; Challis, C. D.; Chapman, I. T.; Chapman, S.; Chernyshova, M.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Clairet, F.; Classen, I.; Coelho, R.; Coenen, J. W.; Colas, L.; Conway, G.; Corre, Y.; Costea, S.; Crisanti, F.; Cruz, N.; Cseh, G.; Czarnecka, A.; D'Arcangelo, O.; De Angeli, M.; De Masi, G.; De Temmerman, G.; De Tommasi, G.; Decker, J.; Delogu, R. S.; Dendy, R.; Denner, P.; Di Troia, C.; Dimitrova, M.; D'Inca, R.; Dorić, V.; Douai, D.; Drenik, A.; Dudson, B.; Dunai, D.; Dunne, M.; Duval, B. P.; Easy, L.; Elmore, S.; Erdös, B.; Esposito, B.; Fable, E.; Faitsch, M.; Fanni, A.; Fedorczak, N.; Felici, F.; Ferreira, J.; Février, O.; Ficker, O.; Fietz, S.; Figini, L.; Figueiredo, A.; Fil, A.; Fishpool, G.; Fitzgerald, M.; Fontana, M.; Ford, O.; Frassinetti, L.; Fridström, R.; Frigione, D.; Fuchert, G.; Fuchs, C.; Furno Palumbo, M.; Futatani, S.; Gabellieri, L.; Gałązka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Gao, Y.; Garavaglia, S.; Garcia, J.; Garcia-Carrasco, A.; Garcia-Lopez, J.; Garcia-Munoz, M.; Gardarein, J.-L.; Garzotti, L.; Gaspar, J.; Gauthier, E.; Geelen, P.; Geiger, B.; Ghendrih, P.; Ghezzi, F.; Giacomelli, L.; Giannone, L.; Giovannozzi, E.; Giroud, C.; Gleason González, C.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Gruber, M.; Gude, A.; Guimarais, L.; Guirlet, R.; Gunn, J.; Hacek, P.; Hacquin, S.; Hall, S.; Ham, C.; Happel, T.; Harrison, J.; Harting, D.; Hauer, V.; Havlickova, E.; Hellsten, T.; Helou, W.; Henderson, S.; Hennequin, P.; Heyn, M.; Hnat, B.; Hölzl, M.; Hogeweij, D.; Honoré, C.; Hopf, C.; Horáček, J.; Hornung, G.; Horváth, L.; Huang, Z.; Huber, A.; Igitkhanov, J.; Igochine, V.; Imrisek, M.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A. S.; Jacquet, P.; Jakubowski, M.; Jardin, A.; Jaulmes, F.; Jenko, F.; Jensen, T.; Jeppe Miki Busk, O.; Jessen, M.; Joffrin, E.; Jones, O.; Jonsson, T.; Kallenbach, A.; Kallinikos, N.; Kálvin, S.; Kappatou, A.; Karhunen, J.; Karpushov, A.; Kasilov, S.; Kasprowicz, G.; Kendl, A.; Kernbichler, W.; Kim, D.; Kirk, A.; Kjer, S.; Klimek, I.; Kocsis, G.; Kogut, D.; Komm, M.; Korsholm, S. B.; Koslowski, H. R.; Koubiti, M.; Kovacic, J.; Kovarik, K.; Krawczyk, N.; Krbec, J.; Krieger, K.; Krivska, A.; Kube, R.; Kudlacek, O.; Kurki-Suonio, T.; Labit, B.; Laggner, F. M.; Laguardia, L.; Lahtinen, A.; Lalousis, P.; Lang, P.; Lauber, P.; Lazányi, N.; Lazaros, A.; Le, H. B.; Lebschy, A.; Leddy, J.; Lefévre, L.; Lehnen, M.; Leipold, F.; Lessig, A.; Leyland, M.; Li, L.; Liang, Y.; Lipschultz, B.; Liu, Y. Q.; Loarer, T.; Loarte, A.; Loewenhoff, T.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Lupelli, I.; Lux, H.; Lyssoivan, A.; Madsen, J.; Maget, P.; Maggi, C.; Maggiora, R.; Magnussen, M. L.; Mailloux, J.; Maljaars, B.; Malygin, A.; Mantica, P.; Mantsinen, M.; Maraschek, M.; Marchand, B.; Marconato, N.; Marini, C.; Marinucci, M.; Markovic, T.; Marocco, D.; Marrelli, L.; Martin, Y.; Solis, J. R. Martin; Martitsch, A.; Mastrostefano, S.; Mattei, M.; Matthews, G.; Mavridis, M.; Mayoral, M.-L.; Mazon, D.; McCarthy, P.; McAdams, R.; McArdle, G.; McCarthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meisl, G.; Merle, A.; Meyer, O.; Milanesio, D.; Militello, F.; Miron, I. G.; Mitosinkova, K.; Mlynar, J.; Mlynek, A.; Molina, D.; Molina, P.; Monakhov, I.; Morales, J.; Moreau, D.; Morel, P.; Moret, J.-M.; Moro, A.; Moulton, D.; Müller, H. W.; Nabais, F.; Nardon, E.; Naulin, V.; Nemes-Czopf, A.; Nespoli, F.; Neu, R.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nimb, S.; Nocente, M.; Nouailletas, R.; Nowak, S.; Oberkofler, M.; Oberparleiter, M.; Ochoukov, R.; Odstrčil, T.; Olsen, J.; Omotani, J.; O'Mullane, M. G.; Orain, F.; Osterman, N.; Paccagnella, R.; Pamela, S.; Pangione, L.; Panjan, M.; Papp, G.; Papřok, R.; Parail, V.; Parra, F. I.; Pau, A.; Pautasso, G.; Pehkonen, S.-P.; Pereira, A.; Perelli Cippo, E.; Pericoli Ridolfini, V.; Peterka, M.; Petersson, P.; Petrzilka, V.; Piovesan, P.; Piron, C.; Pironti, A.; Pisano, F.; Pisokas, T.; Pitts, R.; Ploumistakis, I.; Plyusnin, V.; Pokol, G.; Poljak, D.; Pölöskei, P.; Popovic, Z.; Pór, G.; Porte, L.; Potzel, S.; Predebon, I.; Preynas, M.; Primc, G.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rack, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravera, G.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reinke, M.; Reiser, D.; Resnik, M.; Reux, C.; Ripamonti, D.; Rittich, D.; Riva, G.; Rodriguez-Ramos, M.; Rohde, V.; Rosato, J.; Ryter, F.; Saarelma, S.; Sabot, R.; Saint-Laurent, F.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, M.; Schneider, B.; Schneider, P.; Schneller, M.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Seidl, J.; Sertoli, M.; Šesnić, S.; Shabbir, A.; Shalpegin, A.; Shanahan, B.; Sharapov, S.; Sheikh, U.; Sias, G.; Sieglin, B.; Silva, C.; Silva, A.; Silva Fuglister, M.; Simpson, J.; Snicker, A.; Sommariva, C.; Sozzi, C.; Spagnolo, S.; Spizzo, G.; Spolaore, M.; Stange, T.; Stejner Pedersen, M.; Stepanov, I.; Stober, J.; Strand, P.; Šušnjara, A.; Suttrop, W.; Szepesi, T.; Tál, B.; Tala, T.; Tamain, P.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thornton, A.; Tolias, P.; Tophøj, L.; Treutterer, W.; Trevisan, G. L.; Tripsky, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Uccello, A.; Urban, J.; Valisa, M.; Vallejos, P.; Valovic, M.; Van den Brand, H.; Vanovac, B.; Varoutis, S.; Vartanian, S.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Vignitchouk, L.; Vijvers, W. A. J.; Villone, F.; Viola, B.; Vlahos, L.; Voitsekhovitch, I.; Vondráček, P.; Vu, N. M. T.; Wagner, D.; Walkden, N.; Wang, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Westerhof, E.; Wiesenberger, M.; Willensdorfer, M.; Wischmeier, M.; Wodniak, I.; Wolfrum, E.; Yadykin, D.; Zagórski, R.; Zammuto, I.; Zanca, P.; Zaplotnik, R.; Zestanakis, P.; Zhang, W.; Zoletnik, S.; Zuin, M.; ASDEX Upgrade, the; MAST; TCV Teams

2017-10-01

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n  =  2 RMP maintaining good confinement {{H}\\text{H≤ft(98,\\text{y}2\\right)}}≈ 0.95 . Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes. In the future we will refer to the author list of the paper as the EUROfusion MST1 Team.

ELM mitigation techniques

NASA Astrophysics Data System (ADS)

Evans, T. E.

2013-07-01

Large edge-localized mode (ELM) control techniques must be developed to help ensure the success of burning and ignited fusion plasma devices such as tokamaks and stellarators. In full performance ITER tokamak discharges, with QDT = 10, the energy released by a single ELM could reach ˜30 MJ which is expected to result in an energy density of 10-15 MJ/m2on the divertor targets. This will exceed the estimated divertor ablation limit by a factor of 20-30. A worldwide research program is underway to develop various types of ELM control techniques in preparation for ITER H-mode plasma operations. An overview of the ELM control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER. Particular emphasis is given to the primary approaches, pellet pacing and resonant magnetic perturbation fields, currently being considered for ITER.

Area of Stochastic Scrape-Off Layer for a Single-Null Divertor Tokamak Using Simple Map

NASA Astrophysics Data System (ADS)

Fisher, Tiffany; Verma, Arun; Punjabi, Alkesh

1996-11-01

The magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The Simple map is characterized by a single parameter k representing the toroidal asymmetry. The width of the stochastic scrape-off layer and its area varies with the map parameter k. We calculate the area of the stochastic scrape-off layer for different k's and obtain a parametric expression for the area in terms of k and y _LastGoodSurface(k). This work is supported by US DOE OFES. Tiffany Fisher is a HU CFRT Summer Fusion High school Workshop Scholar from New Bern High School in North Carolina. She is supported by NASA SHARP Plus Program.

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 where snakes are present—various physics models of the sawteeth are now being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER.

Experimental results from an X-ray imaging crystal spectrometer utilizing multi-wire proportional counter for KSTAR

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

Lee, S. G., E-mail: sglee@nfri.re.kr; Kim, Y. S.; Yoo, J. W.

2016-11-15

The inconsistency of the first experimental results from the X-ray imaging crystal spectrometer for the Korea Superconducting Tokamak Advanced Research device utilizing a multi-wire proportional counter (MWPC) is clarified after improving the photon-count rate of the data acquisition system for the MWPC and ground loop isolator for the whole spectrometer system. The improved MWPC is successfully applied to pure Ohmic plasmas as well as plasmas with high confinement modes.

Enabling High Fidelity Measurements of Energy and Pitch Angle for Escaping Energetic Ions with a Fast Ion Loss Detector

NASA Astrophysics Data System (ADS)

Chaban, R.; Pace, D. C.; Marcy, G. R.; Taussig, D.

2016-10-01

Energetic ion losses must be minimized in burning plasmas to maintain fusion power, and existing tokamaks provide access to energetic ion parameter regimes that are relevant to burning machines. A new Fast Ion Loss Detector (FILD) probe on the DIII-D tokamak has been optimized to resolve beam ion losses across a range of 30 - 90 keV in energy and 40° to 80° in pitch angle, thereby providing valuable measurements during many different experiments. The FILD is a magnetic spectrometer; once inserted into the tokamak, the magnetic field allows energetic ions to pass through a collimating aperture and strike a scintillator plate that is imaged by a wide view camera and narrow view photomultiplier tubes (PMTs). The design involves calculating scintillator strike patterns while varying probe geometry. Calculated scintillator patterns are then used to design an optical system that allows adjustment of the focus regions for the 1 MS/s resolved PMTs. A synthetic diagnostic will be used to determine the energy and pitch angle resolution that can be attained in DIII-D experiments. Work supported in part by US DOE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698.

Studies of runaway electrons via Cherenkov effect in tokamaks

NASA Astrophysics Data System (ADS)

Zebrowski, J.; Jakubowski, L.; Rabinski, M.; Sadowski, M. J.; Jakubowski, M. J.; Kwiatkowski, R.; Malinowski, K.; Mirowski, R.; Mlynar, J.; Ficker, O.; Weinzettl, V.; Causa, F.; COMPASS; FTU Teams

2018-01-01

The paper concerns measurements of runaway electrons (REs) which are generated during discharges in tokamaks. The control of REs is an important task in experimental studies within the ITER-physics program. The NCBJ team proposed to study REs by means of Cherenkov-type detectors several years ago. The Cherenkov radiation, induced by REs in appropriate radiators, makes it possible to identify fast electron beams and to determine their spatial- and temporal-characteristics. The results of recent experimental studies of REs, performed in two tokamaks - COMPASS in Prague and FTU in Frascati, are summarized and discussed in this paper. Examples of the electron-induced signals, as recorded at different experimental conditions and scenarios, are presented. Measurements performed with a three-channel Cherenkov-probe in COMPASS showed that the first fast electron peaks can be observed already during the current ramp-up phase. A strong dependence of RE-signals on the radial position of the Cherenkov probe was observed. The most distinct electron peaks were recorded during the plasma disruption. The Cherenkov signals confirmed the appearance of post-disruptive RE beams in circular-plasma discharges with massive Ar-puffing. During experiments at FTU a clear correlation between the Cherenkov detector signals and the rotation of magnetic islands was identified.

The Research of EAST Pedestal Structure and Preliminary Application

NASA Astrophysics Data System (ADS)

Wang, Tengfei; Zang, Qing; Han, Xiaofeng; Xiao, Shumei; Hu, Ailan; Zhao, Junyu

2016-10-01

The pedestal characteristic is an important basis for high confinement mode (H-mode) research. Because of the finite spatial resolution of Thomson scattering (TS) diagnostic on Experimental Advanced Superconducting Tokamak (EAST), it is necessary to characterize the pedestal with a suitable functional form. Based on simulated and experimental data of EAST, it is shown that the two-line method with a bilinear fitting has better reproducibility of pedestal parameters than hyperbolic tangent (tanh) and modified hyperbolic tangent (mtanh) methods. This method has been applied to EAST type I edge localized mode (ELM) discharges, and the electron pedestal density is found to be proportional to the line-averaged density and the edge pressure gradient is found to be proportional to the pedestal pressure. Furthermore, the ion poloidal gyro-radius has been identified as the suitable parameter to describe the pedestal pressure width. supported by National Natural Science Foundation of China (Nos. 11275233 and 11405206), and the National Magnetic Confinement Fusion Science Program of China (No. 2013GB112003), and Science Foundation of Institute of Plasma Physics, Chinese Academy of Sciences (No. DSJJ-15-JC01)

High Harmonic Fast Wave Damping on an Ion Beam: NSTX and DIII-D Regimes Compared

NASA Astrophysics Data System (ADS)

Pinsker, R. I.; Choi, C. C.; Petty, C. C.; Porkolab, M.; Wilson, J. R.; Murakami, M.; Harvey, R. W.

2004-11-01

Both NSTX and DIII-D use the combination of fast Alfven waves (FW) and neutral beam injection (NBI) for central electron heating and current drive. Damping of the fast wave on the beam ions at moderate to high harmonics (4th--20th) of the beam ion cyclotron frequency represents a loss process. In DIII-D current drive experiments at low density in which 4th and 8th harmonics were compared, damping at the 8th harmonic damping was much weaker than at the 4th [1]. However, recent simulations have predicted that in higher density and higher beam power regimes (of interest to the Advanced Tokamak program) the beam ion absorption will transition to the unmagnetized ion regime, where the damping is significant and essentially independent of harmonic number. In the present work, the transition from magnetized to unmagnetized ion regimes for the NSTX and DIII-D HHFW experiments is studied theoretically, with a combination of simple semi-analytic models and numerical models. \\vspace0.25 em [1] C.C. Petty, et al., Plasma Phys. and Contr. Fusion 43, 1747 (2001).

Toroidal Alfvénic Eigenmodes Driven by Energetic Particles with Maxwell and Slowing-down Distributions

NASA Astrophysics Data System (ADS)

Hou, Yawei; Zhu, Ping; Zou, Zhihui; Kim, Charlson C.; Hu, Zhaoqing; Wang, Zhengxiong

2016-10-01

The energetic-particle (EP) driven toroidal Alfvén eigenmodes (TAEs) in a circular-shaped large aspect ratio tokamak are studied using the hybrid kinetic-MHD model in the NIMROD code, where the EPs are advanced using the δf particle-in-cell (PIC) method and their kinetic effects are coupled to the bulk plasma through moment closures. Two initial distributions of EPs, Maxwell and slowing-down, are considered. The influence of EP parameters, including density, temperature and density gradient, on the frequency and the growth rate of TAEs are obtained and benchmarked with theory and gyrokinetic simulations for the Maxwell distribution with good agreement. When the density and temperature of EPs are above certain thresholds, the transition from TAE to energetic particle modes (EPM) occurs and the mode structure also changes. Comparisons between Maxwell and slowing-down distributions in terms of EP-driven TAEs and EPMs will also be presented and discussed. Supported by the National Magnetic Confinement Fusion Science Program of China Grant Nos. 2014GB124002 and 2015GB101004, and the Natural Science Foundation of China Grant No. 11205194.

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.

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.

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

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

NASA Astrophysics Data System (ADS)

Moreau, D.; Artaud, J. F.; Ferron, J. R.; Holcomb, C. T.; Humphreys, D. A.; Liu, F.; Luce, T. C.; Park, J. M.; Prater, R.; Turco, F.; Walker, M. L.

2015-06-01

This paper shows that semi-empirical data-driven models based on a two-time-scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, βN, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off-axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated open-loop data obtained using a rapidly converging plasma transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0D scaling laws and 1.5D ordinary differential equations. The paper discusses the results of closed-loop METIS simulations, using the near-optimal ARTAEMIS control algorithm (Moreau D et al 2013 Nucl. Fusion 53 063020) for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and βN are satisfactorily tracked with a time scale of about 10 s, despite large disturbances applied to the feedforward powers and plasma parameters. The robustness of the control algorithm with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.

The accurate particle tracer code

NASA Astrophysics Data System (ADS)

Wang, Yulei; Liu, Jian; Qin, Hong; Yu, Zhi; Yao, Yicun

2017-11-01

The Accurate Particle Tracer (APT) code is designed for systematic large-scale applications of geometric algorithms for particle dynamical simulations. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and nonlinear problems. To provide a flexible and convenient I/O interface, the libraries of Lua and Hdf5 are used. Following a three-step procedure, users can efficiently extend the libraries of electromagnetic configurations, external non-electromagnetic forces, particle pushers, and initialization approaches by use of the extendible module. APT has been used in simulations of key physical problems, such as runaway electrons in tokamaks and energetic particles in Van Allen belt. As an important realization, the APT-SW version has been successfully distributed on the world's fastest computer, the Sunway TaihuLight supercomputer, by supporting master-slave architecture of Sunway many-core processors. Based on large-scale simulations of a runaway beam under parameters of the ITER tokamak, it is revealed that the magnetic ripple field can disperse the pitch-angle distribution significantly and improve the confinement of energetic runaway beam on the same time.

Combined Langmuir-magnetic probe measurements of type-I ELMy filaments in the EAST tokamak

NASA Astrophysics Data System (ADS)

Qingquan, YANG; Fangchuan, ZHONG; Guosheng, XU; Ning, YAN; Liang, CHEN; Xiang, LIU; Yong, LIU; Liang, WANG; Zhendong, YANG; Yifeng, WANG; Yang, YE; Heng, ZHANG; Xiaoliang, Li

2018-06-01

Detailed investigations on the filamentary structures associated with the type-I edge-localized modes (ELMs) should be helpful for protecting the materials of a plasma-facing wall on a future large device. Related experiments have been carefully conducted in the Experimental Advanced Superconducting Tokamak (EAST) using combined Langmuir-magnetic probes. The experimental results indicate that the radially outward velocity of type-I ELMy filaments can be up to 1.7 km s‑1 in the far scrape-off layer (SOL) region. It is remarkable that the electron temperature of these filaments is detected to be ∼50 eV, corresponding to a fraction of 1/6 to the temperature near the pedestal top, while the density (∼ 3× {10}19 {{{m}}}-3) of these filaments could be approximate to the line-averaged density. In addition, associated magnetic fluctuations have been clearly observed at the same time, which show good agreement with the density perturbations. A localized current on the order of ∼100 kA could be estimated within the filaments.

ATCA digital controller hardware for vertical stabilization of plasmas in tokamaks

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

Batista, A. J. N.; Sousa, J.; Varandas, C. A. F.

2006-10-15

The efficient vertical stabilization (VS) of plasmas in tokamaks requires a fast reaction of the VS controller, for example, after detection of edge localized modes (ELM). For controlling the effects of very large ELMs a new digital control hardware, based on the Advanced Telecommunications Computing Architecture trade mark sign (ATCA), is being developed aiming to reduce the VS digital control loop cycle (down to an optimal value of 10 {mu}s) and improve the algorithm performance. The system has 1 ATCA trade mark sign processor module and up to 12 ATCA trade mark sign control modules, each one with 32 analogmore » input channels (12 bit resolution), 4 analog output channels (12 bit resolution), and 8 digital input/output channels. The Aurora trade mark sign and PCI Express trade mark sign communication protocols will be used for data transport, between modules, with expected latencies below 2 {mu}s. Control algorithms are implemented on a ix86 based processor with 6 Gflops and on field programmable gate arrays with 80 GMACS, interconnected by serial gigabit links in a full mesh topology.« 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.

Operating characteristics of a new ion source for KSTAR neutral beam injection system.

PubMed

Kim, Tae-Seong; Jeong, Seung Ho; Chang, Doo-Hee; Lee, Kwang Won; In, Sang-Ryul

2014-02-01

A new positive ion source for the Korea Superconducting Tokamak Advanced Research neutral beam injection (KSTAR NBI-1) system was designed, fabricated, and assembled in 2011. The characteristics of the arc discharge and beam extraction were investigated using hydrogen and helium gas to find the optimum operating parameters of the arc power, filament voltage, gas pressure, extracting voltage, accelerating voltage, and decelerating voltage at the neutral beam test stand at the Korea Atomic Energy Research Institute in 2012. Based on the optimum operating condition, the new ion source was then conditioned, and performance tests were primarily finished. The accelerator system with enlarged apertures can extract a maximum 65 A ion beam with a beam energy of 100 keV. The arc efficiency and optimum beam perveance, at which the beam divergence is at a minimum, are estimated to be 1.0 A/kW and 2.5 uP, respectively. The beam extraction tests show that the design goal of delivering a 2 MW deuterium neutral beam into the KSTAR Tokamak plasma is achievable.

Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

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

Xu, X Q

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

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.

Damping Rate Measurements of Medium n Alfv'en Eigenmodes in JET

NASA Astrophysics Data System (ADS)

Klein, Alexander; Testa, Duccio; Snipes, Joseph; Fasoli, Ambrogio; Carfantan, Hervé

2007-11-01

Alfv'en Eigenmodes (AE's) with mode numbers 5 < n < 20 are expected to be unstable in burning tokamaks and may lead to loss of fast particle confinement. The active MHD spectroscopy program at JET has already provided a wealth of information about low n (n <= 2) AE's in the past decade, but a recently installed array of four antennas is capable of driving higher mode numbered (n < 100, 30 < f < 350 kHz) perturbations. In the latest JET campaign, the damping rates for several types of AE's were measured parasitically in a wide range of tokamak scenarios. We review the active MHD diagnostic and present the first measurements of medium-n AE stability on JET, then describe future plans for the active MHD spectroscopy project. The data analysis involves a novel method for resolving multiple AE's that exist at identical frequencies, which uses techniques based on the SparSpec code.

New Technique of AC drive in Tokamak using Permanent Magnets

NASA Astrophysics Data System (ADS)

Matteucci, Jackson; Zolfaghari, Ali

2013-10-01

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

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

NASA Astrophysics Data System (ADS)

Yan, Xingting; Zhu, Ping; Sun, Youwen

2016-10-01

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

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak.

PubMed

Xu, J C; Wang, L; Xu, G S; Luo, G N; Yao, D M; Li, Q; Cao, L; Chen, L; Zhang, W; Liu, S C; Wang, H Q; Jia, M N; Feng, W; Deng, G Z; Hu, L Q; Wan, B N; Li, J; Sun, Y W; Guo, H Y

2016-08-01

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

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.

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

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

Xu, J. C.; Jia, M. N.; Feng, W.

2016-08-15

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triplemore » probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.« less

Bifurcated helical core equilibrium states in tokamaks

NASA Astrophysics Data System (ADS)

Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

2013-07-01

Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

Cherenkov neutron detector for fusion reaction and runaway electron diagnostics

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

Cheon, MunSeong, E-mail: munseong@nfri.re.kr; Kim, Junghee

2015-08-15

A Cherenkov-type neutron detector was newly developed and neutron measurement experiments were performed at Korea Superconducting Tokamak Advanced Research. It was shown that the Cherenkov neutron detector can monitor the time-resolved neutron flux from deuterium-fueled fusion plasmas. Owing to the high temporal resolution of the detector, fast behaviors of runaway electrons, such as the neutron spikes, could be observed clearly. It is expected that the Cherenkov neutron detector could be utilized to provide useful information on runaway electrons as well as fusion reaction rate in fusion plasmas.

Modeling of the EAST ICRF antenna with ICANT Code

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

Qin Chengming; Zhao Yanping; Colas, L.

2007-09-28

A Resonant Double Loop (RDL) antenna for ion-cyclotron range of frequencies (ICRF) on Experimental Advanced Superconducting Tokamak (EAST) is under construction. The new antenna is analyzed using the antenna coupling code ICANT which self-consistently determines the surface currents on all antenna parts. In this work, the modeling of the new ICRF antenna using this code is to assess the near-fields in front of the antenna and analysis its coupling capabilities. Moreover, the antenna reactive radiated power computed by ICANT and shows a good agreement with deduced from Transmission Line (TL) theory.

Modeling of the EAST ICRF antenna with ICANT Code

NASA Astrophysics Data System (ADS)

Qin, Chengming; Zhao, Yanping; Colas, L.; Heuraux, S.

2007-09-01

A Resonant Double Loop (RDL) antenna for ion-cyclotron range of frequencies (ICRF) on Experimental Advanced Superconducting Tokamak (EAST) is under construction. The new antenna is analyzed using the antenna coupling code ICANT which self-consistently determines the surface currents on all antenna parts. In this work, the modeling of the new ICRF antenna using this code is to assess the near-fields in front of the antenna and analysis its coupling capabilities. Moreover, the antenna reactive radiated power computed by ICANT and shows a good agreement with deduced from Transmission Line (TL) theory.

Investigation of Plasma Surface Interactions with the PISCES ELM Laser System

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Baldwin, M.; Hanna, J.; Doerner, R.; Lynch, T.; Palmer, T.; Tynan, G. R.

2007-11-01

When an ELM occurs in tokamaks, up to 30% of the pedestal energy can be deposited on the wall of the tokamak causing heating & material loss due to sublimation, evaporation and melt splashing of plasma facing components (PFCs) and expansion of the ejected material into the plasma. We have explored heat pulses using an electrical power circuit to draw electrons from the plasma to heat samples ohmically. This system is limited in power to ˜250kJ/m^2 at the minimum pulse width of 10ms and depletes the plasma column, complicating spectroscopy. We have completed calculations that indicate that a pulsed laser system can be used to simulate the heat pulse of ELMs. We are integrating laser systems into the existing PFC research program in PISCES, a laboratory facility capable of reproducing plasma-materials interactions expected during normal operation of large tokamaks. Two Nd:YAG lasers capable of delivering up to 50J of energy over various pulsewidths are used for the experiments. Laser heat pulse only, H+/D+ plasma only, and laser+plasma experiments were conducted and initial results indicate that metals behave very differently while exposed to plasma and simultaneous heat pulses. We will also discuss initial results for carbon PFCs and material transport into the plasma. Supported by US DoE grant DE-FG02-07ER-54912.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Last Good Surface

NASA Astrophysics Data System (ADS)

Phan, Huong; Ali, Halima; Punjabi, Alkesh

2000-10-01

The Simple Map^1 is a representation of the magnetic field inside a single-null divertor tokamak. It is given by the equations: X_n+1=X_n kYn (1-Y_n), Y_n+1= Y_n+kX_n+1. These equations mimic the motion of the magnetic field lines in a single-null divertor tokamak. The fixed stable point is (0,0) and the unstable fixed oint is (0,1). k is fixed at 0.60. In our work, the starting values of Y in the map is kept in the interval of 0 to 1, and the starting value of X is 0. Using the successive bifurcation method, we first run these equations for 10^6 iterations to find the approximate value of Y when chaos occurs. We examine the neighborhood of this Y value to find the exact value of Y for the last good surface. We call this value Y_lgs. We find Y_lgs to be 0.997135768 for k=0.60 and X=0. This work is supported by US DOE OFES. Ms. Huong Phan is a HU CFRT Summer Fusion High School Workshop Scholar from Andrew P. Hill High School in California. She is supported by NASA SHARP Plus Program. 1. Punjabi A, Verma A and Boozer A, Phys Rev Lett 69 3322 (1992) and J Plasma Phys 52 91 (1994)

MSE commissioning and other major diagnostic updates on KSTAR

NASA Astrophysics Data System (ADS)

Ko, Jinseok; Kstar Team

2015-11-01

The motional Stark effect (MSE) diagnostic based on the photoelastic-modulator (PEM) approach has been commissioned for the Korea Superconducting Tokamak Advanced Research (KSTAR). The 25-channel MSE system with the polarization-preserving front optics and precise tilt-tuning narrow bandpass filters provides the spatial resolution less than 1 cm in most of the plasma cross section and about 10 millisecond of time resolution. The polarization response curves with the daily Faraday rotation correction provides reliable pitch angle profiles for the KSTAR discharges with the MSE-optimized energy combination in the three-ion-source neutral beam injection. Some major diagnostic advances such as the poloidal charge exchange spectroscopy, the improved Thomson-scatting system, and the divertor infrared TV are reported as well. Work supported by the Ministry of Science, ICT and Future Planning, Korea.

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

NASA Astrophysics Data System (ADS)

Menard, Jonathan

2016-10-01

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

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-AC52-07NA27344, DE-AC02-09CH11466, and the NMCFP of China under 2015GB110000 and 2015GB102000.

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.

Maximizing MST's inductive capability with a Bp programmable power supply

NASA Astrophysics Data System (ADS)

Chapman, B. E.; Holly, D. J.; Jacobson, C. M.; McCollam, K. J.; Morin, J. C.; Sarff, J. S.; Squitieri, A.

2016-10-01

A major goal of the MST program is the advancement of inductive control for the development of both the RFP's fusion potential and, synergistically, the predictive capability of fusion science. This entails programmable power supplies (PPS's) for the Bt and Bp circuits. A Bt PPS is already in place, allowing advanced RFP operation and the production of tokamak plasmas, and a Bp PPS prototype is under construction. To explore some of the new capabilities to be provided by the Bp PPS, the existing Bt PPS has been temporarily connected to the Bp circuit. One key result is new-found access to very low Ip (20 kA) and very low Lundquist number, S (104). At this low S, simulation of RFP plasmas with the MHD code NIMROD is readily achievable, and work toward validation of extended MHD models using NIMROD is underway with direct comparisons to these MST plasmas. The full Bp PPS will also provide higher Ip and S than presently possible, allowing MST to produce plasmas with S spanning as much as five orders of magnitude, a dramatic extension of MST's capability. In these initial tests, the PPS has also increased five-fold MST's Ip flattop duration, to about 100 ms. This, coupled with the recently demonstrated PPS ability to drive large-amplitude sinusoidal oscillations in Ip, will allow tests of extended-duration oscillating field current drive, the goal of which is ac sustainment of a quasi-dc plasma current. Work supported by US DOE.

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

Twenty Years of Research on the Alcator C-Mod Tokamak

NASA Astrophysics Data System (ADS)

Greenwald, Martin

2013-10-01

Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

Advanced control of neutral beam injected power in DIII-D

DOE PAGES

Pawley, Carl J.; Crowley, Brendan J.; Pace, David C.; ...

2017-03-23

In the DIII-D tokamak, one of the most powerful techniques to control the density, temperature and plasma rotation is by eight independently modulated neutral beam sources with a total power of 20 MW. The rapid modulation requires a high degree of reproducibility and precise control of the ion source plasma and beam acceleration voltage. Recent changes have been made to the controls to provide a new capability to smoothly vary the beam current and beam voltage during a discharge, while maintaining the modulation capability. The ion source plasma inside the arc chamber is controlled through feedback from the Langmuir probesmore » measuring plasma density near the extraction end. To provide the new capability, the plasma control system (PCS) has been enabled to change the Langmuir probe set point and the beam voltage set point in real time. When the PCS varies the Langmuir set point, the plasma density is directly controlled in the arc chamber, thus changing the beam current (perveance) and power going into the tokamak. Alternately, the PCS can sweep the beam voltage set point by 20 kV or more and adjust the Langmuir probe setting to match, keeping the perveance constant and beam divergence at a minimum. This changes the beam power and average neutral particle energy, which changes deposition in the tokamak plasma. The ion separating magnetic field must accurately match the beam voltage to protect the beam line. To do this, the magnet current control accurately tracks the beam voltage set point. In conclusion, these new capabilities allow continuous in-shot variation of neutral beam ion energy to complement« less

Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

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

Zhu, Y. L.; Xie, J. L., E-mail: jlxie@ustc.edu.cn; Yu, C. X.

2016-11-15

Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This “4th generation” MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven bymore » fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy “general optics structure” has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.« less

Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

NASA Astrophysics Data System (ADS)

Zhu, Y. L.; Xie, J. L.; Yu, C. X.; Zhao, Z. L.; Gao, B. X.; Chen, D. X.; Liu, W. D.; Liao, W.; Qu, C. M.; Luo, C.; Hu, X.; Spear, A. G.; Luhmann, N. C.; Domier, C. W.; Chen, M.; Ren, X.; Tobias, B. J.

2016-11-01

Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This "4th generation" MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven by fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy "general optics structure" has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.

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.

Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

NASA Astrophysics Data System (ADS)

Pace, D. C.; Austin, M. E.; Bardoczi, L.; Collins, C. S.; Crowley, B.; Davis, E.; Du, X.; Ferron, J.; Grierson, B. A.; Heidbrink, W. W.; Holcomb, C. T.; McKee, G. R.; Pawley, C.; Petty, C. C.; Podestà, M.; Rauch, J.; Scoville, J. T.; Spong, D. A.; Thome, K. E.; Van Zeeland, M. A.; Varela, J.; Victor, B.

2018-05-01

An engineering upgrade to the neutral beam system at the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic ( E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2 MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities and results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.

Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

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

Austin, Max E.; Bardoczi, Laszlo; Collins, Cami S.

Here, an engineering upgrade to the neutral beam system at the DIII-D tokamak enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic (E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities andmore » results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.« less

Dynamic neutral beam current and voltage control to improve beam efficacy in tokamaks

DOE PAGES

Austin, Max E.; Bardoczi, Laszlo; Collins, Cami S.; ...

2018-04-20

Here, an engineering upgrade to the neutral beam system at the DIII-D tokamak enables time-dependent programming of the beam voltage and current. Initial application of this capability involves pre-programmed beam voltage and current injected into plasmas that are known to be susceptible to instabilities that are driven by energetic (E ≥ 40 keV) beam ions. These instabilities, here all Alfvén eigenmodes (AEs), increase the transport of the beam ions beyond a classical expectation based on particle drifts and collisions. Injecting neutral beam power, P beam ≥ 2MW, at reduced voltage with increased current reduces the drive for Alfvénic instabilities andmore » results in improved ion confinement. In lower-confinement plasmas, this technique is applied to eliminate the presence of AEs across the mid-radius of the plasmas. Simulations of those plasmas indicate that the mode drive is decreased and the radial extent of the remaining modes is reduced compared to a higher beam voltage case. In higher-confinement plasmas, this technique reduces AE activity in the far edge and results in an interesting scenario of beam current drive improving as the beam voltage reduces from 80 kV to 65 kV.« less

The Simple Map for a Single-null Divertor Tokamak: How to Look for Self-Similarity in Chaos

NASA Astrophysics Data System (ADS)

Nguyen, Christina; Ali, Halima; Punjabi, Alkesh

2000-10-01

The movement of magnetic field lines inside a single-null divertor tokamak can be described by the Simple Map^1. The Simple Map in the Poincaré Surface of Section is given by the equations: X_1=X_0-KY_0(1-Y_0) and Y_1=Y_0+KX_1. In these equations, K remains constant at 0.60. However, the values for X0 and Y0 are changed. These values are changed so that we can zoom into chaos. Chaos lies between the region (0,0.997) and (0,1). In chaos, there lies order. As we zoom into chaos, we again find chaos and order that looks like the original good surfaces and chaos. This phenomenon is called self-similarity. Self-similarity can occur for an infinite number of times if one magnifies into the chaotic region. For this work, we write a program in a computer language called Fortran 77 and Gnuplot. This work is supported by US DOE OFES. Ms. Christina Nguyen is a HU CFRT Summer Fusion High School Workshop Scholar from Andrew Hill High School in California. She is supported by NASA SHARP Plus Program. 1. Punjabi A, Verma A and Boozer A, Phys Rev Lett 69 3322 (1992) and J Plasma Phys 52 91 (1994)

The Helicity Injected Torus (HIT) Program

NASA Astrophysics Data System (ADS)

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

2000-10-01

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

Simulations of the effects of density and temperature profile on SMBI penetration depth based on the HL-2A tokamak configuration

NASA Astrophysics Data System (ADS)

Wu, Xueke; Li, Huidong; Wang, Zhanhui; Feng, Hao; Zhou, Yulin

2017-06-01

Not Available Project supported by the National Natural Science Foundation for Young Scientists of China (Grant No. 11605143), the Undergraduate Training Programs for Innovation and Entrepreneurship of Sichuan Province, China (Grant No. 05020732), the National Natural Science Foundation of China (Grant No. 11575055), the Fund from the Department of Education in Sichuan Province of China (Grant No. 15ZB0129), the China National Magnetic Confinement Fusion Science Program (Grant No. 2013GB107001), the National ITER Program of China (Contract No. 2014GB113000), and the Funds of the Youth Innovation Team of Science and Technology in Sichuan Province of China (Grant No. 2014TD0023).

Collision broadened resonance localization in tokamaks excited with ICRF waves

NASA Astrophysics Data System (ADS)

Kerbel, G. D.; McCoy, M. G.

1985-08-01

Advanced wave models used to evaluate ICRH in tokamaks typically use warm plasma theory and allow inhomogeneity in one dimension. The authors have developed a bounce-averaged Fokker-Planck quasilinear computational model which evolves the population of particles on more realistic orbits. Each wave-particle resonance has its own specific interaction amplitude within any given volume element. These data need only be generated once, and appropriately stored for efficient retrieval. The wave-particle resonant interaction then serves as a mechanism by which the diffusion of particle populations can proceed among neighboring orbits. Collisions affect the absorption of RF energy by two quite distinct processes: In addition to the usual relaxation towards the Maxwellian distribution creating velocity gradients which drive quasilinear diffusion, collisions also affect the wave-particle resonance through the mechanism of gyro-phase diffusion. The local specific spectral energy absorption rate is directly calculable once the orbit geometry and populations are determined. The code is constructed in such fashion as to accommodate wave propagation models which provide the wave spectral energy density on a poloidal cross-section. Information provided by the calculation includes the local absorption properties of the medium which can then be exploited to evolve the wave field.

Resonance localization in tokamaks excited with ICRF waves

NASA Astrophysics Data System (ADS)

Kerbel, G. D.; McCoy, M. G.

1985-06-01

Advanced wave model used to evaluate ICRH in tokamaks typically used warm plasma theory and allow inhomogeneity in one dimension. The majority of these calculations neglect the fact that gyrocenters experience the inhomogeneity via their motion parallel to the magnetic field. In strongly driven systems, wave damping can distort the particle distribution function supporting the wave and this produces changes in the absorption. A bounce-averaged Fokker-Planck quasilinear computational model which evolves the population of particles on more realistic orbits is presented. Each wave-particle resonance has its own specific interaction amplitude within any given volume element; these data need only be generated once, and appropriately stored for efficient retrieval. The wave-particle resonant interaction then serves as a mechanism by which the diffusion of particle populations can proceed among neighboring orbits. The local specific spectral energy absorption rate is directly calculable once the orbit geometry and populations are determined. The code is constructed in such fashion as to accommodate wave propagation models which provide the wave spectral energy density on a poloidal cross-section. Information provided by the calculation includes the local absorption properties of the medium which can then be exploited to evolve the wave field.

Compatibility of internal transport barrier with steady-state operation in the high bootstrap fraction regime on DIII-D

DOE PAGES

Garofalo, Andrea M.; Gong, Xianzu; Grierson, Brian A.; ...

2015-11-16

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by ≥30% relative to earlier work. The advancement was enabled by improved understanding of the “relaxation oscillations”, previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the “relaxation oscillations” are coupled core-edge modes 2 amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced tomore » classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ≥ 80%, β N ≤ 4 , β P ≥ 3 , and β T ≥ 2%. Finally, these results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.« less

The accurate particle tracer code

DOE PAGES

Wang, Yulei; Liu, Jian; Qin, Hong; ...

2017-07-20

The Accurate Particle Tracer (APT) code is designed for systematic large-scale applications of geometric algorithms for particle dynamical simulations. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and nonlinear problems. To provide a flexible and convenient I/O interface, the libraries of Lua and Hdf5 are used. Following a three-step procedure, users can efficiently extend the libraries of electromagnetic configurations, external non-electromagnetic forces, particle pushers, and initialization approaches by use of the extendible module. APT has been used in simulations of key physical problems, such as runawaymore » electrons in tokamaks and energetic particles in Van Allen belt. As an important realization, the APT-SW version has been successfully distributed on the world’s fastest computer, the Sunway TaihuLight supercomputer, by supporting master–slave architecture of Sunway many-core processors. Here, based on large-scale simulations of a runaway beam under parameters of the ITER tokamak, it is revealed that the magnetic ripple field can disperse the pitch-angle distribution significantly and improve the confinement of energetic runaway beam on the same time.« less

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

Lampert, M.; BME NTI, Budapest; Anda, G.

A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, whilemore » a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.« less

The accurate particle tracer code

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

Wang, Yulei; Liu, Jian; Qin, Hong

The Accurate Particle Tracer (APT) code is designed for systematic large-scale applications of geometric algorithms for particle dynamical simulations. Based on a large variety of advanced geometric algorithms, APT possesses long-term numerical accuracy and stability, which are critical for solving multi-scale and nonlinear problems. To provide a flexible and convenient I/O interface, the libraries of Lua and Hdf5 are used. Following a three-step procedure, users can efficiently extend the libraries of electromagnetic configurations, external non-electromagnetic forces, particle pushers, and initialization approaches by use of the extendible module. APT has been used in simulations of key physical problems, such as runawaymore » electrons in tokamaks and energetic particles in Van Allen belt. As an important realization, the APT-SW version has been successfully distributed on the world’s fastest computer, the Sunway TaihuLight supercomputer, by supporting master–slave architecture of Sunway many-core processors. Here, based on large-scale simulations of a runaway beam under parameters of the ITER tokamak, it is revealed that the magnetic ripple field can disperse the pitch-angle distribution significantly and improve the confinement of energetic runaway beam on the same time.« less

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.

Revitalizing Fusion via Fission Fusion

NASA Astrophysics Data System (ADS)

Manheimer, Wallace

2001-10-01

Existing tokamaks could generate significant nuclear fuel. TFTR, operating steady state with DT might generate enough fuel for a 300 MW nuclear reactor. The immediate goals of the magnetic fusion program would necessarily shift from a study of advanced plasma regimes in larger sized devices, to mostly known plasmas regimes, but at steady state or high duty cycle operation in DT plasmas. The science and engineering of breeding blankets would be equally important. Follow on projects could possibly produce nuclear fuel in large quantity at low price. Although today there is strong opposition to nuclear power in the United States, in a 21st century world of 10 billion people, all of whom will demand a middle class life style, nuclear energy will be important. Concern over greenhouse gases will also drive the world toward nuclear power. There are studies indicating that the world will need 10 TW of carbon free energy by 2050. It is difficult to see how this can be achieved without the breeding of nuclear fuel. By using the thorium cycle, proliferation risks are minimized. [1], [2]. 1 W. Manheimer, Fusion Technology, 36, 1, 1999, 2.W. Manheimer, Physics and Society, v 29, #3, p5, July, 2000

Development of Integrated Magnetic and Kinetic Control-oriented Transport Model for q-profile Response Prediction in EAST Discharges

NASA Astrophysics Data System (ADS)

Wang, Hexiang; Schuster, Eugenio; Rafiq, Tariq; Kritz, Arnold; Ding, Siye

2016-10-01

Extensive research has been conducted to find high-performance operating scenarios characterized by high fusion gain, good confinement, plasma stability and possible steady-state operation. A key plasma property that is related to both the stability and performance of these advanced plasma scenarios is the safety factor profile. A key component of the EAST research program is the exploration of non-inductively driven steady-state plasmas with the recently upgraded heating and current drive capabilities that include lower hybrid current drive and neutral beam injection. Anticipating the need for tight regulation of the safety factor profile in these plasma scenarios, a first-principles-driven (FPD)control-oriented model is proposed to describe the safety factor profile evolution in EAST in response to the different actuators. The TRANSP simulation code is employed to tailor the FPD model to the EAST tokamak geometry and to convert it into a form suitable for control design. The FPD control-oriented model's prediction capabilities are demonstrated by comparing predictions with experimental data from EAST. Supported by the US DOE under DE-SC0010537,DE-FG02-92ER54141 and DE-SC0013977.

High-Performance Computational Modeling of ICRF Physics and Plasma-Surface Interactions in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Jenkins, Thomas; Smithe, David

2016-10-01

Inefficiencies and detrimental physical effects may arise in conjunction with ICRF heating of tokamak plasmas. Large wall potential drops, associated with sheath formation near plasma-facing antenna hardware, give rise to high-Z impurity sputtering from plasma-facing components and subsequent radiative cooling. Linear and nonlinear wave excitations in the plasma edge/SOL also dissipate injected RF power and reduce overall antenna efficiency. Recent advances in finite-difference time-domain (FDTD) modeling techniques allow the physics of localized sheath potentials, and associated sputtering events, to be modeled concurrently with the physics of antenna near- and far-field behavior and RF power flow. The new methods enable time-domain modeling of plasma-surface interactions and ICRF physics in realistic experimental configurations at unprecedented spatial resolution. We present results/animations from high-performance (10k-100k core) FDTD/PIC simulations spanning half of Alcator C-Mod at mm-scale resolution, exploring impurity production due to localized sputtering (in response to self-consistent sheath potentials at antenna surfaces) and the physics of parasitic slow wave excitation near the antenna hardware and SOL. Supported by US DoE (Award DE-SC0009501) and the ALCC program.

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.

Study of high field side/low field side asymmetry in the electron temperature profile with electron cyclotron emission

NASA Astrophysics Data System (ADS)

Gugliada, V. R.; Austin, M. E.; Brookman, M. W.

2017-10-01

Electron cyclotron emission (ECE) provides high resolution measurements of electron temperature profiles (Te(R , t)) in tokamaks. Calibration accuracy of this data can be improved using a sawtooth averaging technique. This improved calibration will then be utilized to determine the symmetry of Te profiles by comparing low field side (LFS) and high field side (HFS) measurements. Although Te is considered constant on flux surfaces, cases have been observed in which there are pronounced asymmetries about the magnetic axis, particularly with increased pressure. Trends in LFS/HFS overlap are examined as functions of plasma pressure, MHD mode presence, heating techniques, and other discharge conditions. This research will provide information on the accuracy of the current two-dimensional mapping of flux surfaces in the tokamak. Findings can be used to generate higher quality EFITs and inform ECE calibration. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER549698.

Recent progress of the Laser-driven Ion-beam Trace Probe

NASA Astrophysics Data System (ADS)

Yang, Xiaoyi; Xiao, Chijie; Chen, Yihang; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

2017-10-01

The Laser-driven Ion-beam Trace Probe (LITP) is a new method to diagnose the poloidal magnetic field and radial electric field in tokamaks. Recently significant progresses have been made as follows. 1) The experimental system has been set up on the PKU Plasma Test (PPT) linear device and begun to validate the principle of LITP, including the ion source, the ion detector and the poloidal magnetic field cable. Preliminary experimental results matched the theoretical prediction well. 2) The reconstruction principle has been improved including the nonlinear effect. 3) Tomography methods have been applied in the reconstruction codes. Now the laser-driven ion-beam accelerator has been setup on the PPT device, and further test of LITP will start soon. After that a prototype of LITP system will be designed and setup on the HL-2A tokamak device. This work was supported by the CHINA MOST under 2012YQ030142, ITER-CHINA program 2015GB120001 and National Natural Science Foundation of China under 11575014 and 11375053.

Development of KSTAR Thomson scattering system.

PubMed

Lee, J H; Oh, S T; Wi, H M

2010-10-01

To measure the electron temperature (T(e)) and electron density (n(e)) profiles in the Korean Superconducting Tokamak Advanced Research (KSTAR) device for the KSTAR third campaign (September 2010), we designed and installed a Thomson scattering system. The KSTAR Thomson scattering system is designed as a tangential Thomson scattering system and utilizes the N-, L-, and B-ports. The N-port is designed for the collection optics with a cassette system, the L-port is the laser input port, and the B-port is the location of the beam dump. In this paper, we will describe the final design of the KSTAR Thomson scattering system.

Princeton Plasma Physics Laboratory: Annual report, October 1, 1986--September 30, 1987

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

Not Available

1987-01-01

This report contains papers on the following topics: Principle Parameters Achieved in Experimental Devices (FY87); Tokamak Fusion Test Reactor; Princeton Beta Experiment-Modification; S-1 Spheromak; Current-Drive Experiment; X-Ray Laser Studies; Theoretical Division; Tokamak Modeling; Compact Ignition Tokamak; Engineering Department; Project Planning and Safety Office; Quality Assurance and Reliability; Administrative Operations; and PPPL Patent Invention Disclosures (FY87).

Alternative approaches to plasma confinement

NASA Technical Reports Server (NTRS)

Roth, J. R.

1977-01-01

The potential applications of fusion reactors, the desirable properties of reactors intended for various applications, and the limitations of the Tokamak concept are discussed. The principles and characteristics of 20 distinct alternative confinement concepts are described, each of which may be an alternative to the Tokamak. The devices are classed as Tokamak-like, stellarator-like, mirror machines, bumpy tori, electrostatically assisted, migma concept, and wall-confined plasma.

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.

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

Stewart Zweben; Samuel Cohen; Hantao Ji

Small ''concept exploration'' experiments have for many years been an important part of the fusion research program at the Princeton Plasma Physics Laboratory (PPPL). this paper describes some of the present and planned fusion concept exploration experiments at PPPL. These experiments are a University-scale research level, in contrast with the larger fusion devices at PPPL such as the National Spherical Torus Experiment (NSTX) and the Tokamak Fusion Test Reactor (TFTR), which are at ''proof-of-principle'' and ''proof-of-performance'' levels, respectively.

Dusty (complex) plasmas: recent developments, advances, and unsolved problems

NASA Astrophysics Data System (ADS)

Popel, Sergey

The area of dusty (complex) plasma research is a vibrant subfield of plasma physics that be-longs to frontier research in physical sciences. This area is intrinsically interdisciplinary and encompasses astrophysics, planetary science, atmospheric science, magnetic fusion energy sci-ence, and various applied technologies. The research in dusty plasma started after two major discoveries in very different areas: (1) the discovery by the Voyager 2 spacecraft in 1980 of the radial spokes in Saturn's B ring, and (2) the discovery of the early 80's growth of contaminating dust particles in plasma processing. Dusty plasmas are ubiquitous in the universe; examples are proto-planetary and solar nebulae, molecular clouds, supernovae explosions, interplanetary medium, circumsolar rings, and asteroids. Within the solar system, we have planetary rings (e.g., Saturn and Jupiter), Martian atmosphere, cometary tails and comae, dust clouds on the Moon, etc. Close to the Earth, there are noctilucent clouds and polar mesospheric summer echoes, which are clouds of tiny (charged) ice particles that are formed in the summer polar mesosphere at the altitudes of about 82-95 km. Dust and dusty plasmas are also found in the vicinity of artificial satellites and space stations. Dust also turns out to be common in labo-ratory plasmas, such as in the processing of semiconductors and in tokamaks. In processing plasmas, dust particles are actually grown in the discharge from the reactive gases used to form the plasmas. An example of the relevance of industrial dusty plasmas is the growth of silicon microcrystals for improved solar cells in the future. In fact, nanostructured polymorphous sili-con films provide solar cells with high and time stable efficiency. These nano-materials can also be used for the fabrication of ultra-large-scale integration circuits, display devices, single elec-tron devices, light emitting diodes, laser diodes, and others. In microelectronic industries, dust has to be kept under control in the manufacture of microchips, otherwise charged dust particles (also known as killer particles) can destroy electronic circuits. In magnetic fusion research using tokamaks, one realizes that the absorption of tritium by dust fragments could cause a serious health hazard. The evaporation of dust particles could also lead to bremsstrahlung adversely affecting the energy gain of the tokamaks or other fusion devices. The specific features of dusty plasmas are a possibility of the formation of dust Coulomb lattices and the anomalous dissi-pation arising due to the interplay between plasmas and charged dust grains. These features determine new physics of dusty plasmas including, in particular, phase transitions and critical point phenomena, wave propagation, nonlinear effects and turbulence, dissipative and coherent structures, etc. The present review covers the main aspects of the area of dusty (complex) plasma research. The author acknowledges the financial support of the Division of Earth Sci-ences, Russian Academy of Sciences (the basic research program "Nanoscale particles in nature and technogenic products: conditions of existence, physical and chemical properties, and mech-anisms of formation"'), of the Division of Physical Sciences, Russian Academy of Sciences (the basic research program "Plasma physics in the Solar system"), of the Dynasty Foundation, as well as of the Russian Foundation for Basic Research.

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.

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

The ITPA disruption database

DOE PAGES

Eidietis, N. W.; Gerhardt, S. P.; Granetz, R. S.; ...

2015-05-22

A multi-device database of disruption characteristics has been developed under the auspices of the International Tokamak Physics Activity magneto hydrodynamics topical group. The purpose of this ITPA Disruption Database (IDDB) is to find the commonalities between the disruption and disruption mitigation characteristics in a wide variety of tokamaks in order to elucidate the physics underlying tokamak disruptions and to extrapolate toward much larger devices, such as ITER and future burning plasma devices. Conversely, in order to previous smaller disruption data collation efforts, the IDDB aims to provide significant context for each shot provided, allowing exploration of a wide array ofmore » relationships between pre-disruption and disruption parameters. Furthermore, the IDDB presently includes contributions from nine tokamaks, including both conventional aspect ratio and spherical tokamaks. An initial parametric analysis of the available data is presented. Our analysis includes current quench rates, halo current fraction and peaking, and the effectiveness of massive impurity injection. The IDDB is publicly available, with instruction for access provided herein.« less

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

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.

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.

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.

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 needed particularly for the 3D divertor tokamak configuration.« less

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 needed particularly for the 3D divertor tokamak configuration.« less

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.

U. S. fusion programs: Struggling to stay in the game

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

Crawford, M.

Funding for the US fusion energy program has suffered and will probably continue to suffer major cuts. A committee hand-picked by Energy Secretary James Watkins urged the Department of Energy to mount an aggressive program to develop fusion power, but congress cut funding from $323 million in 1990 to $275 million in 1991. This portends dire conditions for fusion research and development. Projects to receive top priority are concerned with the tokamaks and to keep the next big machine, the Burning Plasma Experiment, scheduled for beginning of construction in 1993 on schedule. Secretary Watkins is said to want to keepmore » the International Thermonuclear Energy Reactor (ITER) on schedule. ITER would follow the Burning Plasma Experiment.« less

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

NASA Astrophysics Data System (ADS)

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

2004-05-01

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

SciDAC GSEP: Gyrokinetic Simulation of Energetic Particle Turbulence and Transport

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

Lin, Zhihong

Energetic particle (EP) confinement is a key physics issue for burning plasma experiment ITER, the crucial next step in the quest for clean and abundant energy, since ignition relies on self-heating by energetic fusion products (α-particles). Due to the strong coupling of EP with burning thermal plasmas, plasma confinement property in the ignition regime is one of the most uncertain factors when extrapolating from existing fusion devices to the ITER tokamak. EP population in current tokamaks are mostly produced by auxiliary heating such as neutral beam injection (NBI) and radio frequency (RF) heating. Remarkable progress in developing comprehensive EP simulationmore » codes and understanding basic EP physics has been made by two concurrent SciDAC EP projects GSEP funded by the Department of Energy (DOE) Office of Fusion Energy Science (OFES), which have successfully established gyrokinetic turbulence simulation as a necessary paradigm shift for studying the EP confinement in burning plasmas. Verification and validation have rapidly advanced through close collaborations between simulation, theory, and experiment. Furthermore, productive collaborations with computational scientists have enabled EP simulation codes to effectively utilize current petascale computers and emerging exascale computers. We review here key physics progress in the GSEP projects regarding verification and validation of gyrokinetic simulations, nonlinear EP physics, EP coupling with thermal plasmas, and reduced EP transport models. Advances in high performance computing through collaborations with computational scientists that enable these large scale electromagnetic simulations are also highlighted. These results have been widely disseminated in numerous peer-reviewed publications including many Phys. Rev. Lett. papers and many invited presentations at prominent fusion conferences such as the biennial International Atomic Energy Agency (IAEA) Fusion Energy Conference and the annual meeting of the American Physics Society, Division of Plasma Physics (APS-DPP).« 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.

Method of sustaining a radial electric field and poloidal plasma rotation over most of the cross-section of a tokamak

DOEpatents

Darrow, Douglass S.; Ono, Masayuki

1990-03-06

A radial electric field of a desired magnitude and configuration is created throughout a substantial portion of the cross-section of the plasma of a tokamak. The radial electric field is created by injection of a unidirectional electron beam. The magnitude and configuration of the radial electric field may be controlled by the strength of the toroidal magnetic field of the tokamak.

Method of sustaining a radial electric field and poloidal plasma rotation over most of the cross-section of a tokamak

DOEpatents

Darrow, Douglass S.; Ono, Masayuki

1990-01-01

A radial electric field of a desired magnitude and configuration is created hroughout a substantial portion of the cross-section of the plasma of a tokamak. The radial electric field is created by injection of a unidirectional electron beam. The magnitude and configuration of the radial electric field may be controlled by the strength of the toroidal magnetic field of the tokamak.

Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak

DOE PAGES

Mitarai, O.; Xiao, C.; McColl, D.; ...

2015-03-24

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. Our results suggest a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. Finally, the effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments inmore » the STOR-M tokamak.« less

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

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.

Newly developed double neural network concept for reliable fast plasma position control

NASA Astrophysics Data System (ADS)

Jeon, Young-Mu; Na, Yong-Su; Kim, Myung-Rak; Hwang, Y. S.

2001-01-01

Neural network is considered as a parameter estimation tool in plasma controls for next generation tokamak such as ITER. The neural network has been reported to be so accurate and fast for plasma equilibrium identification that it may be applied to the control of complex tokamak plasmas. For this application, the reliability of the conventional neural network needs to be improved. In this study, a new idea of double neural network is developed to achieve this. The new idea has been applied to simple plasma position identification of KSTAR tokamak for feasibility test. Characteristics of the concept show higher reliability and fault tolerance even in severe faulty conditions, which may make neural network applicable to plasma control reliably and widely in future tokamaks.

Plasma current start-up by the outer ohmic heating coils in the Saskatchewan TORus Modified (STOR-M) iron core tokamak

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

Mitarai, O.; Xiao, C.; McColl, D.

A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. Our results suggest a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. Finally, the effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments inmore » the STOR-M tokamak.« less

Advancing Pre-college Science and Mathematics Education

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

Lee, Rick

With support from the US Department of Energy, Office of Science, Fusion Energy Sciences, and General Atomics, an educational and outreach program primarily for grades G6-G13 was developed using the basic science of plasma and fusion as the content foundation. The program period was 1994 - 2015 and provided many students and teachers unique experiences such as a visit to the DIII-D National Fusion Facility to tour the nation’s premiere tokamak facility or to interact with interesting and informative demonstration equipment and have the opportunity to increase their understanding of a wide range of scientific content, including states of matter,more » the electromagnetic spectrum, radiation & radioactivity, and much more. Engaging activities were developed for classroom-size audiences, many made by teachers in Build-it Day workshops. Scientist and engineer team members visited classrooms, participated in science expositions, held workshops, produced informational handouts in paper, video, online, and gaming-CD format. Participants could interact with team members from different institutions and countries and gain a wider view of the world of science and engineering educational and career possibilities. In addition, multiple science stage shows were presented to audiences of up to 700 persons in a formal theatre setting over a several day period at Science & Technology Education Partnership (STEP) Conferences. Annually repeated participation by team members in various classroom and public venue events allowed for the development of excellent interactive skills when working with students, teachers, and educational administrative staff members. We believe this program has had a positive impact in science understanding and the role of the Department of Energy in fusion research on thousands of students, teachers, and members of the general public through various interactive venues.« less

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.

Liquid Metals as Plasma-facing Materials for Fusion Energy Systems: From Atoms to Tokamaks

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

Stone, Howard A.; Koel, Bruce E.; Bernasek, Steven L.

The objective of our studies was to advance our fundamental understanding of liquid metals as plasma-facing materials for fusion energy systems, with a broad scope: from atoms to tokamaks. The flow of liquid metals offers solutions to significant problems of the plasma-facing materials for fusion energy systems. Candidate metals include lithium, tin, gallium, and their eutectic combinations. However, such liquid metal solutions can only be designed efficiently if a range of scientific and engineering issues are resolved that require advances in fundamental fluid dynamics, materials science and surface science. In our research we investigated a range of significant and timelymore » problems relevant to current and proposed engineering designs for fusion reactors, including high-heat flux configurations that are being considered by leading fusion energy groups world-wide. Using experimental and theoretical tools spanning atomistic to continuum descriptions of liquid metals, and bridging surface chemistry, wetting/dewetting and flow, our research has advanced the science and engineering of fusion energy materials and systems. Specifically, we developed a combined experimental and theoretical program to investigate flows of liquid metals in fusion-relevant geometries, including equilibrium and stability of thin-film flows, e.g. wetting and dewetting, effects of electromagnetic and thermocapillary fields on liquid metal thin-film flows, and how chemical interactions and the properties of the surface are influenced by impurities and in turn affect the surface wetting characteristics, the surface tension, and its gradients. Because high-heat flux configurations produce evaporation and sputtering, which forces rearrangement of the liquid, and any dewetting exposes the substrate to damage from the plasma, our studies addressed such evaporatively driven liquid flows and measured and simulated properties of the different bulk phases and material interfaces. The range of our studies included (i) quantum mechanical calculations that allow inclusion of many thousands of atoms for the characterization of the interface of liquid metals exposed to continuous bombardment by deuterium and tritium as expected in fusion, (ii) molecular dynamics studies of the phase behavior of liquid metals, which (a) utilize thermodynamic properties computed using our quantum mechanical calculations and (b) establish material and wetting properties of the liquid metals, including relevant eutectics, (iii) experimental investigations of the surface science of liquid metals, interacting both with the solid substrate as well as gaseous species, and (iv) fluid dynamical studies that incorporate the material and surface science results of (ii) and (iii) in order to characterize flow in capillary porous materials and the thin-film flow along curved boundaries, both of which are potentially major components of plasma-facing materials. The outcome of these integrated studies was new understanding that enables developing design rules useful for future developments of the plasma-facing components critical to the success of fusion energy systems.« less

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

Likelihood of Alfvénic instability bifurcation in experiments

NASA Astrophysics Data System (ADS)

Duarte, Vinicius; Gorelenkov, Nikolai; Schneller, Mirjam; Fredrickson, Eric; Berk, Herbert; Canal, Gustavo; Heidbrink, William; Kaye, Stanley; Podesta, Mario; van Zeeland, Michael; Wang, Weixing

2017-10-01

We apply a criterion for the likely nature of fast ion redistribution in tokamaks to be in the convective or diffusive nonlinear regimes. The criterion, which is shown to be rather sensitive to the relative strength of collisional or micro-turbulent scattering and drag processes, ultimately translates into a condition for the applicability of reduced quasilinear modeling for realistic tokamak eigenmodes scenarios. The criterion is tested and validated against different machines, where the chirping mode behavior is shown to be in accord with the model. It has been found that the anomalous fast ion transport is a likely mediator of the bifurcation between the fixed-frequency mode behavior and rapid chirping in tokamaks. In addition, micro-turbulence appears to resolve the disparity with respect to the ubiquitous chirping observation in spherical tokamaks and its rarer occurrence in conventional tokamaks. In NSTX, the tendency for chirping is further studied in terms of the beam beta and the plasma rotation shear. For more accurate quantitative assessment, numerical simulations of the effects of electrostatic ion temperature gradient turbulence on chirping are presently being pursued using the GTS code.

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.

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.

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.

Hybrid simulation of fishbone instabilities in the EAST tokamak

DOE PAGES

Shen, Wei; Wang, Feng; Fu, G. Y.; ...

2017-08-11

Hybrid simulations with the global kinetic-magnetohydrodynamic (MHD) code M3D-K have been carried out to investigate the linear stability and nonlinear dynamics of beam-driven fishbone in the experimental advanced superconducting tokamak (EAST) experiment. Linear simulations show that a low frequency fishbone instability is excited at experimental value of beam ion pressure. The mode is mainly driven by low energy beam ions via precessional resonance. Our results are consistent with the experimental measurement with respect to mode frequency and mode structure. When the beam ion pressure is increased to exceed a critical value, the low frequency mode transits to a beta-induced Alfvenmore » eigenmode (BAE) with much higher frequency. This BAE is driven by higher energy beam ions. Nonlinear simulations show that the frequency of the low frequency fishbone chirps up and down with corresponding hole-clump structures in phase space, consistent with the Berk-Breizman theory. In addition to the low frequency mode, the high frequency BAE is excited during the nonlinear evolution. Furthermore, for the transient case of beam pressure fraction where the low and high frequency modes are simultaneously excited in the linear phase, only one dominant mode appears in the nonlinear phase with frequency jumps up and down during nonlinear evolution.« less

Direct measurements of safety factor profiles with motional Stark effect for KSTAR tokamak discharges with internal transport barriers

NASA Astrophysics Data System (ADS)

Ko, J.; Chung, J.

2017-06-01

The safety factor profile evolutions have been measured from the plasma discharges with the external current drive mechanism such as the multi-ion-source neutral beam injection for the Korea Superconducting Tokamak Advanced Research (KSTAR) for the first time. This measurement has been possible by the newly installed motional Stark effect (MSE) diagnostic system that utilizes the polarized Balmer-alpha emission from the energetic neutral deuterium atoms induced by the Stark effect under the Lorentz electric field. The 25-channel KSTAR MSE diagnostic is based on the conventional photoelastic modulator approach with the spatial and temporal resolutions less than 2 cm (for the most of the channels except 2 to 3 channels inside the magnetic axis) and about 10 ms, respectively. The strong Faraday rotation imposed on the optical elements in the diagnostic system is calibrated out from a separate and well-designed polarization measurement procedure using an in-vessel reference polarizer during the toroidal-field ramp-up phase before the plasma experiment starts. The combination of the non-inductive current drive during the ramp-up and shape control enables the formation of the internal transport barrier where the pitch angle profiles indicate flat or slightly hollow profiles in the safety factor.

Installation and pre-commissioning of the cryogenic system of JT-60SA tokamak

NASA Astrophysics Data System (ADS)

Hoa, C.; Michel, F.; Roussel, P.; Fejoz, P.; Girard, S.; Goncalves, R.; Lamaison, V.; Natsume, K.; Kizu, K.; Koide, Y.; Yoshida, K.; Cardella, A.; Portone, A.; Verrecchia, M.; Wanner, M.; Beauvisage, J.; Bertholat, F.; Gaillard, G.; Heloin, V.; Langevin, B.; Legrand, J.; Maire, S.; Perrier, J. M.; Pudys, V.

2017-02-01

The cryogenic system for the superconducting tokamak JT-60SA is currently being commissioned in Naka, Japan and shall be ready for operation in summer 2016. This contribution is part of the Broader Approach agreement between Japan and Europe. With an equivalent refrigeration capacity of about 9.5 kW at 4.5 K the cryogenic system will supply cryo-pump panels at 3.7 K, superconducting magnets and their structures at 4.4 K, high temperature superconducting current leads at 50 K and thermal shields between 80 K and 100 K. The system has been specifically designed to handle large pulse loads at 4.4 K during plasma operation. The mechanical and electrical assembly of the cryogenic system has been achieved within six months by October 2015. The main contractor Air Liquide Advanced Technology (AL-aT) have supplied eight parallel working screw compressors with a common oil removal and dryer system, a Refrigeration Cold Box and an Auxiliary Cold box with cold rotating machines. F4E has provided six GHe storage vessels and QST has provided the complete infrastructure and the facilities for the utilities. The paper gives an overview of the main design features, the infrastructure and the status of installation and pre-commissioning.

3-D plasma boundary and plasma wall interaction research at UW-Madison

NASA Astrophysics Data System (ADS)

Schmitz, Oliver; Akerson, Adrian; Bader, Aaron; Barbui, Tullio; Effenberg, Florian; Flesch, Kurt; Frerichs, Heinke; Green, Jonathan; Hinson, Edward; Kremeyer, Thierry; Norval, Ryan; Stephey, Laurie; Waters, Ian; Winters, Victoria

2016-10-01

The necessity of considering 3-D effects on the plasma boundary and plasma wall interaction (PWI) in tokamaks, stellarators and reversed field pinches has been highlighted by abundant experimental and numerical results in the recent past. Prominent examples with 3-D boundary situations are numerous: ELM controlled H-modes by RMP fields in tokamaks, research on boundary plasmas and PWI in stellarators in general, quasi-helical states in RFPs, asymmetric fueling situations, and structural and wall elements which are not aligned with the magnetic guiding fields. A systematic approach is being taken at UW-Madison to establish a targeted experimental basis for identifying the most significant effects for plasma edge transport and resulting PWI in such 3-D plasma boundary situations. We deploy advanced 3-D modeling using the EMC3-EIRENE, ERO and MCI codes in combination with laboratory experiments at UW-Madison to investigate the relevance of 3-D effects in large scale devices with a concerted approach on DIII-D, NSTX-U, and Wendelstein 7-X. Highlights of experimental results from the on-site laboratory activities at UW-Madison and the large scale facilities are presented and interlinks will be discussed. This work was supported by US DOE DE-SC0013911, DE-SC00012315 and DE-SC00014210.

LETTER: ECH pre-ionization and assisted startup in the fully superconducting KSTAR tokamak using second harmonic

NASA Astrophysics Data System (ADS)

Bae, Y. S.; Jeong, J. H.; Park, S. I.; Joung, M.; Kim, J. H.; Hahn, S. H.; Yoon, S. W.; Yang, H. L.; Kim, W. C.; Oh, Y. K.; England, A. C.; Namkung, W.; Cho, M. H.; Jackson, G. L.; Bak, J. S.; KSTAR Team

2009-02-01

This letter reports on the successful demonstration of the second harmonic electron cyclotron heating (ECH)-assisted startup in the first plasma experiments recently completed in the fully superconducting Korea Superconducting Tokamak Advanced Research (KSTAR) device whose major and minor radii are 1.8 m and 0.5 m, respectively. For the second harmonic ECH-assisted startup, an 84 GHz EC wave at 0.35 MW was launched before the onset of the toroidal electric field of the Ohmic system. And it was observed that this was sufficient to achieve breakdown in the ECH pre-ionization phase, allow burn-through and sustain the plasma during the current ramp with a low loop voltage of 2.0 V and a corresponding toroidal electric field of 0.24 V m-1at the innermost vacuum vessel wall (R = 1.3 m). This is a lower value than 0.3 Vm-1 which is the maximum electric field in ITER. Due to the limited volt-seconds and the loop voltage of the Ohmic power system, the extended pulse duration of the ECH power up to 180 ms allowed the plasma current to rise up to more than 100 kA with a ramp-up rate of 0.8 MA s-1.

Operational present status and reliability analysis of the upgraded EAST cryogenic system

NASA Astrophysics Data System (ADS)

Zhou, Z. W.; Y Zhang, Q.; Lu, X. F.; Hu, L. B.; Zhu, P.

2017-12-01

Since the first commissioning in 2005, the cryogenic system for EAST (Experimental Advanced Superconducting Tokamak) has been cooled down and warmed up for thirteen experimental campaigns. In order to promote the refrigeration efficiencies and reliability, the EAST cryogenic system was upgraded gradually with new helium screw compressors and new dynamic gas bearing helium turbine expanders with eddy current brake to improve the original poor mechanical and operational performance from 2012 to 2015. Then the totally upgraded cryogenic system was put into operation in the eleventh cool-down experiment, and has been operated for the latest several experimental campaigns. The upgraded system has successfully coped with various normal operational modes during cool-down and 4.5 K steady-state operation under pulsed heat load from the tokamak as well as the abnormal fault modes including turbines protection stop. In this paper, the upgraded EAST cryogenic system including its functional analysis and new cryogenic control networks will be presented in detail. Also, its operational present status in the latest cool-down experiments will be presented and the system reliability will be analyzed, which shows a high reliability and low fault rate after upgrade. In the end, some future necessary work to meet the higher reliability requirement for future uninterrupted long-term experimental operation will also be proposed.

Improvement of Current Drive Efficiency in Projected FNSF Discharges

NASA Astrophysics Data System (ADS)

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

2012-10-01

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

Overview, Progress, and Plans for the Compact Toroidal Hybrid Experiment

NASA Astrophysics Data System (ADS)

Hartwell, G. J.; Allen, N. R.; Ennis, D. A.; Hanson, J. D.; Howell, E. C.; Johnson, C. A.; Knowlton, S. F.; Kring, J. D.; Ma, X.; Maurer, D. A.; Ross, K. G.; Schmitt, J. C.; Traverso, P. J.; Williamson, E. N.

2017-10-01

The Compact Toroidal Hybrid (CTH) is an l = 2 , m = 5 torsatron/tokamak hybrid (R0 = 0.75 m, ap 0.2 m, and | B | <= 0.7 T) which generates highly configurable confining magnetic fields solely with external coils but typically uses up to 80 kA of plasma current for heating and disruption studies. The main goals of the CTH experiment are to study disruptive behavior as a function of applied 3D magnetic shaping, and to test and advance the V3FIT reconstruction code and NIMROD modeling of CTH. The disruptive density limit is observed to exceed the Greenwald limit as the vacuum transform is increased with no observed threshold for avoidance. Low-q operations (1.1 < q(a) < 2.0) are routine, with disruptions ceasing if the vacuum transform is raised above 0.07. Sawteeth are observed in CTH and have a similar phenomenology to tokamak sawteeth despite employing a 3D confining field. Application of vacuum transform has been demonstrated to reduce and eliminate the vertical drift of elongated discharges. Internal SXR diagnostics, in conjunction with external magnetics, extend the range of reconstruction accuracy into the plasma core. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

Three-dimensional propagation and absorption of high frequency Gaussian beams in magnetoactive plasmas

NASA Astrophysics Data System (ADS)

Nowak, S.; Orefice, A.

1994-05-01

In today's high frequency systems employed for plasma diagnostics, power heating, and current drive the behavior of the wave beams is appreciably affected by the self-diffraction phenomena due to their narrow collimation. In the present article the three-dimensional propagation of Gaussian beams in inhomogeneous and anisotropic media is analyzed, starting from a properly formulated dispersion relation. Particular attention is paid, in the case of electromagnetic electron cyclotron (EC) waves, to the toroidal geometry characterizing tokamak plasmas, to the power density evolution on the advancing wave fronts, and to the absorption features occurring when a beam crosses an EC resonant layer.

The R&D progress of 4 MW EAST-NBI high current ion source.

PubMed

Xie, Yahong; Hu, Chundong; Liu, Sheng; Xu, Yongjian; Liang, Lizhen; Xie, Yuanlai; Sheng, Peng; Jiang, Caichao; Liu, Zhimin

2014-02-01

A high current ion source, which consists of the multi-cusp bucket plasma generator and tetrode accelerator with multi-slot apertures, is developed and tested for the Experimental Advanced Superconducting Tokamak neutral beam injector. Three ion sources are tested on the test bed with arc power of 80 kW, beam voltage of 80 keV, and beam power of 4 MW. The arc regulation technology with Langmuir probes is employed for the long pulse operation of ion source, and the long pulse beam of 50 keV @ 15.5 A @ 100 s and 80 keV @ 52A @ 1s are extracted, respectively.

Arc discharge regulation of a megawatt hot cathode bucket ion source for the experimental advanced superconducting tokamak neutral beam injector

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

Xie Yahong; Hu Chundong; Liu Sheng

2012-01-15

Arc discharge of a hot cathode bucket ion source tends to be unstable what attributes to the filament self-heating and energetic electrons backstreaming from the accelerator. A regulation method, which based on the ion density measurement by a Langmuir probe, is employed for stable arc discharge operation and long pulse ion beam generation. Long pulse arc discharge of 100 s is obtained based on this regulation method of arc power. It establishes a foundation for the long pulse arc discharge of a megawatt ion source, which will be utilized a high power neutral beam injection device.

Arc discharge regulation of a megawatt hot cathode bucket ion source for the experimental advanced superconducting tokamak neutral beam injector.

PubMed

Xie, Yahong; Hu, Chundong; Liu, Sheng; Jiang, Caichao; Li, Jun; Liang, Lizhen

2012-01-01

Arc discharge of a hot cathode bucket ion source tends to be unstable what attributes to the filament self-heating and energetic electrons backstreaming from the accelerator. A regulation method, which based on the ion density measurement by a Langmuir probe, is employed for stable arc discharge operation and long pulse ion beam generation. Long pulse arc discharge of 100 s is obtained based on this regulation method of arc power. It establishes a foundation for the long pulse arc discharge of a megawatt ion source, which will be utilized a high power neutral beam injection device.

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.

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

Electric Tokamak Results and Plans

NASA Astrophysics Data System (ADS)

Taylor, R. J.; Gauvreau, J.-L.; Gourdain, P.-A.; Kissick, M. W.; Leboeuf, J.-N.; Schmitz, L. W.

1999-11-01

Initial plasmas have been obtained in ET (R=5 m, a=1 m) at 100 G previously with plasma currents up to 100 kA. Firsts full field plasma at 2.5 kG with the possibility of electrode and RF driven H-modes are expected in August 1999. The initial goal of the Electric Tokamak, ET(See Web Site for more details.), is to "drift lock" the thermal ions to the magnetic surfaces by rapid poloidal rotation ω_pol > ω_bounce. The required radial electric field is Er = Vi × B/A. ET physics is an outcome of the TTF effort of the last decade with a "unity beta vision" added to it by Cowley(S. C. Cowley, P. K. Kaw, R. S. Kelly, R. M. Kulsrud, \\underlinePhys. fluids B) 3 (1991) 2066. Classical heat transport is expected at 2 keV temperatures if rotation and beta can be combined, leading to the realization of an omnigeneous magnetic confinement zone(D. Palumbo, \\underlineNuovo Cimento) 53B (1967) 507.. The plasma startup experience and the scientific program plan will be discussed.

Scaling Relationships for ELM Diverter Heat Flux on DIII D

NASA Astrophysics Data System (ADS)

Peters, E. A.; Makowski, M. A.; Leonard, A. W.

2015-11-01

Edge Localized Modes (ELMs) are periodic plasma instabilities that occur during H-mode operation in tokamaks. Left unmitigated, these instabilities result in concentrated particle and heat fluxes at the divertor and stand to cause serious damage to the plasma facing components of tokamaks. The purpose of this research is to find scaling relationships that predict divertor heat flux due to ELMs based on plasma parameters at the time of instability. This will be accomplished by correlating characteristic ELM parameters with corresponding plasma measurements and analyzing the data for trends. One early assessment is the effect of the heat transmission coefficient ? on the in/out asymmetry of the calculated ELM heat fluxes. Using IR camera data, further assessments in this study will continue to emphasize in/out asymmetry in ELMs, as this has important implications for ITER operation. Work supported in part by the US DOE, DE-AC52-07NA27344, DE-FC02-04ER54698, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

Fast, multi-channel real-time processing of signals with microsecond latency using graphics processing units.

PubMed

Rath, N; Kato, S; Levesque, J P; Mauel, M E; Navratil, G A; Peng, Q

2014-04-01

Fast, digital signal processing (DSP) has many applications. Typical hardware options for performing DSP are field-programmable gate arrays (FPGAs), application-specific integrated DSP chips, or general purpose personal computer systems. This paper presents a novel DSP platform that has been developed for feedback control on the HBT-EP tokamak device. The system runs all signal processing exclusively on a Graphics Processing Unit (GPU) to achieve real-time performance with latencies below 8 μs. Signals are transferred into and out of the GPU using PCI Express peer-to-peer direct-memory-access transfers without involvement of the central processing unit or host memory. Tests were performed on the feedback control system of the HBT-EP tokamak using forty 16-bit floating point inputs and outputs each and a sampling rate of up to 250 kHz. Signals were digitized by a D-TACQ ACQ196 module, processing done on an NVIDIA GTX 580 GPU programmed in CUDA, and analog output was generated by D-TACQ AO32CPCI modules.

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

Stabilization of Tokamak Plasmas by the Addition of Nonaxisymmetric Coils

NASA Astrophysics Data System (ADS)

Reiman, Allan

2008-11-01

It has been recognized since the early days of the fusion program that stellarator coils can be used to stabilize current carrying, toroidal, magnetically confined plasmas.[1] More recently, it has been shown that the vertical mode in a tokamak can be stabilized by a relatively simple set of parallelogram-shaped, localized, nonaxisymmetric coils.[2] We show that by superposing sets of these parallelogram-shaped, nonaxisymmetric coils at different locations, it is possible to reproduce the coil current patterns for conventional stellarator coils as well as those for Furth-Hartman coils[3]. This allows us to gain insight into the physics of stabilization produced by various sets of nonaxisymmetric coils by analysis of the effect on stability of localized coils at differing locations. In particular, the relationship between the stabilization effect and the rotational transform generated by the nonaxisymmetric coils is clarified. [1] J. L. Johnson, C. R. Oberman, R. M. Kulsrud, and E. A. Frieman, Phys. Fluids 1, 281 (1958) [2] A. Reiman, Phys. Rev. Lett. 99, 135007, (2007). [3] H.P. Furth and C.W. Hartman, Phys. Fluids 11, 408 (1968).

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.

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.

Compact advanced extreme-ultraviolet imaging spectrometer for spatiotemporally varying tungsten spectra from fusion plasmas.

PubMed

Song, Inwoo; Seon, C R; Hong, Joohwan; An, Y H; Barnsley, R; Guirlet, R; Choe, Wonho

2017-09-01

A compact advanced extreme-ultraviolet (EUV) spectrometer operating in the EUV wavelength range of a few nanometers to measure spatially resolved line emissions from tungsten (W) was developed for studying W transport in fusion plasmas. This system consists of two perpendicularly crossed slits-an entrance aperture and a space-resolved slit-inside a chamber operating as a pinhole, which enables the system to obtain a spatial distribution of line emissions. Moreover, a so-called v-shaped slit was devised to manage the aperture size for measuring the spatial resolution of the system caused by the finite width of the pinhole. A back-illuminated charge-coupled device was used as a detector with 2048 × 512 active pixels, each with dimensions of 13.5 × 13.5 μm 2 . After the alignment and installation on Korea superconducting tokamak advanced research, the preliminary results were obtained during the 2016 campaign. Several well-known carbon atomic lines in the 2-7 nm range originating from intrinsic carbon impurities were observed and used for wavelength calibration. Further, the time behavior of their spatial distributions is presented.

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 increased q0 and slightly decreased q98. The magnetic helicity was found to change between -15% and +28% over an IRE.

INTRODUCTION: Award of the 2004 Hannes Alfvén Prize of the European Physical Society to J W Connor, R J Hastie and J B Taylor

NASA Astrophysics Data System (ADS)

Lister, Jo, Dr

2004-12-01

Jack Connor, Jim Hastie and Bryan Taylor The Hannes Alfvén Prize of the European Physical Society for Outstanding Contributions to Plasma Physics (2004) has been awarded to Jack Connor, Jim Hastie and Bryan Taylor `for their seminal contributions to a wide range of issues of fundamental importance to the success of magnetic confinement fusion, including: the development of gyro-kinetic theory; the prediction of the bootstrap current; dimensionless scaling laws; pressure-limiting instabilities, and micro-stability and transport theory'. Jack Connor, Jim Hastie and Bryan Taylor form one of the most successful teams of theoretical physicists in the history of magnetic confinement fusion. They have made important contributions individually, but their greatest discoveries have mostly been accomplished jointly, either in pairs or as a team involving all three. Their early work, in the 1960s, included the development of the gyro-kinetic theory for fine-scale plasma instabilities, which today forms the basis of the most advanced turbulence simulation codes in tokamak and stellarator research. The theoretical prediction of the bootstrap current, made in 1970-71 was not confirmed experimentally for over a decade but is now regarded as crucial to the success of the tokamak as a steady-state fusion power source. Their work on collisional transport also included the prediction of impurity ion accumulation, which is observed in internal transport barriers and is a key concern for long-pulse tokamak operation. The relativistic threshold for runaway electrons, identified in 1975, forms the basis of the most recent tokamak disruption mitigation schemes. In the late 1970s, the team developed the theory for ballooning instabilities, which provided an important ingredient in the `Troyon-Sykes' β-limit—an expression that is still used as a guide to the performance of tokamaks and in the design of ITER. Ballooning mode theory has also contributed to the understanding of instabilities in space plasmas such as magnetospheres and the solar corona. Finally, coming right up to date, the ballooning mode is thought to be a key ingredient in edge-localized modes (ELMs), which are a main issue for ITER, and ballooning stability is an important feature of modern stellarators. In the late 1970s and through the 1980s, the concept of dimensionless scaling laws was introduced and developed (following work by Kadomtsev), enabling scalings for transport coefficients to be derived without tackling all the details of the plasma turbulence. The same ideas are still used today to provide various constraints on confinement scaling laws, for example, on which the ITER design is largely based. The linear theory of toroidal drift waves was also developed by the team during this period, and into the 1990s. Key results on the role of shear damping in toroidal geometry, the identification of modes with extended radial correlation lengths, and the role of flow shear in reducing these correlation lengths (and hence transport) were deduced. All of these are key ideas that are often components in theoretical models for tokamak confinement and the generation of transport barriers. This laudation can only address a small number of the areas in which this formidable team of theoretical plasma physicists have made great contributions to our understanding of magnetically confined plasmas. It is appropriate and timely that their contributions are recognized as they approach the end of their careers.

Research on long pulse ECRH system of EAST in support of ITER

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

Wang, Xiaojie, E-mail: xjiew@ipp.ac.cn; Liu, Fukun; Shan, Jiafang

2015-12-10

Experimental Advanced Superconducting Tokamak (EAST), as a fully superconducting tokamak in China, aims to achieve high performance plasma under steady-state operation. To fulfill the physical objectives of EAST, a program of 4-MW long pulse electron cyclotron resonance heating and current drive (EC H&CD) system, which would offer greater flexibility for plasma shape and plasma stabilization has been launched on EAST since 2011. The system, composed of 4 gyrotrons with nominal 1MW output power and 1000s pulse length each, is designed with the feature of steerable power handling capabilities at 140 GHz, using second harmonic of the extraordinary mode(X2). The missions ofmore » the ECRH system are to provide plasma heating, current drive, plasma profile tailoring and control of magneto-hydrodynamic (MHD) instabilities. Presently, the first two 140-GHz 1-MW gyrotrons, provided by GYCOM and CPI, respectively, have been tested at long pulse operation. The tubes, the associated power supplies, cooling system, cryogenic plant, 2 transmission lines and an equatorial launcher are now installed at EAST. The power generated from each tube will be transmitted by an evacuated corrugated waveguide transmission line and injected into plasma from the low field side (radial port) through a front steering equatorial launcher. Considering the diverse applications of the EC system, the beam’s launch angles can be continuously varied with the optimized scanning range of over 30° in poloidal direction and ±25° in toroidal, as well as the polarization could be adjusted during the discharge by the orientations of a pair of polarizers in the transmission line to maintain the highest absorption for different operational scenarios. The commissioning of the first 2MW ECRH plant for EAST is under way. The design, R&D activities and recent progress of the long pulse 140-GHz ECRH system are presented in this paper. As the technological requirements for EAST ECRH have many similarities with ITER devices, the installation and experience of EAST ECRH system may provide valuable data for the ITER.« less

Compact Torus Injection Experiments on the H.I.T. teststand and the JFT-2M tokamak

NASA Astrophysics Data System (ADS)

Fukumoto, Naoyuki; Fujiwara, Makoto; Kuramoto, Keiji; Ageishi, Masaya; Nagata, Masayoshi; Uyama, Tadao; Ogawa, Hiroaki; Kasai, Satoshi; Hasegawa, Kouichi; Shibata, Takatoshi

1997-11-01

A spheromak-type compact torus (CT) acceleration and injection experiment has been carried out using the Himeji Institute of Technology Compact Torus Injector (HIT-CTI). We investigate the possibility of refueling, density control, current drive, and edge electric field control of tokamak plasmas by means of CT injection. The HIT-CTI produces a CT with a speed of 200 km/s and a density of 1× 10^21m-3. We have constructed new electrodes and power supplies, and will install the HIT-CTI on the JFT-2M tokamak at JAERI in Autumn 1997. The outer electrode serves as a common ground for both the formation bank (144μF, 20kV) and the acceleration bank (92.4μF, 40kV). If the external toroidal field of the tokamak is applied across the CT acceleration region, the CT kinetic energy might decrease during penetration into the field lines joining the inner and outer electrode. This could result in the CT not being able to reach the core of the tokamak plasma. Determining the optimum position of the inner electrode is one of the near term goals of this research. We will present magnetic probe, He-Ne interferometer and fast framing camera data from experiments at H.I.T., where a CT was accelerated into a transverse field. We will also present initial results from the operation of the HIT-CTI on the JFT-2M tokamak.

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

Burning plasma regime for Fussion-Fission Research Facility

NASA Astrophysics Data System (ADS)

Zakharov, Leonid E.

2010-11-01

The basic aspects of burning plasma regimes of Fusion-Fission Research Facility (FFRF, R/a=4/1 m/m, Ipl=5 MA, Btor=4-6 T, P^DT=50-100 MW, P^fission=80-4000 MW, 1 m thick blanket), which is suggested as the next step device for Chinese fusion program, are presented. The mission of FFRF is to advance magnetic fusion to the level of a stationary neutron source and to create a technical, scientific, and technology basis for the utilization of high-energy fusion neutrons for the needs of nuclear energy and technology. FFRF will rely as much as possible on ITER design. Thus, the magnetic system, especially TFC, will take advantage of ITER experience. TFC will use the same superconductor as ITER. The plasma regimes will represent an extension of the stationary plasma regimes on HT-7 and EAST tokamaks at ASIPP. Both inductive discharges and stationary non-inductive Lower Hybrid Current Drive (LHCD) will be possible. FFRF strongly relies on new, Lithium Wall Fusion (LiWF) plasma regimes, the development of which will be done on NSTX, HT-7, EAST in parallel with the design work. This regime will eliminate a number of uncertainties, still remaining unresolved in the ITER project. Well controlled, hours long inductive current drive operation at P^DT=50-100 MW is predicted.

Signatures of the electron saddle swaps mechanism in the photon spectra following charge-exchange collisions

NASA Astrophysics Data System (ADS)

Otranto, Sebastian

2014-10-01

During the last few years, several experimental and theoretical studies have focused on state selective charge exchange processes between charged ions and alkali metals. These data are of particular importance for the tokamak nuclear fusion reactor program, since diagnostics on the plasma usually rely on charge-exchange spectroscopy. In this sense, alkali metals, have been proposed as potential alternatives to excited hydrogen/deuterium for which laboratory experiments are not feasible at present. In this talk, we present our recent work involving ion collisions with alkali metals. Oscillatory structures in the angular differential charge-exchange cross sections obtained using the MOTRIMS technique are correctly described by classical trajectory Monte Carlo simulations. These oscillations are found to originate from the number of swaps the electron undergoes around the projectile-target potential saddle before capture takes place and are very prominent at impact energies below 10 keV/amu. Moreover, cross sections of higher order of differentiability also indicate that the swaps leave distinctive signatures in the (n,l)-state selective cross sections and in the photon line emission cross sections. Oscillatory structures for the x-ray hardness ratio parameter are also predicted. In collaboration with Ronnie Hoekstra, Zernike Institute for Advanced Materials, University of Groningen and Ronald Olson, Department of Physics, Missouri University of Science and Technology.

In-vessel tritium retention and removal in ITER

NASA Astrophysics Data System (ADS)

Federici, G.; Anderl, R. A.; Andrew, P.; Brooks, J. N.; Causey, R. A.; Coad, J. P.; Cowgill, D.; Doerner, R. P.; Haasz, A. A.; Janeschitz, G.; Jacob, W.; Longhurst, G. R.; Nygren, R.; Peacock, A.; Pick, M. A.; Philipps, V.; Roth, J.; Skinner, C. H.; Wampler, W. R.

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER - and more generically in any other next-step experimental fusion facility fuelled with tritium - the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

Analysis of Radiation Transport Due to Activated Coolant in the ITER Neutral Beam Injection Cell

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

Royston, Katherine; Wilson, Stephen C.; Risner, Joel M.

Detailed spatial distributions of the biological dose rate due to a variety of sources are required for the design of the ITER tokamak facility to ensure that all radiological zoning limits are met. During operation, water in the Integrated loop of Blanket, Edge-localized mode and vertical stabilization coils, and Divertor (IBED) cooling system will be activated by plasma neutrons and will flow out of the bioshield through a complex system of pipes and heat exchangers. This paper discusses the methods used to characterize the biological dose rate outside the tokamak complex due to 16N gamma radiation emitted by the activatedmore » coolant in the Neutral Beam Injection (NBI) cell of the tokamak building. Activated coolant will enter the NBI cell through the IBED Primary Heat Transfer System (PHTS), and the NBI PHTS will also become activated due to radiation streaming through the NBI system. To properly characterize these gamma sources, the production of 16N, the decay of 16N, and the flow of activated water through the coolant loops were modeled. The impact of conservative approximations on the solution was also examined. Once the source due to activated coolant was calculated, the resulting biological dose rate outside the north wall of the NBI cell was determined through the use of sophisticated variance reduction techniques. The AutomateD VAriaNce reducTion Generator (ADVANTG) software implements methods developed specifically to provide highly effective variance reduction for complex radiation transport simulations such as those encountered with ITER. Using ADVANTG with the Monte Carlo N-particle (MCNP) radiation transport code, radiation responses were calculated on a fine spatial mesh with a high degree of statistical accuracy. In conclusion, advanced visualization tools were also developed and used to determine pipe cell connectivity, to facilitate model checking, and to post-process the transport simulation results.« less

A dynamic monitoring approach for the surface morphology evolution measurement of plasma facing components by means of speckle interferometry

NASA Astrophysics Data System (ADS)

Wang, Hongbei; Cui, Xiaoqian; Feng, Chunlei; Li, Yuanbo; Zhao, Mengge; Luo, Guangnan; Ding, Hongbin

2017-11-01

Plasma Facing Components (PFCs) in a magnetically confined fusion plasma device will be exposed to high heat load and particle fluxes, and it would cause PFCs' surface morphology to change due to material erosion and redeposition from plasma wall interactions. The state of PFCs' surface condition will seriously affect the performance of long-pulse or steady state plasma discharge in a tokamak; it will even constitute an enormous threat to the operation and the safety of fusion plasma devices. The PFCs' surface morphology evolution measurement could provide important information about PFCs' real-time status or damage situation and it would help to a better understanding of the plasma wall interaction process and mechanism. Meanwhile through monitoring the distribution of dust deposition in a tokamak and providing an upper limit on the amount of loose dust, the PFCs' surface morphology measurement could indirectly contribute to keep fusion operational limits and fusion device safety. Aiming at in situ dynamic monitoring PFCs' surface morphology evolution, a laboratory experimental platform DUT-SIEP (Dalian University of Technology-speckle interferometry experimental platform) based on the speckle interferometry technique has been constructed at Dalian University of Technology (DUT) in China. With directional specific designing and focusing on the real detection condition of EAST (Experimental Advanced Superconducting Tokamak), the DUT-SIEP could realize a variable measurement range, widely increased from 0.1 μm to 300 μm, with high spatial resolution (<1 mm) and ultra-high time resolution (<2 s for EAST measuring conditions). Three main components of the DUT-SIEP are all integrated and synchronized by a time schedule control and data acquisition terminal and coupled with a three-dimensional phase unwrapping algorithm, the surface morphology information of target samples can be obtained and reconstructed in real-time. A local surface morphology of the real divertor tiles adopted from EAST has been measured, and the feasibility and reliability of this new experimental platform have been demonstrated.

Analysis of Radiation Transport Due to Activated Coolant in the ITER Neutral Beam Injection Cell

DOE PAGES

Royston, Katherine; Wilson, Stephen C.; Risner, Joel M.; ...

2017-07-26

Detailed spatial distributions of the biological dose rate due to a variety of sources are required for the design of the ITER tokamak facility to ensure that all radiological zoning limits are met. During operation, water in the Integrated loop of Blanket, Edge-localized mode and vertical stabilization coils, and Divertor (IBED) cooling system will be activated by plasma neutrons and will flow out of the bioshield through a complex system of pipes and heat exchangers. This paper discusses the methods used to characterize the biological dose rate outside the tokamak complex due to 16N gamma radiation emitted by the activatedmore » coolant in the Neutral Beam Injection (NBI) cell of the tokamak building. Activated coolant will enter the NBI cell through the IBED Primary Heat Transfer System (PHTS), and the NBI PHTS will also become activated due to radiation streaming through the NBI system. To properly characterize these gamma sources, the production of 16N, the decay of 16N, and the flow of activated water through the coolant loops were modeled. The impact of conservative approximations on the solution was also examined. Once the source due to activated coolant was calculated, the resulting biological dose rate outside the north wall of the NBI cell was determined through the use of sophisticated variance reduction techniques. The AutomateD VAriaNce reducTion Generator (ADVANTG) software implements methods developed specifically to provide highly effective variance reduction for complex radiation transport simulations such as those encountered with ITER. Using ADVANTG with the Monte Carlo N-particle (MCNP) radiation transport code, radiation responses were calculated on a fine spatial mesh with a high degree of statistical accuracy. In conclusion, advanced visualization tools were also developed and used to determine pipe cell connectivity, to facilitate model checking, and to post-process the transport simulation results.« less

Plasma Inter-Particle and Particle-Wall Interactions

NASA Astrophysics Data System (ADS)

Patino, Marlene Idy

An improved understanding of plasma inter-particle and particle-wall interactions is critical to the advancement of plasma devices used for space electric propulsion, fusion, high-power communications, and next-generation energy systems. Two interactions of particular importance are (1) ion-atom collisions in the plasma bulk and (2) secondary electron emission from plasma-facing materials. For ion-atom collisions, interactions between fast ions and slow atoms are commonly dominated by charge-exchange and momentum-exchange collisions that are important to understanding the performance and behavior of many plasma devices. To investigate this behavior, this work developed a simple, well-characterized experiment that accurately measures the effects of high energy xenon ions incident on a background of xenon neutral atoms. By comparing these results to both analytical and computational models of ion-atom interactions, we discovered the importance of (1) accurately treating the differential cross-sections for momentum-exchange and charge-exchange collisions over all neutral background pressures, and (2) commonly overlooked interactions, including ion-induced electron emission and neutral-neutral ionization collisions, at high pressures. Data provide vital information on the angular scattering distributions of charge-exchange and momentum-exchange ions at 1.5 keV relevant for ion thrusters, and serve as canonical data for validation of plasma models. This work also investigates electron-induced secondary electron emission behavior relevant to materials commonly considered for plasma thrusters, fusion systems, and many other plasma devices. For such applications, secondary electron emission can alter the sheath potential, which can significantly affect device performance and life. Secondary electron emission properties were measured for materials that are critical to the efficient operation of many plasma devices, including: graphite (for tokamaks, ion thrusters, and traveling wave tubes), lithium (for tokamak walls), tungsten (the most promising material for future tokamaks such as ITER), and nickel (for plasma-enhanced chemistry). Measurements were made for incident electron energies up to 1.5 keV and angles between 0 and 78°. The most significant results from these measurements are as follows: (1) first-ever measurements of naturally-forming tungsten fuzz show a more than 40% reduction in secondary electron emission and an independence on incidence angle; (2) original measurements of lithium oxide show a 2x and 6x increase in secondary electron emission for 17% and 100% oxidation; and (3) unique measurements of Ni(110) single crystal show extrema in secondary electron emission when incidence angle is varied and an up to 36% increase at 0° over polycrystalline nickel. Each of these results are important discoveries for improving plasma devices. For example, from (1), the growth of tungsten fuzz in tokamaks is desirable for minimizing adverse secondary electron emission effects. From (2), the opposite is true for tokamaks with lithium coatings which are oxidized by typical residual gases. From (3), secondary electron emission from Ni(110) catalysts in plasma-enhanced chemistry may facilitate further reactions.

Homoclinic tangle of the ideal separatrix in the DIII-D tokamak from (30, 10) + (40, 10) perturbation

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

Punjabi, Alkesh

Trajectories of magnetic field lines are a 1½ degree of freedom Hamiltonian system. The perturbed separatrix in a divertor tokamak is radically different from the unperturbed one. This is because magnetic asymmetries cause the separatrix to form extremely complicated structures called homoclinic tangles. The shape of flux surfaces in the edge region of divertor tokamaks such as the DIII (J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)) is fundamentally different from near-circular. Recently, a new method is developed to calculate the homoclinic tangle and lobes of the separatrix of divertor tokamaks in physical space (A. Punjabimore » and A. Boozer, Phys. Lett. A 378, 2410 (2014)). This method is based on three elements: preservation of the two invariants—symplectic and topological neighborhood—and a new set of canonical coordinates called the natural canonical coordinates. The very complicated shape of edge surfaces can be represented very accurately and very realistically in these new coordinates (A. Punjabi and H. Ali, Phys. Plasmas 15, 122502 (2008); A. Punjabi, Nucl. Fusion 49, 115020 (2009)). A symplectic map in the new coordinates can advance the separatrix manifold forward and backward in time. Every time the two manifolds meet in a fixed poloidal plane, they intersect and form homoclinic tangle to preserve the two invariants. The new coordinates can be mapped to physical space and the dynamical evolution of the homoclinic tangle can be seen and pictured in physical space. Here, the new method is applied to the DIII-D tokamak to study the basic features of the homoclinic tangle of the unperturbed separatrix from two Fourier components, which represent the peeling-ballooning modes of equal amplitude and no radial dependence, and the results are analyzed. Homoclinic tangle has a very complicated structure and becomes extremely complicated for as the lines take more toroidal turns, especially near the X-point. Homoclinic tangle is the most complicated near the X-point and forms the largest lobes there. On average, the field lines cover a distance of about 9 m per turn. Poloidal rotation of the lines has large gradients in the poloidal direction. The average normal displacement of the lines on the separatrix varies from 5 mm to 7 cm. Average outward displacement of the lines is considerably larger than the inward displacement; however, on the average more lines are displaced inside than outside of the separatrix. The field line diffusion normal to the separatrix has extremely wide variation and very large poloidal gradients. Half of all the lines are lost in less than 6 turns. Complicated electric potentials will be required to maintain the neutrality of the plasma, and the E × B drifts from these fields can modify plasma confinement and influence the edge physics (A. Punjabi and A. Boozer, Phys. Lett. A 378, 2410 (2014))« less

Homoclinic tangle of the ideal separatrix in the DIII-D tokamak from (30, 10) + (40, 10) perturbation

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh

2014-12-01

Trajectories of magnetic field lines are a 1½ degree of freedom Hamiltonian system. The perturbed separatrix in a divertor tokamak is radically different from the unperturbed one. This is because magnetic asymmetries cause the separatrix to form extremely complicated structures called homoclinic tangles. The shape of flux surfaces in the edge region of divertor tokamaks such as the DIII (J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)) is fundamentally different from near-circular. Recently, a new method is developed to calculate the homoclinic tangle and lobes of the separatrix of divertor tokamaks in physical space (A. Punjabi and A. Boozer, Phys. Lett. A 378, 2410 (2014)). This method is based on three elements: preservation of the two invariants—symplectic and topological neighborhood—and a new set of canonical coordinates called the natural canonical coordinates. The very complicated shape of edge surfaces can be represented very accurately and very realistically in these new coordinates (A. Punjabi and H. Ali, Phys. Plasmas 15, 122502 (2008); A. Punjabi, Nucl. Fusion 49, 115020 (2009)). A symplectic map in the new coordinates can advance the separatrix manifold forward and backward in time. Every time the two manifolds meet in a fixed poloidal plane, they intersect and form homoclinic tangle to preserve the two invariants. The new coordinates can be mapped to physical space and the dynamical evolution of the homoclinic tangle can be seen and pictured in physical space. Here, the new method is applied to the DIII-D tokamak to study the basic features of the homoclinic tangle of the unperturbed separatrix from two Fourier components, which represent the peeling-ballooning modes of equal amplitude and no radial dependence, and the results are analyzed. Homoclinic tangle has a very complicated structure and becomes extremely complicated for as the lines take more toroidal turns, especially near the X-point. Homoclinic tangle is the most complicated near the X-point and forms the largest lobes there. On average, the field lines cover a distance of about 9 m per turn. Poloidal rotation of the lines has large gradients in the poloidal direction. The average normal displacement of the lines on the separatrix varies from 5 mm to 7 cm. Average outward displacement of the lines is considerably larger than the inward displacement; however, on the average more lines are displaced inside than outside of the separatrix. The field line diffusion normal to the separatrix has extremely wide variation and very large poloidal gradients. Half of all the lines are lost in less than 6 turns. Complicated electric potentials will be required to maintain the neutrality of the plasma, and the E × B drifts from these fields can modify plasma confinement and influence the edge physics (A. Punjabi and A. Boozer, Phys. Lett. A 378, 2410 (2014)).

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.

Superconducting magnet development for tokamaks and mirrors: a technical assessment

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

Laverick, C.; Jacobs, R. B.; Boom, R. W.

1977-11-01

The role of superconducting magnets in Magnetic Fusion Energy Research and Development is assessed from a consideration of program plans and schedules, the present status of the programs and the research and development suggestions arising from recent studies and workshops. A principal conclusion is that the large superconducting magnet systems needed for commercial magnetic fusion reactors can be constructed. However such magnets working under severe conditions, with increasingly stringent reliability, safety and cost restrictions can never be built unless experience is first gained in a number of important installations designed to prove physics and technology steps on the way tomore » commercial power demonstration. The immediate problem is to design a technology program in the absence of definite device needs and specifications, giving a priority weighting to the multiplicity of good, high quality development program suggestions when all proposals cannot be supported.« less

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.

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.

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.

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.

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 needed particularly for the 3D divertor tokamak configuration.« less

Conversion of magnetic energy to runaway kinetic energy during the termination of runaway current on the J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Dai, A. J.; Chen, Z. Y.; Huang, D. W.; Tong, R. H.; Zhang, J.; Wei, Y. N.; Ma, T. K.; Wang, X. L.; Yang, H. Y.; Gao, H. L.; Pan, Y.; the J-TEXT Team

2018-05-01

A large number of runaway electrons (REs) with energies as high as several tens of mega-electron volt (MeV) may be generated during disruptions on a large-scale tokamak. The kinetic energy carried by REs is eventually deposited on the plasma-facing components, causing damage and posing a threat on the operation of the tokamak. The remaining magnetic energy following a thermal quench is significant on a large-scale tokamak. The conversion of magnetic energy to runaway kinetic energy will increase the threat of runaway electrons on the first wall. The magnetic energy dissipated inside the vacuum vessel (VV) equals the decrease of initial magnetic energy inside the VV plus the magnetic energy flowing into the VV during a disruption. Based on the estimated magnetic energy, the evolution of magnetic-kinetic energy conversion are analyzed through three periods in disruptions with a runaway current plateau.

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

ETF system code: composition and applications

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

Reid, R.L.; Wu, K.F.

1980-01-01

A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies,more » such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system.« less

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

NASA Astrophysics Data System (ADS)

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

2018-03-01

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

Destruction of tungsten limiters in the T-10 Tokamak under high plasma heat loads

NASA Astrophysics Data System (ADS)

Grashin, S. A.; Arkhipov, I. I.; Budaev, V. P.; Giniyatulin, R. N.; Karpov, A. V.; Klyuchnikov, L. A.; Krupin, V. A.; Litunovskiy, N. V.; Masul, I. V.; Makhankov, F. N.; Martynenko, Yu V.; Sarytchev, D. V.; Solomatin, R. Yu; Khimchenko, L. N.

2017-10-01

Tungsten limiters were tested in the T-10 tokamak. The limiters were made from the ITER-grade WMP “POLEMA” tungsten. The influence of the edge tokamak plasma on tungsten limiters leads to significant cracking of tungsten. The heat load of up to 2 MW · m-2 leads to the micro-crack development at the grain boundaries accompanied by the loss of grains. The heat loads that exceed 5 MW · m-2 lead to the macro crack development. Under the present T-10 tokamak conditions, the heat and particle fluxes in the edge plasma lead to the significant destruction of tungsten limiters during the experimental campaign. During the disruption and runaway electron formation, extreme heat loads of more than 1 GW/m2 cause strong melting of tungsten on the inner and outer part of the ring limiter.

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.

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.

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.

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

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.

Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak

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

Maurya, Gulab Singh; Kumar, Rohit; Rai, Awadhesh Kumar, E-mail: awadheshkrai@rediffmail.com

2015-12-15

In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known asmore » “back collection method” to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.« 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.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Footprint of Field lines

NASA Astrophysics Data System (ADS)

Figgins, Montoya; Ali, Halima; Punjabi, Alkesh

2000-10-01

We are working with the Simple Map^1 to find the footprint of field lines on the diverter plate in a single-null tokamak. Footprint of a field line is the position of the line when it escapes across the divertor plate. The Simple Map represents the magnetic field in a single-null divertor tokamak. The path of a field line is given by the equations: X_n+1=X_n-kY_n(1-Y_n) and Y_n+1=Y_n+kX_n+1. In order to find the footprint, we must first find the last good surface which is Y=0.997135768 and X=0. The value of k is fixed at 0.6. The starting values X0 are fixed at X_0=0. We use 10,000 points between the last good surface and the X-point. The X-point is located at (0,1). We also use the Continuous Analog of the Simple Map given by the equations: X(φ)=X_0-kY0 (1-Y_0)φ and Y(φ)=Y_0+kX(φ)φ. This will tell us what the (φ,X) is which represents the field lines crossing the divertor plate. The divertor plate is located at Y=1. When graphed, the footprint of field lines looks like the rings of Saturn. This work is supported by US DOES OFES. Ms. Montoya Figgins is HU CFRT Summer Fusion High School Scholar from E. E. Smith High School in North Carolina. She is supported by NASA under its NASA SHARP Plus Program. 1. Punjabi A, Verma A, and Boozer A, Phys Rev Lett, 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994)

An intelligent remote control system for ECEI on EAST

NASA Astrophysics Data System (ADS)

Chen, Dongxu; Zhu, Yilun; Zhao, Zhenling; Qu, Chengming; Liao, Wang; Xie, Jinlin; Liu, Wandong

2017-08-01

An intelligent remote control system based on a power distribution unit (PDU) and Arduino has been designed for the electron cyclotron emission imaging (ECEI) system on Experimental Advanced Superconducting Tokamak (EAST). This intelligent system has three major functions: ECEI system reboot, measurement region adjustment and signal amplitude optimization. The observation region of ECEI can be modified for different physics proposals by remotely tuning the optical and electronics systems. Via the remote adjustment of the attenuation level, the ECEI intermediate frequency signal amplitude can be efficiently optimized. The remote control system provides a feasible and reliable solution for the improvement of signal quality and the efficiency of the ECEI diagnostic system, which is also valuable for other diagnostic systems.

Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

PubMed

McAdams, R

2014-02-01

In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

Design of the klystron filament power supply control system for EAST LHCD

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

Wu, Zege; Wang, Mao; Hu, Huaichuan

A filament is a critical component of the klystron used to heat the cathode. There are totally 44 klystrons in experimental advanced superconducting tokamak (EAST) lower hybrid current drive (LHCD) systems. All klystron filaments are powered by AC power suppliers through isolated transformers. In order to achieve better klystron preheat, a klystron filament power supply control system is designed to obtain the automatic control of all filament power suppliers. Klystron filament current is measured by PLC and the interlock between filament current and klystron high voltage system is also implemented. This design has already been deployed in two LHCD systemsmore » and proves feasible completely.« less

Physics-model-based nonlinear actuator trajectory optimization and safety factor profile feedback control for advanced scenario development in DIII-D

DOE PAGES

Barton, Justin E.; Boyer, Mark D.; Shi, Wenyu; ...

2015-07-30

DIII-D experimental results are reported to demonstrate the potential of physics-model-based safety factor profile control for robust and reproducible sustainment of advanced scenarios. In the absence of feedback control, variability in wall conditions and plasma impurities, as well as drifts due to external disturbances, can limit the reproducibility of discharges with simple pre-programmed scenario trajectories. The control architecture utilized is a feedforward + feedback scheme where the feedforward commands are computed off-line and the feedback commands are computed on-line. In this work, firstly a first-principles-driven (FPD), physics-based model of the q profile and normalized beta (β N) dynamics is embeddedmore » into a numerical optimization algorithm to design feedforward actuator trajectories that sheer the plasma through the tokamak operating space to reach a desired stationary target state that is characterized by the achieved q profile and β N. Good agreement between experimental results and simulations demonstrates the accuracy of the models employed for physics-model-based control design. Secondly, a feedback algorithm for q profile control is designed following a FPD approach, and the ability of the controller to achieve and maintain a target q profile evolution is tested in DIII-D high confinement (H-mode) experiments. The controller is shown to be able to effectively control the q profile when β N is relatively close to the target, indicating the need for integrated q profile and β N control to further enhance the ability to achieve robust scenario execution. Furthermore, the ability of an integrated q profile + β N feedback controller to track a desired target is demonstrated through simulation.« less

Current status and future R&D for reduced-activation ferritic/martensitic steels

NASA Astrophysics Data System (ADS)

Hishinuma, A.; Kohyama, A.; Klueh, R. L.; Gelles, D. S.; Dietz, W.; Ehrlich, K.

1998-10-01

International research and development programs on reduced-activation ferritic/martensitic steels, the primary candidate-alloys for a DEMO fusion reactor and beyond, are briefly summarized, along with some information on conventional steels. An International Energy Agency (IEA) collaborative test program to determine the feasibility of reduced-activation ferritic/martensitic steels for fusion is in progress and will be completed within this century. Baseline properties including typical irradiation behavior for Fe-(7-9)%Cr reduced-activation ferritic steels are shown. Most of the data are for a heat of modified F82H steel, purchased for the IEA program. Experimental plans to explore possible problems and solutions for fusion devices using ferromagnetic materials are introduced. The preliminary results show that it should be possible to use a ferromagnetic vacuum vessel in tokamak devices.

Prediction of nonlinear evolution character of energetic-particle-driven instabilities

DOE PAGES

Duarte, Vinicius N.; Berk, H. L.; Gorelenkov, N. N.; ...

2017-03-17

A general criterion is proposed and found to successfully predict the emergence of chirping oscillations of unstable Alfvénic eigenmodes in tokamak plasma experiments. The model includes realistic eigenfunction structure, detailed phase-space dependences of the instability drive, stochastic scattering and the Coulomb drag. The stochastic scattering combines the effects of collisional pitch angle scattering and micro-turbulence spatial diffusion. Furthermore, the latter mechanism is essential to accurately identify the transition between the fixed-frequency mode behavior and rapid chirping in tokamaks and to resolve the disparity with respect to chirping observation in spherical and conventional tokamaks.

Prediction of nonlinear evolution character of energetic-particle-driven instabilities

NASA Astrophysics Data System (ADS)

Duarte, V. N.; Berk, H. L.; Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; Nazikian, R.; Pace, D. C.; Podestà, M.; Tobias, B. J.; Van Zeeland, M. A.

2017-05-01

A general criterion is proposed and found to successfully predict the emergence of chirping oscillations of unstable Alfvénic eigenmodes in tokamak plasma experiments. The model includes realistic eigenfunction structure, detailed phase-space dependences of the instability drive, stochastic scattering and the Coulomb drag. The stochastic scattering combines the effects of collisional pitch angle scattering and micro-turbulence spatial diffusion. The latter mechanism is essential to accurately identify the transition between the fixed-frequency mode behavior and rapid chirping in tokamaks and to resolve the disparity with respect to chirping observation in spherical and conventional tokamaks.

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.

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.}

A fluid model for the edge pressure pedestal height and width in tokamaks based on the transport constraint of particle, energy, and momentum balance

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

Stacey, W. M., E-mail: weston.stacey@nre.gatech.edu

2016-06-15

A fluid model for the tokamak edge pressure profile required by the conservation of particles, momentum and energy in the presence of specified heating and fueling sources and electromagnetic and geometric parameters has been developed. Kinetics effects of ion orbit loss are incorporated into the model. The use of this model as a “transport” constraint together with a “Peeling-Ballooning (P-B)” instability constraint to achieve a prediction of edge pressure pedestal heights and widths in future tokamaks is discussed.

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.

Li Experiments at the Tokamak T-11M Toward PFC Concept of Steady State Tokamak-Reactor

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2009-11-01

As practical method of using a liquid lithium as a renewable plasma-facing component (PCF) for steady state tokamak-reactor the concept of lithium emitter-collector is considered [1]. It is based on lithium filled capillary porous system proposed by V.A. Evtikhin et al. (1996). The lithium circulation process consists of four steps: (1) Li emission from the PFC emitter into the plasma; (2) plasma boundary cooling by non-coronal Li radiation; (3) Li ion capture by the collector (before they are lost to the tokamak chamber wall); (4) Li return from the collector to the emitter. T-11M tokamak experiments have used three local rail limiters made from lithium, molybdenum and graphite as lithium collectors. The lithium behavior was studied by analysis of the witness samples, and by a mobile graphite probe. The key findings are: (1) lithium collection on the ion side of the lithium limiter is 2-3 times larger than on the electron side; (2) total efficiency of Li collection integrated over all three rail limiters can reach 50-70% of the lithium emission during the discharge pulse, while the theoretical limit is about 90%. [1] S.V. Mirnov, J. Nucl. Mat., 390-391, 876 (2009).

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.

Hiro and Evans currents in Vertical Disruption Event

NASA Astrophysics Data System (ADS)

Zakharov, Leonid; Xujing Li Team; Sergei Galkin Team

2014-10-01

The notion of Tokamak Magneto-Hydrodynamics (TMHD), which explicitly reflects the anisotropy of a high temperature tokamak plasma is introduced. The set of TMHD equations is formulated for simulations of macroscopic plasma dynamics and disruptions in tokamaks. Free from the Courant restriction on the time step, this set of equations is appropriate for high performance plasmas and does not require any extension of the MHD plasma model. At the same time, TMHD requires the use of magnetic field aligned numerical grids. The TMHD model was used for creation of theory of the Wall Touching Kink and Vertical Modes (WTKM and WTVM), prediction of Hiro and Evans currents, design of an innovative diagnostics for Hiro current measurements, installed on EAST device. While Hiro currents have explained the toroidal asymmetry in the plasma current measurements in JET disruptions, the Evans currents explain the tile current measurements in tokamaks. The recently developed Vertical Disruption Code (VDE) have demonstrated 5 regimes of VDE and confirmed the generation of both Hiro and Evans currents. The results challenge the 24 years long misinterpretation of the tile currents in tokamaks as ``halo'' currents, which were a product of misuse of equilibrium reconstruction for VDE. This work is supported by US DoE Contract No. DE-AC02-09-CH1146.

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.

The Pegasus-Upgrade Experiment

NASA Astrophysics Data System (ADS)

Fonck, R. J.; Bongard, M. W.; Barr, J. L.; Frerichs, H. G.; Lewicki, B. T.; Reusch, J. A.; Schmitz, O.; Winz, G. R.

2015-11-01

Tokamak operation at near-unity aspect ratio provides access to advanced tokamak physics at modest parameters. High plasma current is accessible at very low toroidal field. This offers H-mode performance at Te levels that allow use of electrostatic and magnetic probe arrays through the edge pedestal region into the plasma core. An upgrade to the Pegasus ST is planned to exploit these features and pursue unique studies in three areas: local measurements of pedestal and ELM dynamics at Alfvenic timescales; direct measurement of the local plasma response to application of 3D magnetic perturbations with high spectral flexibility; and extension of Local Helicity Injection for nonsolenoidal startup to NSTX-U-relevant confinement and stability regimes. Significant but relatively low-cost upgrades to the facility are proposed: a new centerstack with larger solenoid and 2x the number of toroidal field conductors; a new TF power supply and conversion of the 200 MVA OH power supply to a cascaded multilevel inverter configuration; and installation of an extensive 3D-magnetic perturbation coil system for ELM mitigation and suppression studies. The upgraded facility will provide 0.3 MA plasmas with pulse lengths of 50-100 msec flattop, aspect ratio <1.25, and toroidal field up to 0.4 T. These research activities will be integrated into related efforts on DIII-D and NSTX-U. Work supported by US DOE grant DE-FG02-96ER54375.

First measurement of the edge charge exchange recombination spectroscopy on EAST tokamak

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

Li, Y. Y., E-mail: liyy@ipp.ac.cn; Fu, J.; Jiang, D.

2016-11-15

An edge toroidal charge exchange recombination spectroscopy (eCXRS) diagnostic, based on a heating neutral beam injection (NBI), has been deployed recently on the Experimental Advanced Superconducting Tokamak (EAST). The eCXRS, which aims to measure the plasma ion temperature and toroidal rotation velocity in the edge region simultaneously, is a complement to the exiting core CXRS (cCXRS). Two rows with 32 fiber channels each cover a radial range from ∼2.15 m to ∼2.32 m with a high spatial resolution of ∼5-7 mm. Charge exchange emission of Carbon VI CVI at 529.059 nm induced by the NBI is routinely observed, but canmore » be tuned to any interested wavelength in the spectral range from 400 to 700 nm. Double-slit fiber bundles increase the number of channels, the fibers viewing the same radial position are binned on the CCD detector to improve the signal-to-noise ratio, enabling shorter exposure time down to 5 ms. One channel is connected to a neon lamp, which provides the real-time wavelength calibration on a shot-to-shot basis. In this paper, an overview of the eCXRS diagnostic on EAST is presented and the first results from the 2015 experimental campaign will be shown. Good agreements in ion temperature and toroidal rotation are obtained between the eCXRS and cCXRS systems.« less

Effect of pedestal fluctuation on ELM frequency in the EAST tokamak

NASA Astrophysics Data System (ADS)

Zhong, F. B.; Zhang, T.; Liu, Z. X.; Qu, H.; Liu, H. Q.; Li, G. Q.; Liu, Y.; Gao, W.; Duan, Y. M.; Yang, Y.; Zeng, L.; Xiang, H. M.; Geng, K. N.; Wen, F.; Zhang, S. B.; Gao, X.; the EAST Team

2018-05-01

The dependence of ELM frequency on heating power has been studied on the Experimental Advanced Superconducting Tokamak (EAST). It is found that the ELM frequency (f ELM) generally increases with the power through separatrix (P sep), indicating type-I ELM in these plasmas. However, there are two data points, named ‘abnormal ELM’ in the present paper, which have much lower f ELM than the ‘normal ELM’, while both types of ELM have similar ELM energy losses. The ‘abnormal ELM’ occurs at a phase with increased radiation power due to metal impurity influx events. The increased radiation power cannot explain the much lower f ELM for ‘abnormal ELM’, since the reduction of P sep is weaker than proportional to the observed reduction of the ELM frequency. The ‘abnormal ELM’ feature can be attributed to the enhanced amplitude of a coherent mode in the pedestal region. Comparing the pedestal evolutions for the two types of ELM with similar separatrix power P sep, it is actually found that the more pronounced pedestal coherent mode in the plasma with ‘abnormal ELM’ leads to a slower pressure pedestal recovery during the inter-ELM phase. This experimental result implies that the physical mechanism for ‘abnormal ELM’ is that the more pronounced pedestal fluctuation induces larger outward transport, slows down the pedestal evolution and leads to longer inter-ELM phase, i.e. lower ELM frequency.

Overview of results from the MST reversed field pinch experiment

NASA Astrophysics Data System (ADS)

Sarff, J. S.; Almagri, A. F.; Anderson, J. K.; Borchardt, M.; Carmody, D.; Caspary, K.; Chapman, B. E.; Den Hartog, D. J.; Duff, J.; Eilerman, S.; Falkowski, A.; Forest, C. B.; Goetz, J. A.; Holly, D. J.; Kim, J.-H.; King, J.; Ko, J.; Koliner, J.; Kumar, S.; Lee, J. D.; Liu, D.; Magee, R.; McCollam, K. J.; McGarry, M.; Mirnov, V. V.; Nornberg, M. D.; Nonn, P. D.; Oliva, S. P.; Parke, E.; Reusch, J. A.; Sauppe, J. P.; Seltzman, A.; Sovinec, C. R.; Stephens, H.; Stone, D.; Theucks, D.; Thomas, M.; Triana, J.; Terry, P. W.; Waksman, J.; Bergerson, W. F.; Brower, D. L.; Ding, W. X.; Lin, L.; Demers, D. R.; Fimognari, P.; Titus, J.; Auriemma, F.; Cappello, S.; Franz, P.; Innocente, P.; Lorenzini, R.; Martines, E.; Momo, B.; Piovesan, P.; Puiatti, M.; Spolaore, M.; Terranova, D.; Zanca, P.; Belykh, V.; Davydenko, V. I.; Deichuli, P.; Ivanov, A. A.; Polosatkin, S.; Stupishin, N. V.; Spong, D.; Craig, D.; Harvey, R. W.; Cianciosa, M.; Hanson, J. D.

2013-10-01

An overview of recent results from the MST programme on physics important for the advancement of the reversed field pinch (RFP) as well as for improved understanding of toroidal magnetic confinement more generally is reported. Evidence for the classical confinement of ions in the RFP is provided by analysis of impurity ions and energetic ions created by 1 MW neutral beam injection (NBI). The first appearance of energetic-particle-driven modes by NBI in a RFP plasma is described. MST plasmas robustly access the quasi-single-helicity state that has commonalities to the stellarator and ‘snake’ formation in tokamaks. In MST the dominant mode grows to 8% of the axisymmetric field strength, while the remaining modes are reduced. Predictive capability for tearing mode behaviour has been improved through nonlinear, 3D, resistive magnetohydrodynamic computation using the measured resistivity profile and Lundquist number, which reproduces the sawtooth cycle dynamics. Experimental evidence and computational analysis indicates two-fluid effects, e.g., Hall physics and gyro-viscosity, are needed to understand the coupling of parallel momentum transport and current profile relaxation. Large Reynolds and Maxwell stresses, plus separately measured kinetic stress, indicate an intricate momentum balance and a possible origin for MST's intrinsic plasma rotation. Gyrokinetic analysis indicates that micro-tearing modes can be unstable at high beta, with a critical gradient for the electron temperature that is larger than for tokamak plasmas by roughly the aspect ratio.

New Frontier Science Campaign on DIII-D launched in 2017

NASA Astrophysics Data System (ADS)

Koepke, M.; Buttery, R.; Carter, T.; Egedal, J.; Forest, C.; Fox, W.; Ji, H.; Howes, G.; Piovesan, P.; Sarff, J.; Skiff, F.; Spong, D.; DIII-D FSE Collaboration Collaboration

2017-10-01

The DIII-D Frontier Science Experiments initiative explores the potential to use the DIII-D tokamak facility to investigate questions of value beyond the usual fusion-energy science mission of DIII-D. The campaign is unique within DOE-SC-FES because the DIII-D tokamak supplied a multi-day-shot platform for non-fusion-energy-motivated research for the first time. All selected FSE campaign projects competed on the basis of potential intellectual impact and on the degree to which the ability to achieve success as a transformational advance relied on the capabilities of DIII-D. The motivation of the following FSE projects, as well as the selection process, will be summarized (1) Self-organization of Unstable Flux Ropes: Universal Structures in Space/Astrophysical Plasmas (2) Impact of Magnetic Perturbations on Turbulence: Zonal Flow Interactions and Saturation (3) Interaction of Alfvén/whistler fluctuations and Runaway Electrons (4) Self-consistent chaos in magnetic field dynamics These basic-plasma experiments, conducted in collaboration with the DIII-D team, were carried out during 5 shot days in FY2017. Additional days are earmarked in FY2018. Future studies with additional FSE-community members are envisioned. Opportunities exist to piggy back with DIII-D research A proper solicitation and peer review would be appropriate going forward if this activity on DIII-D continues Funding from U.S. DOE is gratefully acknowledged.

High Performance Regimes in Alcator C-Mod at High Magnetic Field

NASA Astrophysics Data System (ADS)

Marmar, E. S.; Alcator C-Mod Team

2017-10-01

Alcator is the only divertor tokamak in the world capable of operating at magnetic fields up to 8 T, equaling and exceeding that planned for ITER. Using RF and microwave tools for auxiliary heating and current drive, C-Mod accesses high pressure, high density, reactor-relevant regimes with no external torque and equilibrated electrons and ions, with exclusive use of high-Z metal plasma-facing components. The 2016 experimental campaign focused on naturally ELM-suppressed, enhanced energy confinement regimes (including I-mode and EDA H-mode, and approaches to super-H-mode), with emphasis on operation at the highest fields (52 atm.) was achieved. Taken together, combined with previous results from C-Mod and the world tokamak database, these results form a strong foundation for the high field, compact approach to achieving fusion energy production. New advances in high temperature, high field superconductors open the possibilities for practical development of this path for commercial fusion. Supported by USDOE.

Simulation of ITG instabilities with fully kinetic ions and drift-kinetic electrons in tokamaks

NASA Astrophysics Data System (ADS)

Hu, Youjun; Chen, Yang; Parker, Scott

2017-10-01

A turbulence simulation model with fully kinetic ions and drift-kinetic electrons is being developed in the toroidal electromagnetic turbulence code GEM. This is motivated by the observation that gyrokinetic ions are not well justified in simulating turbulence in tokamak edges with steep density profile, where ρi / L is not small enough to be used a small parameter needed by the gyrokinetic ordering (here ρi is the gyro-radius of ions and L is the scale length of density profile). In this case, the fully kinetic ion model may be useful. Our model uses an implicit scheme to suppress high-frequency compressional Alfven waves and waves associated with the gyro-motion of ions. The ion orbits are advanced by using the well-known Boris scheme, which reproduces correct drift-motion even with large time-step comparable to the ion gyro-period. The field equation in this model is Ampere's law with the magnetic field eliminated by using an implicit scheme of Faraday's law. The current contributed by ions are computed by using an implicit δf method. A flux tube approximation is adopted, which makes the field equation much easier to solve. Numerical results of electromagnetic ITG obtained from this model will be presented and compared with the gyrokinetic results. This work is supported by U.S. Department of Energy, Office of Fusion Energy Sciences under Award No. DE-SC0008801.

Ion cyclotron emission studies: Retrospects and prospects

DOE PAGES

Gorelenkov, N. N.

2016-06-05

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

Advanced simulation of mixed-material erosion/evolution and application to low and high-Z containing plasma facing components

NASA Astrophysics Data System (ADS)

Brooks, J. N.; Hassanein, A.; Sizyuk, T.

2013-07-01

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (˜4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (˜50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

Ion cyclotron emission studies: Retrospects and prospects

NASA Astrophysics Data System (ADS)

Gorelenkov, N. N.

2016-05-01

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

\\mathscr{H}_2 optimal control techniques for resistive wall mode feedback in tokamaks

NASA Astrophysics Data System (ADS)

Clement, Mitchell; Hanson, Jeremy; Bialek, Jim; Navratil, Gerald

2018-04-01

DIII-D experiments show that a new, advanced algorithm enables resistive wall mode (RWM) stability control in high performance discharges using external coils. DIII-D can excite strong, locked or nearly locked external kink modes whose rotation frequencies and growth rates are on the order of the magnetic flux diffusion time of the vacuum vessel wall. Experiments have shown that modern control techniques like linear quadratic Gaussian (LQG) control require less current than the proportional controller in use at DIII-D when using control coils external to DIII-D’s vacuum vessel. Experiments were conducted to develop control of a rotating n  =  1 perturbation using an LQG controller derived from VALEN and external coils. Feedback using this LQG algorithm outperformed a proportional gain only controller in these perturbation experiments over a range of frequencies. Results from high βN experiments also show that advanced feedback techniques using external control coils may be as effective as internal control coil feedback using classical control techniques.

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

DOE PAGES

Pinsker, R. I.; Austin, M. E.; Diem, S. J.; ...

2014-02-12

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ~2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedlymore » strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. As a result, the AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.« less

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

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

Pinsker, R. I.; Jackson, G. L.; Luce, T. C.

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ∼2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedlymore » strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. The AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.« less

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

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

Pinsker, R. I.; Austin, M. E.; Diem, S. J.

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ~2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedlymore » strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. As a result, the AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.« less

Physics objectives of PI3 spherical tokamak program

NASA Astrophysics Data System (ADS)

Howard, Stephen; Laberge, Michel; Reynolds, Meritt; O'Shea, Peter; Ivanov, Russ; Young, William; Carle, Patrick; Froese, Aaron; Epp, Kelly

2017-10-01

Achieving net energy gain with a Magnetized Target Fusion (MTF) system requires the initial plasma state to satisfy a set of performance goals, such as particle inventory (1021 ions), sufficient magnetic flux (0.3 Wb) to confine the plasma without MHD instability, and initial energy confinement time several times longer than the compression time. General Fusion (GF) is now constructing Plasma Injector 3 (PI3) to explore the physics of reactor-scale plasmas. Energy considerations lead us to design around an initial state of Rvessel = 1 m. PI3 will use fast coaxial helicity injection via a Marshall gun to create a spherical tokamak plasma, with no additional heating. MTF requires solenoid-free startup with no vertical field coils, and will rely on flux conservation by a metal wall. PI3 is 5x larger than SPECTOR so is expected to yield magnetic lifetime increase of 25x, while peak temperature of PI3 is expected to be similar (400-500 eV) Physics investigations will study MHD activity and the resistive and convective evolution of current, temperature and density profiles. We seek to understand the confinement physics, radiative loss, thermal and particle transport, recycling and edge physics of PI3.

Initial Simulations of RF Waves in Hot Plasmas Using the FullWave Code

NASA Astrophysics Data System (ADS)

Zhao, Liangji; Svidzinski, Vladimir; Spencer, Andrew; Kim, Jin-Soo

2017-10-01

FullWave is a simulation tool that models RF fields in hot inhomogeneous magnetized plasmas. The wave equations with linearized hot plasma dielectric response are solved in configuration space on adaptive cloud of computational points. The nonlocal hot plasma dielectric response is formulated by calculating the plasma conductivity kernel based on the solution of the linearized Vlasov equation in inhomogeneous magnetic field. In an rf field, the hot plasma dielectric response is limited to the distance of a few particles' Larmor radii, near the magnetic field line passing through the test point. The localization of the hot plasma dielectric response results in a sparse matrix of the problem thus significantly reduces the size of the problem and makes the simulations faster. We will present the initial results of modeling of rf waves using the Fullwave code, including calculation of nonlocal conductivity kernel in 2D Tokamak geometry; the interpolation of conductivity kernel from test points to adaptive cloud of computational points; and the results of self-consistent simulations of 2D rf fields using calculated hot plasma conductivity kernel in a tokamak plasma with reduced parameters. Work supported by the US DOE ``SBIR program.

High heat flux testing of CFC composites for the tokamak physics experiment

NASA Astrophysics Data System (ADS)

Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

1996-10-01

High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

Synchrotron emission diagnostic of full-orbit kinetic simulations of runaway electrons in tokamaks plasmas

NASA Astrophysics Data System (ADS)

Carbajal Gomez, Leopoldo; Del-Castillo-Negrete, Diego

2017-10-01

Developing avoidance or mitigation strategies of runaway electrons (RE) for the safe operation of ITER is imperative. Synchrotron radiation (SR) of RE is routinely used in current tokamak experiments to diagnose RE. We present the results of a newly developed camera diagnostic of SR for full-orbit kinetic simulations of RE in DIII-D-like plasmas that simultaneously includes: full-orbit effects, information of the spectral and angular distribution of SR of each electron, and basic geometric optics of a camera. We observe a strong dependence of the SR measured by the camera on the pitch angle distribution of RE, namely we find that crescent shapes of the SR on the camera pictures relate to RE distributions with small pitch angles, while ellipse shapes relate to distributions of RE with larger pitch angles. A weak dependence of the SR measured by the camera with the RE energy, value of the q-profile at the edge, and the chosen range of wavelengths is found. Furthermore, we observe that oversimplifying the angular distribution of the SR changes the synchrotron spectra and overestimates its amplitude. Research sponsored by the LDRD Program of ORNL, managed by UT-Battelle, LLC, for the U. S. DoE.

The Effects of Aperiodic Perturbation on the Last Good Surface of a Single-Null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Jordan, Joseph; Punjabi, Alkesh; Ali, Halima

2003-10-01

An area-preserving discrete mapping, called the Simple Map is used to analyze the effects of aperiodic perturbation on the last good magnetic surface of a single-null divertor tokamak. The SM was developed by Punjabi and Boozer /1/. The SM equations are modified to reflect the effect of aperiodic perturbation: Xn + 1 = Xn - (k+δ_aperiodics(n))Y_n(1-Y_n), Y_n+1 = Yn + kX_n+1 where δ_aperiodic is the amplitude of the aperiodic perturbation. The equations are coded into FORTRAN and used in conjunction with a graphing program to show the trajectories for different amplitudes of aperiodic perturbation. As the aperiodic perturbation is increased, the trajectories break from a clean curve and develop strange behavior as they approach chaos. The progression of the trajectories from a clean curve to a chaotic one is observed, and the results are presented in this paper. This project is supported by NASA SHARP and US DOE Grant Number DE-F02-02ER54673. The research was done under the mentorship of Profs. Punjabi and Ali. 1. Punjabi A., Verma A., and Boozer A., Phys. Rev. Lett, 69, 3322 (1992)

Origin of Non-Gaussian Spectra Observed via the Charge Exchange Recombination Spectroscopy Diagnostic in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Sulyman, Alex; Chrystal, Colin; Haskey, Shaun; Burrell, Keith; Grierson, Brian

2017-10-01

The possible observation of non-Maxwellian ion distribution functions in the pedestal of DIII-D will be investigated with a synthetic diagnostic that accounts for the effect of finite neutral beam size. Ion distribution functions in tokamak plasmas are typically assumed to be Maxwellian, however non-Gaussian features observed in impurity charge exchange spectra have challenged this concept. Two possible explanations for these observations are spatial averaging over a finite beam size and a local ion distribution that is non-Maxwellian. Non-Maxwellian ion distribution functions could be driven by orbit loss effects in the edge of the plasma, and this has implications for momentum transport and intrinsic rotation. To investigate the potential effect of finite beam size on the observed spectra, a synthetic diagnostic has been created that uses FIDAsim to model beam and halo neutral density. Finite beam size effects are investigated for vertical and tangential views in the core and pedestal region with varying gradient scale lengths. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program, DE-FC02-04ER54698, and DE-AC02-09CH11466.

Analysis of the Zeeman effect on D α spectra on the EAST tokamak

NASA Astrophysics Data System (ADS)

Gao, Wei; Huang, Juan; Wu, Chengrui; Xu, Zong; Hou, Yumei; Jin, Zhao; Chen, Yingjie; Zhang, Pengfei; Zhang, Ling; Wu, Zhenwei; EAST Team

2017-04-01

Based on the passive spectroscopy, the {{{D}}}α atomic emission spectra in the boundary region of the plasma have been measured by a high resolution optical spectroscopic multichannel analysis (OSMA) system in EAST tokamak. The Zeeman splitting of the {{{D}}}α spectral lines has been observed. A fitting procedure by using a nonlinear least squares method was applied to fit and analyze all polarization π and +/- σ components of the {{{D}}}α atomic spectra to acquire the information of the local plasma. The spectral line shape was investigated according to emission spectra from different regions (e.g., low-field side and high-field side) along the viewing chords. Each polarization component was fitted and classified into three energy categories (the cold, warm, and hot components) based on different atomic production processes, in consistent with the transition energy distribution by calculating the gradient of the {{{D}}}α spectral profile. The emission position, magnetic field intensity, and flow velocity of a deuterium atom were also discussed in the context. Project supported by the National Natural Science Foundation of China (Grant Nos. 11275231 and 11575249) and the National Magnetic Confinement Fusion Energy Research Program of China (Grant No. 2015GB110005).

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.

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

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.

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

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.

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)].

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

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

Simple Map with Low MN Perturbation for a Single-Null Divertor Tokamak with Constant Width of Stochastic Layer

NASA Astrophysics Data System (ADS)

Verma, Arun; Smith, Terry; Punjabi, Alkesh; Boozer, Allen

1996-11-01

In this work, we investigate the effects of low MN perturbations in a single-null divertor tokamak with stochastic scrape-off layer. The unperturbed magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). We choose the combinations of the map parameter k, and the strength of the low MN perturbation such that the width of stochastic layer remains unchanged. We give detailed results on the effects of low MN perturbation on the magnetic topology of the stochastic layer and on the footprint of field lines on the divertor plate given the constraint of constant width of the stochastic layer. The low MN perturbations occur naturally and therefore their effects are of considerable importance in tokamak divertor physics. This work is supported by US DOE OFES. Use of CRAY at HU and at NERSC is gratefully acknowledged.

Overview of MST Research

NASA Astrophysics Data System (ADS)

Sarff, J. S.; MST Team

2011-10-01

MST progress in advancing the RFP for (1) fusion plasma confinement with minimal external magnetization, (2) toroidal confinement physics, and (3) basic plasma physics is summarized. New tools and diagnostics are accessing physics barely studied in the RFP. Several diagnostic advances are important for ITER/burning plasma. A 1 MW neutral beam injector operates routinely for fast ion, heating, and transport investigations. Energetic ions are also created spontaneously by tearing mode reconnection, reminiscent of astrophysical plasmas. Classical confinement of impurity ions is measured in reduced-tearing plasmas. Fast ion slowing-down is also classical. Alfven-eigenmode-like activity occurs with NBI, but apparently not TAE. Stellarator-like helical structure appears in the core of high current plasmas, with improved confinement characteristics. FIR interferometry, Thomson scattering, and HIBP diagnostics are beginning to explore microturbulence scales, an opportunity to exploit the RFP's high beta and strong magnetic shear parameter space. A programmable power supply for the toroidal field flexibly explores scenarios from advanced inductive profile control to low current tokamak operation. A 1 MW 5.5 GHz source for electron Bernstein wave injection is nearly complete to investigate heating and current drive in over-dense plasmas. Supported by DOE and NSF.

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

DOE PAGES

Moreau, Didier; Artaud, J. F.; Ferron, John R.; ...

2015-05-01

This paper shows that semi-empirical data-driven models based on a twotime- scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, β N, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated data obtained using a rapidly converging plasmamore » transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0-D scaling laws and 1.5-D ordinary differential equations. A number of open loop simulations were performed, in which the heating and current drive (H&CD) sources were randomly modulated around the typical values of a reference AT discharge on DIIID. Using these simulated data, a two-time-scale state space model was obtained for the coupled evolution of the poloidal flux profile and βN parameter, and a controller was synthesized based on the near-optimal ARTAEMIS algorithm [D. Moreau et al., Nucl. Fusion 53 (2013) 063020]. The paper discusses the results of closed-loop nonlinear simulations, using this controller for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and β N are satisfactorily tracked with a time scale of about ten seconds, despite large disturbances applied to the feedforward powers and plasma parameters. The effectiveness of the control algorithm is thus demonstrated for long pulse and steady state high-β N AT discharges. Its robustness with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.« less

Advanced Tokamak Investigations in Full-Tungsten ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Bock, Alexander

2017-10-01

The tailoring of the q-profile is the foundation of Advanced Tokamak (AT) scenarios. It depends on low collisionality ν* which permits efficient external current drive and high amounts of intrinsic bootstrap current. At constant pressure, lowering ne leads to a strong decrease of ν* Te - 3 . After the conversion of ASDEX Upgrade to fully W-coated plasma facing components, radiative collapses of H-modes with little gas puffing due to central W accumulation could only be avoided partially with central ECRH. Also, operation at high β with low ne presented a challenge for the divertor. Together, these issues prevented meaningful AT investigations. To overcome this, several major feats have been accomplished: Access to lower ne was achieved through a better understanding of the changes to recycling and pumping, and optionally the density pump-out phenomenon due to RMPs. ECRH capacities were substantially expanded for both heating and current drive, and a solid W divertor capable of withstanding the power loads was installed. A major overhaul improved the reliability of the current profile diagnostics. This contribution will detail the efforts needed to re-access AT scenarios and report on the development of candidate steady state scenarios for ITER/DEMO. Starting from the `hybrid scenario,' a non-inductive scenario (q95 = 5.3 , βN = 2.7 , fbs > 40 %) was developed. It can be sustained for many τE, limited only by technical boundaries, and is also independent of the ramp-up scenario. The β-limit is set by ideal modes that convert into NTMs. The Ti-profiles are steeper than predicted by TGLF, but nonlinear electromagnetic gyro-kinetic analyses with GENE including fast particle effects matched the experimental heat fluxes. We will also report on scenarios at higher q95, similar to the EAST/DIII-D steady state scenario. The extrapolation of these scenarios to ITER/DEMO will be discussed.

An innovative small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

Guo, Houyang

2017-10-01

A new Small Angle Slot (SAS) divertor is being developed in DIII-D to address the challenge of efficient divertor heat dispersal at the relatively low plasma density required for non-inductive current drive in future advanced tokamaks. SAS features a small incident angle near the plasma strike point on the divertor target plate with a progressively opening slot. SOLPS (B2-Eirene) edge code analysis finds that SAS can achieve strong plasma cooling when the strike point is placed near the small angle target plate in the slot, leading to low electron temperature Te across the entire divertor target. This is enabled by strong coupling between a gas tight slot and directed neutral recycling by the small angle target to enhance neutral buildup near the target. SOLPS analysis reveals a strong correlation between Te and D2 density at the target for various divertor configurations including the flat target, slanted target, and lower single null divertor. The strong correlation suggests that achievement of low Te may reduce essentially to identifying the divertor baffle geometry that achieves the highest target gas density at a given upstream condition. The SAS divertor concept has recently been tested in DIII-D for a range of plasma configurations and conditions with precise control of slot strike point location. In confirmation of SOLPS predictions, a sharp transition is observed when the strike point is moved to the critical outer corner of SAS. A set of Langmuir probes imbedded in SAS show that the Te radial profile, which is peaked at the strike point when it is located away from the SAS corner, becomes low across the target when the strike point is located near the corner. With further increase in density, deep-slot detachment occurs with Te 1 eV, measured by the unique DIII-D divertor Thomson Scattering diagnostic. Work supported by US DOE under DE-FC02-04ER54698.

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.

Coherent nonlinear coupling between a long-wavelength mode and small-scale turbulence in the TEXT tokamak

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

Tsui, H.Y.W.; Rypdal, K.; Ritz, C.P.

1993-04-26

Bispectral analysis of Langmuir probe data indicates that coherent nonlinear coupling, in addition to the noncoherent turbulent interactions, exists in the edge plasma of the tokamak TEXT. Not all the modes involved reside within the spectral region of the usual broadband turbulence. At a major resonant surface the small-scale turbulent activity interacts [ital coherently] with a localized long-wavelength mode; a signature of regular or coherent structure. By the observed coupling to the transport related turbulence, the long-wavelength mode can influence plasma confinement indirectly. These observations signify the influence of low-order resonant surfaces on the edge turbulence in tokamaks.

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

Rasouli, C.; Abbasi Davani, F., E-mail: fabbasidavani@gmail.com

A series of experiments and numerical calculations have been done on the Damavand tokamak for accurate determination of equilibrium parameters, such as the plasma boundary position and shape. For this work, the pickup coils of the Damavand tokamak were recalibrated and after that a plasma boundary shape identification code was developed for analyzing the experimental data, such as magnetic probes and coils currents data. The plasma boundary position, shape and other parameters are determined by the plasma shape identification code. A free-boundary equilibrium code was also generated for comparison with the plasma boundary shape identification results and determination of requiredmore » fields to obtain elongated plasma in the Damavand tokamak.« less

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.

Study of runaway electrons with Hard X-ray spectrometry of tokamak plasmas

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

Shevelev, A.; Chugunov, I.; Khilkevitch, E.

2014-08-21

Hard-X-ray spectrometry is a tool widely used for diagnostic of runaway electrons in existing tokamaks. In future machines, ITER and DEMO, HXR spectrometry will be useful providing information on runaway electron energy, runaway beam current and its profile during disruption.

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).

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

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.

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.

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

Public Data Set: 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.

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in J.L. Barr et. al, 'A Power-Balance Model for Local Helicity Injection Startup in a Spherical Tokamak,' Nuclear Fusion 58, 076011 (2018).

[The sawtooth oscillation phenomenon of visible spectral signal in HT-6M Tokamak].

PubMed

Xu, W; Fang, Z; Wan, B; Li, J; Luo, J; Yin, F

1997-02-01

The sawtooth oscillation phenomenon of visible spectral signal in HT-6M Tokamak is presented. The influences of electron temperature, electron density and atomic ground density on the spectral signal discussed. This phenomenon results mainly from the change of electron temperature at the edge.

Remote experimental site concept development

NASA Astrophysics Data System (ADS)

Casper, Thomas A.; Meyer, William; Butner, David

1995-01-01

Scientific research is now often conducted on large and expensive experiments that utilize collaborative efforts on a national or international scale to explore physics and engineering issues. This is particularly true for the current US magnetic fusion energy program where collaboration on existing facilities has increased in importance and will form the basis for future efforts. As fusion energy research approaches reactor conditions, the trend is towards fewer large and expensive experimental facilities, leaving many major institutions without local experiments. Since the expertise of various groups is a valuable resource, it is important to integrate these teams into an overall scientific program. To sustain continued involvement in experiments, scientists are now often required to travel frequently, or to move their families, to the new large facilities. This problem is common to many other different fields of scientific research. The next-generation tokamaks, such as the Tokamak Physics Experiment (TPX) or the International Thermonuclear Experimental Reactor (ITER), will operate in steady-state or long pulse mode and produce fluxes of fusion reaction products sufficient to activate the surrounding structures. As a direct consequence, remote operation requiring robotics and video monitoring will become necessary, with only brief and limited access to the vessel area allowed. Even the on-site control room, data acquisition facilities, and work areas will be remotely located from the experiment, isolated by large biological barriers, and connected with fiber-optics. Current planning for the ITER experiment includes a network of control room facilities to be located in the countries of the four major international partners; USA, Russian Federation, Japan, and the European Community.

Long pulse operation of the Kamaboko negative ion source on the MANTIS test bed

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

Tramham, R.; Jacquot, C.; Riz, D.

1998-08-20

Advanced Tokamak concepts and steady state plasma scenarios require external plasma heating and current drive for extended time periods. This poses several problems for the neutral beam injection systems that are currently in use. The power loading of the ion source and accelerator are especially problematic. The Kamaboko negative ion source, a small scale model of the ITER arc source, is being prepared for extended operation of deuterium beams for up to 1000 seconds. The operating conditions of the plasma grid prove to be important for reducing electron power loading of the accelerator. Operation of deuterium beams for extended periodsmore » also poses radiation safety risks which must be addressed.« less

An accurate automated technique for quasi-optics measurement of the microwave diagnostics for fusion plasma

NASA Astrophysics Data System (ADS)

Hu, Jianqiang; Liu, Ahdi; Zhou, Chu; Zhang, Xiaohui; Wang, Mingyuan; Zhang, Jin; Feng, Xi; Li, Hong; Xie, Jinlin; Liu, Wandong; Yu, Changxuan

2017-08-01

A new integrated technique for fast and accurate measurement of the quasi-optics, especially for the microwave/millimeter wave diagnostic systems of fusion plasma, has been developed. Using the LabVIEW-based comprehensive scanning system, we can realize not only automatic but also fast and accurate measurement, which will help to eliminate the effects of temperature drift and standing wave/multi-reflection. With the Matlab-based asymmetric two-dimensional Gaussian fitting method, all the desired parameters of the microwave beam can be obtained. This technique can be used in the design and testing of microwave diagnostic systems such as reflectometers and the electron cyclotron emission imaging diagnostic systems of the Experimental Advanced Superconducting Tokamak.

Kinetics of relativistic runaway electrons

NASA Astrophysics Data System (ADS)

Breizman, B. N.; Aleynikov, P. B.

2017-12-01

This overview covers recent developments in the theory of runaway electrons in tokamaks. Its main purpose is to outline the intuitive basis for first-principle advancements in runaway electron physics. The overview highlights the following physics aspects of the runaway evolution: (1) survival and acceleration of initially hot electrons during thermal quench, (2) effect of magnetic perturbations on runaway confinement, (3) multiplication of the runaways via knock-on collisions with the bulk electrons, (4) slow decay of the runaway current, and (5) runaway-driven micro-instabilities. The scope of the reported studies is governed by the need to understand the behavior of runaway electrons as an essential physics element of the disruption events in ITER in order to develop an effective runaway mitigation scheme. ).

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.

Plasma rotation measurement in small tokamaks using an optical spectrometer and a single photomultiplier as detector.

PubMed

Severo, J H F; Nascimento, I C; Kuznetov, Yu K; Tsypin, V S; Galvão, R M O; Tendler, M

2007-04-01

The method for plasma rotation measurement in the tokamak TCABR is reported in this article. During a discharge, an optical spectrometer is used to scan sequentially spectral lines of plasma impurities and spectral lines of a calibration lamp. Knowing the scanning velocity of the diffraction grating of the spectrometer with adequate precision, the Doppler shifts of impurity lines are determined. The photomultiplier output voltage signals are recorded with adequate sampling rate. With this method the residual poloidal and toroidal plasma rotation velocities were determined, assuming that they are the same as those of the impurity ions. The results show reasonable agreement with the neoclassical theory and with results from similar tokamaks.

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.

Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma.

PubMed

Chrystal, C; Burrell, K H; Grierson, B A; Groebner, R J; Kaplan, D H

2012-10-01

To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

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.

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.

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.

Development of a new gas puff imaging diagnostic on the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Yuan, B.; Xu, M.; Yu, Y.; Zang, L.; Hong, R.; Chen, C.; Wang, Z.; Nie, L.; Ke, R.; Guo, D.; Wu, Y.; Long, T.; Gong, S.; Liu, H.; Ye, M.; Duan, X.; HL-2A team

2018-03-01

A new gas puff imaging (GPI) diagnostic has been developed on the HL-2A tokamak to study two-dimensional plasma edge turbulence in poloidal vs. radial plane. During a discharge, neutral helium or deuterium gas is puffed at the edge of the plasma through a rectangular multi\

Public Data Set: High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak

DOE Data Explorer

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

2016-04-27

This data set contains openly-documented, machine readable digital research data corresponding to figures published in K.E. Thome et al., 'High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak,' Phys. Rev. Lett. 116, 175001 (2016).

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.

Public Data Set: Initiation and Sustainment of Tokamak Plasmas with Local Helicity Injection as the Majority Current Drive

DOE Data Explorer

Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609); Bodner, Grant M. [University of Wisconsin-Madison] (ORCID:0000000324979172); 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); Pachicano, Jessica L. [University of Wisconsin-Madison] (ORCID:0000000207255693); Pierren, Christopher [University of Wisconsin-Madison] (ORCID:0000000228289825); Reusch, Joshua A. [University of Wisconsin-Madison] (ORCID:0000000284249422); Rhodes, Alexander T. [University of Wisconsin-Madison] (ORCID:0000000280735714); Richner, Nathan J. [University of Wisconsin-Madison] (ORCID:0000000155443915); Rodriguez Sanchez, Cuauhtemoc [University of Wisconsin-Madison] (ORCID:0000000334712586); Schaefer, Carolyn E. [University of Wisconsin-Madison] (ORCID:0000000248848727); Weberski, Justin D. [University of Wisconsin-Madison] (ORCID:0000000256267914)

2018-05-22

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in J.M. Perry et al., 'Initiation and Sustainment of Tokamak Plasmas with Local Helicity Injection as the Majority Current Drive,' accepted for publication in Nuclear Fusion.

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-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

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).

Avoiding Tokamak Disruptions by Applying Static Magnetic Fields That Align Locked Modes with Stabilizing Wave-Driven Currents [Avoiding Tokamak Disruptions by Magnetically Aligning Locked Modes with Stabilizing Wave-Driven Currents

DOE PAGES

Volpe, F. A.; Hyatt, Alan; La Haye, Robert J.; ...

2015-10-19

The international ITER tokamak has the objective of demonstrating the scientific feasibility of magnetic confinement fusion as a source of energy. A concern towards the achievement of this goal is represented by major disruptions: complete losses of confinement often initiated by a non-rotating ('locked') magnetic island created by magnetic reconnection. During disruptions, energy and particles accumulated in the plasma volume over many seconds are lost in a few milliseconds and released on the plasma-facing materials. In addition, multi-MA level currents flowing in the tokamak plasma for its sustainment and confinement are lost, also in milliseconds, thus terminating the plasma dischargemore » and causing electromagnetic stresses that, if unmitigated, could lead to excessive device wear. Moreover it is shown that magnetic perturbations can be used to avoid disruptions by "guiding" the magnetic island to lock in a position where it is accessible to millimetre wave beams that fully stabilize it.« less

Neutral helium beam probe

NASA Astrophysics Data System (ADS)

Karim, Rezwanul

1999-10-01

This article discusses the development of a code where diagnostic neutral helium beam can be used as a probe. The code solves numerically the evolution of the population densities of helium atoms at their several different energy levels as the beam propagates through the plasma. The collisional radiative model has been utilized in this numerical calculation. The spatial dependence of the metastable states of neutral helium atom, as obtained in this numerical analysis, offers a possible diagnostic tool for tokamak plasma. The spatial evolution for several hypothetical plasma conditions was tested. Simulation routines were also run with the plasma parameters (density and temperature profiles) similar to a shot in the Princeton beta experiment modified (PBX-M) tokamak and a shot in Tokamak Fusion Test Reactor tokamak. A comparison between the simulation result and the experimentally obtained data (for each of these two shots) is presented. A good correlation in such comparisons for a number of such shots can establish the accurateness and usefulness of this probe. The result can possibly be extended for other plasma machines and for various plasma conditions in those machines.

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

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.

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

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

Paleoclassical transport explains electron transport barriers in RTP and TEXTOR

NASA Astrophysics Data System (ADS)

Hogeweij, G. M. D.; Callen, J. D.; RTP Team; TEXTOR Team

2008-06-01

The recently developed paleoclassical transport model sets the minimum level of electron thermal transport in a tokamak. This transport level has proven to be in good agreement with experimental observations in many cases when fluctuation-induced anomalous transport is small, i.e. in (near-)ohmic plasmas in small to medium size tokamaks, inside internal transport barriers (ITBs) or edge transport barriers (H-mode pedestal). In this paper predictions of the paleoclassical transport model are compared in detail with data from such kinds of discharges: ohmic discharges from the RTP tokamak, EC heated RTP discharges featuring both dynamic and shot-to-shot scans of the ECH power deposition radius and off-axis EC heated discharges from the TEXTOR tokamak. For ohmically heated RTP discharges the Te profiles predicted by the paleoclassical model are in reasonable agreement with the experimental observations, and various parametric dependences are captured satisfactorily. The electron thermal ITBs observed in steady state EC heated RTP discharges and transiently after switch-off of off-axis ECH in TEXTOR are predicted very well by the paleoclassical model.

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

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

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.

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

NASA Astrophysics Data System (ADS)

Adamek, J.; Müller, H. W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horacek, J.; Kurzan, B.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R.; Fernandes, H.; Figueiredo, H.

2016-04-01

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 (ΦBPP) and the floating potential (Vfl) 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 Te = (ΦBPP - Vfl)/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.

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

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

PubMed

Adamek, J; Müller, H W; Silva, C; Schrittwieser, R; Ionita, C; Mehlmann, F; Costea, S; Horacek, J; Kurzan, B; Bilkova, P; Böhm, P; Aftanas, M; Vondracek, P; Stöckel, J; Panek, R; Fernandes, H; Figueiredo, H

2016-04-01

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 (ΦBPP) and the floating potential (Vfl) 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 Te = (ΦBPP - Vfl)/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.

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.

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.

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

Harvey, R. W.

This DOE grant supported fusion energy research, a potential long-term solution to the world's energy needs. Magnetic fusion, exemplified by confinement of very hot ionized gases, i.e., plasmas, in donut-shaped tokamak vessels is a leading approach for this energy source. Thus far, a mixture of hydrogen isotopes has produced 10's of megawatts of fusion power for seconds in a tokamak reactor at Princeton Plasma Physics Laboratory in New Jersey. The research grant under consideration, ER54684, uses computer models to aid in understanding and projecting efficacy of heating and current drive sources in the National Spherical Torus Experiment, a tokamak variant,more » at PPPL. The NSTX experiment explores the physics of very tight aspect ratio, almost spherical tokamaks, aiming at producing steady-state fusion plasmas. The current drive is an integral part of the steady-state concept, maintaining the magnetic geometry in the steady-state tokamak. CompX further developed and applied models for radiofrequency (rf) heating and current drive for applications to NSTX. These models build on a 30 year development of rf ray tracing (the all-frequencies GENRAY code) and higher dimensional Fokker-Planck rf-collisional modeling (the 3D collisional-quasilinear CQL3D code) at CompX. Two mainline current-drive rf modes are proposed for injection into NSTX: (1) electron Bernstein wave (EBW), and (2) high harmonic fast wave (HHFW) modes. Both these current drive systems provide a means for the rf to access the especially high density plasma--termed high beta plasma--compared to the strength of the required magnetic fields. The CompX studies entailed detailed modeling of the EBW to calculate the efficiency of the current drive system, and to determine its range of flexibility for driving current at spatial locations in the plasma cross-section. The ray tracing showed penetration into NSTX bulk plasma, relatively efficient current drive, but a limited ability to produce current over the whole radial plasma cross-section. The actual EBW experiment will cost several million dollars, and remains in the proposal stage. The HHFW current drive system has been experimentally implemented on NSTX, and successfully drives substantial current. The understanding of the experiment is to be accomplished in terms of general concepts of rf current drive, and also detailed modeling of the experiment which can discern the various competing processes which necessarily occur simultaneously in the experiment. An early discovery of the CompX codes, GENRAY and CQL3D, was that there could be significant interference between the neutral beam injection fast ions in the machine (injected for plasma heating) and the HHFW energy. Under many NSTX experimental conditions, power which could go to the fast ions would then be unavailable for current drive by the desired HHFW interaction with electrons. This result has been born out by experiments; the modeling helps in understanding difficulties with HHFW current drive, and has enabled adjustment of the experiment to avoid interaction with neutral beam injected fast ions thereby achieving stronger HHFW current drive. The detailed physics modeling of the various competing processes is almost always required in fusion energy plasma physics, to ensure a reasonably accurate and certain interpretation of the experiment, enabling the confident design of future, more advanced experiments and ultimately a commercial fusion reactor. More recent work entails detailed investigation of the interaction of the HHFW radiation for fast ions, accounting for the particularly large radius orbits in NSTX, and correlations between multiple HHFW-ion interactions. The spherical aspect of the NSTX experiment emphasized particular physics such as the large orbits which are present to some degree in all tokamaks, but gives clearer clues on the resulting physics phenomena since competing physics effects are reduced.« less

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.

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.

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.

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.

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

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.

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)