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Sample records for alcator c-mod tokamak

  1. Divertor bypass in the Alcator C-Mod tokamak

    Science.gov (United States)

    Pitcher, C. S.; LaBombard, B.; Danforth, R.; Pina, W.; Silveira, M.; Parkin, B.

    2001-01-01

    The Alcator C-Mod divertor bypass has for the first time allowed in situ variations to the mechanical baffle design in a tokamak. The design utilizes small coils which interact with the ambient magnetic field inside the vessel to provide the torque required to control small flaps of a Venetian blind geometry. Plasma physics experiments with the bypass have revealed the importance of the divertor baffling to maintain high divertor gas pressures. These experiments have also indicated that the divertor baffling has only a limited effect on the main chamber pressure in C-Mod.

  2. Neutral particle dynamics in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Niemczewski, A.P.

    1995-08-01

    This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism. 146 refs., 82 figs., 14 tabs.

  3. Alcator C-Mod: A high-field divertor tokamak

    Science.gov (United States)

    Lipschultz, B.; Becker, H.; Bonoli, P.; Coleman, J.; Fiore, C.; Golovato, S.; Granetz, R.; Greenwald, M.; Gwinn, D.; Humphries, D.; Hutchinson, I.; Irby, J.; Marmar, E.; Montgomery, D. B.; Najmabadi, F.; Parker, R.; Porkolab, M.; Rice, J.; Sevillano, E.; Takase, Y.; Terry, J.; Watterson, R.; Wolfe, S.

    1989-04-01

    The Alcator C-Mod tokamak is a new device presently under construction at Massachusetts Institute of Technology (M.I.T.) which is scheduled to begin operation in mid-1990. The projected operating parameters are as follows: Toroidal field of 9 T; Ip ≤ 3 MA, R = 66.5 cm, a = 21 cm, κ ≤ 2.0, δ ≤ 0.5, ne ≤ 10 21m-3, PICRF ≤ 6 MW. The divertor configuration includes mechanical baffling as opposed to an 'open' geometry. Under strictly ohmic heating conditions, central Ti and Te are predicted to be in the range 2.5-3.5 keV over the density range (4-8) × 10 20m-3. With the addition of 6 MW of ICRF heating, Ti should vary from 4-8 keV over the same density range (assuming either Kaye-Goldston or Neo-Alcator scalings for electron confinement). Based on edge plasma characterizations from Alcator-C and divertor tokamaks, the scrape-off layer (SOL) properties are predicted to be: λn ≈ 10mm, density at the divertor plate < 2 × 10 21m-3, H 0 ionization mean free path between 1 and 10 mm. Maximum heat loads on various internal components are predicted to be in the range 5-10 MW/m 2. The flexibility of the poloidal field system in forming a number of flux surface geometries will provide further comparisons of the relative impurity control capabilities of double-null, single-null and limiter plasmas.

  4. Edge Turbulence Imaging in the Alcator C-Mod Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben; D.P. Stotler; J.L. Terry; B. LaBombard; M. Greenwald; M. Muterspaugh; C.S. Pitcher; the Alcator C-Mod Group; K. Hallatschek; R.J. Maqueda; B. Rogers; J.L. Lowrance; V.J. Mastrocola; G.F. Renda

    2001-11-26

    The 2-D radial vs. poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nuclear Fusion 41(2001) 1391] was measured using fast cameras and compared with 3-D numerical simulations of edge plasma turbulence. The main diagnostic is Gas Puff Imaging (GPI), in which the visible D(subscript alpha) emission from a localized D(subscript 2) gas puff is viewed along a local magnetic field line. The observed D(subscript alpha) fluctuations have a typical radial and poloidal scale of approximately 1 cm, and often have strong local maxima (''blobs'') in the scrape-off layer. The motion of this 2-D structure motion has also been measured using an ultra-fast framing camera with 12 frames taken at 250,000 frames/sec. Numerical simulations produce turbulent structures with roughly similar spatial and temporal scales and transport levels as that observed in the experiment; however, some differences are also noted, perhaps requiring diagnostic improvement and/or additional physics in the numerical model.

  5. Edge Turbulence Imaging in the Alcator C-Mod Tokamak

    International Nuclear Information System (INIS)

    The 2-D radial vs. poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nuclear Fusion 41(2001) 1391] was measured using fast cameras and compared with 3-D numerical simulations of edge plasma turbulence. The main diagnostic is Gas Puff Imaging (GPI), in which the visible D(subscript alpha) emission from a localized D(subscript 2) gas puff is viewed along a local magnetic field line. The observed D(subscript alpha) fluctuations have a typical radial and poloidal scale of approximately 1 cm, and often have strong local maxima (''blobs'') in the scrape-off layer. The motion of this 2-D structure motion has also been measured using an ultra-fast framing camera with 12 frames taken at 250,000 frames/sec. Numerical simulations produce turbulent structures with roughly similar spatial and temporal scales and transport levels as that observed in the experiment; however, some differences are also noted, perhaps requiring diagnostic improvement and/or additional physics in the numerical model

  6. TSC [Tokamak Simulation Code] simulations of Alcator C-MOD discharges

    International Nuclear Information System (INIS)

    The axisymmetric stability of the single X-point, nominal Alcator C-MOD configuration is investigated with the Tokamak Simulation Code. The resistive wall passive growth rate, in the absence of feedback stabilization, is obtained. The instability is suppressed with an appropriate active feedback system. 2 refs., 25 figs., 1 tab

  7. 20 years of research on the Alcator C-Mod tokamak

    OpenAIRE

    Greenwald, Martin; Bader, A; Baek, S.; M. Bakhtiari; Barnard, H.; Beck, W.; Bergerson, W; Bespamyatnov, I; Bonoli, P.; Brower, D; Brunner, D.; Burke, W.; Candy, J.; Churchill, M; Cziegler, I.

    2014-01-01

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of crit...

  8. Advanced Tokamak Regimes in Alcator C-Mod with Lower Hybrid Current Drive

    Science.gov (United States)

    Parker, R.; Bonoli, P.; Gwinn, D.; Hutchinson, I.; Porkolab, M.; Ramos, J.; Bernabei, S.; Hosea, J.; Wilson, R.

    1999-11-01

    Alcator C-Mod has been proposed as a test-bed for developing advanced tokamak scenarios owing to its strong shaping, relatively long pulse length capability at moderate field, e.g. t ~ L/R at B = 5T and T_eo ~ 7keV, and the availability of strong ICRF heating. We plan to exploit this capability by installing up to 4 MW RF power at 4.6 GHz for efficient off-axis current drive by lower hybrid waves. By launching LH waves with a grill whose n_xx spectrum can be dynamically controlled over the range 2 2. Such reversed or nearly zero shear regimes have already been proposed as the basis of an advanced tokamak burning-plasma experiment-ATBX (M. Porkolab et al, IAEA-CN-69/FTP/13, IAEA,Yokohama 1998.), and could provide the basis for a demonstration power reactor. Theoretical and experimental basis for this advanced tokamak research program on C-Mod, including design of the lower hybrid coupler, its spectrum and current drive capabilities will be presented.

  9. High resolution measurements of neutral density and ionization rate in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R. L.; Hughes, J. W.; LaBombard, B.; Mossessian, D.; Terry, J. L.

    2001-01-01

    Two new high resolution detectors have been installed on the Alcator C-Mod tokamak to measure the neutral density and ionization rate at the edge of the main chamber plasma. Using a silicon detector sensitive to UV light, and a very narrow filter with transmission peaking at 1216 Aa, the Lyman alpha radiation emanating from neutral deuterium (and hydrogen) is measured. The detectors consist of 20 channel arrays which view the plasma tangentially 12.5 cm below the outer midplane, and 10 cm above the inner midplane. The imaging is performed using a 1mmx3mm slit, which gives a nominal radial resolution of 2 and 3 mm, respectively. The local emissivity is obtained via a standard Abel inversion technique. Employing well-known branching ratios, and using measured local electron density and temperature, the neutral density and ionization rate are inferred with similar radial resolution. Details of the setup and sensitivity of the results to plasma conditions are discussed.

  10. High resolution bolometry on the Alcator C-Mod tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R.L.; Goetz, J.A.; Marmar, E.S.; Rice, J.E.; Terry, J.L. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, 175 Albany Street, Cambridge, Massachusetts 02139 (United States)

    1999-01-01

    Recent breakthroughs in silicon detector technology now permit measurement of radiated power over a wide range in photon energies. These detectors (also known as AXUV photodiodes) have a flat spectral power response from ultraviolet to x-ray energies, and with a slightly reduced efficiency all the way down to visible wavelengths. Since they can be made small, multichannel detectors allow high spatial resolution to be combined with an intrinsic high temporal resolution, which can reach the microsecond range, depending on the application. Additional features include ease of use and installation, and relatively low cost compared to other techniques. A combination of two multichannel toroidally viewing systems has been recently installed on the Alcator C-Mod tokamak. The first array, which is composed of 16 channels, sees tangentially the outer-half of the plasma at the midplane, and is used to measure the total power radiated. The second array, also located at the midplane, consists of 19 channels and views the edge of the plasma. This array has a 2 mm radial resolution, allowing, for example, the study of edge dynamics in high confinement (H mode) plasmas. Because these detectors are largely insensitive to neutral particles (at least at particle energies of interest), it is now possible to measure the radial distribution of neutral {open_quotes}radiated{close_quotes} power emissivity, by looking at the difference between these measurements and those obtained with standard bolometers. When neutrals are not important, we found a very good agreement between the AXUV detectors and standard bolometers. Examples of applications of these measurements to the study of edge H-mode dynamics, impurity injection, disruptions, and internal barrier formation, are described. Planned upgrades and new applications for Alcator C-Mod are also discussed. {copyright} {ital 1999 American Institute of Physics.}

  11. The Submillimeter Wave Electron Cyclotron Emission Diagnostic for the Alcator C-Mod Tokamak.

    Science.gov (United States)

    Hsu, Thomas C.

    This thesis describes the engineering design, construction, and operation of a high spatial resolution submillimeter wave diagnostic for electron temperature measurements on Alcator C-Mod. Alcator C-Mod is a high performance compact tokamak capable of producing diverted, shaped plasmas with a major radius of 0.67 meters, minor radius of 0.21 centimeters, plasma current of 3 MA. The maximum toroidal field is 9 Tesla on the magnetic axis. The ECE diagnostic includes three primary components: a 10.8 meter quasioptical transmission line, a rapid scanning Michelson interferometer, and a vacuum compatible calibration source. Due to the compact size and high field of the tokamak the ECE system was designed to have a spectral range from 100 to 1000 GHz with frequency resolution of 5 GHz and spatial resolution of one centimeter. The beamline uses all reflecting optical elements including two off-axis parabolic mirrors with diameters of 20 cm. and focal lengths of 2.7 meters. Techniques are presented for grinding and finishing the mirrors to sufficient surface quality to permit optical alignment of the system. Measurements of the surface figure confirm the design goal of 1/4 wavelength accuracy at 1000 GHz. Extensive broadband tests of the spatial resolution of the ECE system are compared to a fundamental mode Gaussian beam model, a three dimensional vector diffraction model, and a geometric optics model. The Michelson interferometer is a rapid scanning polarization instrument which has an apodized frequency resolution of 5 GHz and a minimum scan period of 7.5 milliseconds. The novel features of this instrument include the use of precision linear bearings to stabilize the moving mirror and active counterbalancing to reduce vibration. Beam collimation within the instrument is done with off-axis parabolic mirrors. The Michelson also includes a 2-50 mm variable aperture and two signal attenuators constructed from crossed wire grid polarizers. To make full use of the advantages

  12. High resolution visible continuum imaging diagnostic on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Marmar, E. S.; Boivin, R. L.; Granetz, R. S.; Hughes, J. W.; Lipschultz, B.; McCool, S.; Mossessian, D.; Pitcher, C. S.; Rice, J. E.; Terry, J. L.

    2001-01-01

    A high spatial resolution CCD based one-dimensional imaging system to measure visible continuum emissivity profiles from Alcator C-Mod tokamak plasmas is described. The instrument has chordal resolution that is better than 1 mm for the edge region of the plasma, where very sharp (1 to 10 mm) gradient lengths in plasma parameters are observed after the formation of the H-mode transport barrier. Each image has up to 2048 pixels, and total spatial coverage goes from 2 cm inside of the magnetic axis to {approx}4 cm outside of the last closed flux surface in the {approx}22 cm horizontal minor radius plasmas. Time resolution can be varied from 0.21 ms to 4 ms; good signal to noise is achieved with 1 ms integration under typical plasma conditions. The emission over most of the plasma volume is dominated by free--free bremsstrahlung, and can be used to infer local values of the average ion charge (Z{sub eff}). Toroidally localized puffing of deuterium, nitrogen, and helium reveals that a significant contribution to the signal in the scrape-off layer at the extreme edge of the plasma can come from diatomic molecular band pseudocontinuum emission.

  13. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    D.P. Stotler; C.S. Pitcher; C.J. Boswell; B. LaBombard; J.L. Terry; J.D. Elder; S. Lisgo

    2002-05-07

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter.

  14. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    International Nuclear Information System (INIS)

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter

  15. Lithium pellet injection experiments on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, D.T.

    1996-06-01

    A pellet enhanced performance mode, showing significantly reduced core transport, is regularly obtained after the injection of deeply penetrating lithium pellets into Alcator C-Mod discharges. These transient modes, which typically persist about two energy confinement times, are characterized by a steep pressure gradient ({ell}{sub p} {le} a/5) in the inner third of the plasma, indicating the presence of an internal transport barrier. Inside this barrier, particle and energy diffusivities are greatly reduced, with ion thermal diffusivity dropping to near neoclassical values. Meanwhile, the global energy confinement time shows a 30% improvement over ITER89-P L-mode scaling. The addition of ICRF auxiliary heating shortly after the pellet injection leads to high fusion reactivity with neutron rates enhanced by an order of magnitude over L-mode discharges with similar input powers. A diagnostic system for measuring equilibrium current density profiles of tokamak plasmas, employing high speed lithium pellets, is also presented. Because ions are confined to move along field lines, imaging the Li{sup +} emission from the toroidally extended pellet ablation cloud gives the direction of the magnetic field. To convert from temporal to radial measurements, the 3-D trajectory of the pellet is determined using a stereoscopic tracking system. These measurements, along with external magnetic measurements, are used to solve the Grad-Shafranov equation for the magnetic equilibrium of the plasma. This diagnostic is used to determine the current density profile of PEP modes by injection of a second pellet during the period of good confinement. This measurement indicates that a region of reversed magnetic shear exists at the plasma core. This current density profile is consistent with TRANSP calculations for the bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasmas are near the n = {infinity} and the n = 1 marginal stability limits.

  16. Electron heating via mode converted ion Bernstein waves in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bonoli, P.T.; OShea, P.; Brambilla, M.; Golovato, S.N.; Hubbard, A.E.; Porkolab, M.; Takase, Y.; Boivin, R.L.; Bombarda, F.; Christensen, C.; Fiore, C.L.; Garnier, D.; Goetz, J.; Granetz, R.; Greenwald, M.; Horne, S.F.; Hutchinson, I.H.; Irby, J.; Jablonski, D.; LaBombard, B.; Lipschultz, B.; Marmar, E.; May, M.; Mazurenko, A.; McCracken, G.; Nachtrieb, R.; Niemczewski, A.; Ohkawa, H.; Pappas, D.A.; Reardon, J.; Rice, J.; Rost, C.; Schachter, J.; Snipes, J.A.; Stek, P.; Takase, K.; Terry, J.; Wang, Y.; Watterson, R.L.; Welch, B.; Wolfe, S.M. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1997-05-01

    Highly localized direct electron heating [full width at half-maximum (FWHM){congruent}0.2a] via mode converted ion Bernstein waves has been observed in the Alcator C-Mod Tokamak [I. H. Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)]. Electron heating at or near the plasma center (r/a{ge}0.3) has been observed in H({sup 3}He) discharges at B{sub 0}=(6.0{endash}6.5)T and n{sub e}(0){congruent}1.8{times}10{sup 20}m{sup {minus}3}. [Here, the minority ion species is indicated parenthetically.] Off-axis heating (r/a{ge}0.5) has also been observed in D({sup 3}He) plasmas at B{sub 0}=7.9T. The concentration of {sup 3}He in these experiments was in the range of n{sub 3{sub He}}/n{sub e}{congruent}(0.2{endash}0.3) and the locations of the mode conversion layer and electron heating peak could be controlled by changing the {sup 3}He concentration or toroidal magnetic field (B{sub 0}). The electron heating profiles were deduced using a rf modulation technique. Detailed comparisons with one-dimensional and toroidal full-wave models in the ion cyclotron range of frequencies have been carried out. One-dimensional full-wave code predictions were found to be in qualitative agreement with the experimental results. Toroidal full-wave calculations indicated the importance of volumetric and wave focusing effects in the interpretation of the experimental results. {copyright} {ital 1997 American Institute of Physics.}

  17. Lithium pellet injection experiments on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    A pellet enhanced performance mode, showing significantly reduced core transport, is regularly obtained after the injection of deeply penetrating lithium pellets into Alcator C-Mod discharges. These transient modes, which typically persist about two energy confinement times, are characterized by a steep pressure gradient (ell p ≤ a/5) in the inner third of the plasma, indicating the presence of an internal transport barrier. Inside this barrier, particle and energy diffusivities are greatly reduced, with ion thermal diffusivity dropping to near neoclassical values. Meanwhile, the global energy confinement time shows a 30% improvement over ITER89-P L-mode scaling. The addition of ICRF auxiliary heating shortly after the pellet injection leads to high fusion reactivity with neutron rates enhanced by an order of magnitude over L-mode discharges with similar input powers. A diagnostic system for measuring equilibrium current density profiles of tokamak plasmas, employing high speed lithium pellets, is also presented. Because ions are confined to move along field lines, imaging the Li+ emission from the toroidally extended pellet ablation cloud gives the direction of the magnetic field. To convert from temporal to radial measurements, the 3-D trajectory of the pellet is determined using a stereoscopic tracking system. These measurements, along with external magnetic measurements, are used to solve the Grad-Shafranov equation for the magnetic equilibrium of the plasma. This diagnostic is used to determine the current density profile of PEP modes by injection of a second pellet during the period of good confinement. This measurement indicates that a region of reversed magnetic shear exists at the plasma core. This current density profile is consistent with TRANSP calculations for the bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasmas are near the n = ∞ and the n = 1 marginal stability limits

  18. Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

    2015-09-15

    The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

  19. ICRF antenna matching system with ferrite tuners for the Alcator C-Mod tokamak

    Science.gov (United States)

    Lin, Y.; Binus, A.; Wukitch, S. J.; Koert, P.; Murray, R.; Pfeiffer, A.

    2015-12-01

    Real-time fast ferrite tuning (FFT) has been successfully implemented on the ICRF antennas on Alcator C-Mod. The former prototypical FFT system on the E-port 2-strap antenna has been upgraded using new ferrite tuners that have been designed specifically for the operational parameters of the Alcator C-Mod ICRF system (˜ 80 MHz). Another similar FFT system, with two ferrite tuners and one fixed-length stub, has been installed on the transmission line of the D-port 2-strap antenna. These two systems share a Linux-server-based real-time controller. These FFT systems are able to achieve and maintain the reflected power to the transmitters to less than 1% in real time during the plasma discharges under almost all plasma conditions, and help ensure reliable high power operation of the antennas. The innovative field-aligned (FA) 4-strap antenna on J-port has been found to have an interesting feature of loading insensitivity vs. plasma conditions. This feature allows us to significantly improve the matching for the FA J-port antenna by installing carefully designed stubs on the two transmission lines. The reduction of the RF voltages in the transmission lines has enabled the FA J-port antenna to deliver 3.7 MW RF power to plasmas out of the 4 MW source power in high performance I-mode plasmas.

  20. Scaling and transport analysis of divertor conditions on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    LaBombard, B.; Goetz, J.; Kurz, C.; Jablonski, D.; Lipschultz, B.; McCracken, G.; Niemczewski, A.; Boivin, R.L.; Bombarda, F.; Christensen, C.; Fairfax, S.; Fiore, C.; Garnier, D.; Graf, M.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Hutchinson, I.; Irby, J.; Kesner, J.; Luke, T.; Marmar, E.; May, M.; O`Shea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Schachter, J.; Snipes, J.; Stek, P.; Takase, Y.; Terry, J.; Tinios, G.; Watterson, R.; Welch, B.; Wolfe, S. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1995-06-01

    Detailed measurements and transport analysis of divertor conditions in Alcator C-Mod [Phys. Plasmas {bold 1}, 1511 (1994)] are presented for a range of line-averaged densities, 0.7{lt}{ital {bar n}}{sub {ital e}}{lt}2.2{times}10{sup 20} m{sup {minus}3}. Three parallel heat transport regimes are evident in the scrape-off layer: sheath-limited conduction, high-recycling divertor, and detached divertor, which can coexist in the same discharge. {ital Local} cross-field pressure gradients are found to scale simply with a {ital local} electron temperature. This scaling is consistent with classical electron parallel conduction being balanced by anomalous cross-field transport ({chi}{sub {perpendicular}}{similar_to}0.2 m{sup 2} s{sup {minus}1}) proportional to the local pressure gradient. A 60%--80% of divertor power is radiated in attached discharges, approaching 100% in detached discharges. Detachment occurs when the heat flux to the plate is low and the plasma pressure is high ({ital T}{sub {ital e}}{similar_to}5 eV). High neutral pressures in the divertor are nearly always present (1--20 mTorr), sufficient to remove parallel momentum via ion--neutral collisions.

  1. Pedestal Stability and Transport on the Alcator C-Mod Tokamak: Experiments in Support of Developing Predictive Capability

    International Nuclear Information System (INIS)

    Full text: New experimental data on the Alcator C-Mod tokamak are used to benchmark predictive modeling of the edge pedestal in various high-confinement regimes, contributing to a greater confidence in projection of pedestal height and width in ITER and reactors. Measurements in conventional Type-I ELMy H-mode have been used to test the theory of peeling-ballooning (PB) stability and pedestal structure predictions from the EPED model, which extends these theoretical comparisons to the highest pressure pedestals of any existing tokamak. Calculations with the ELITE code confirm that C-Mod ELMy H-modes operate near stability limits for ideal PB modes. Experimental C-Mod studies have provided supporting evidence for pedestal width scaling as the square root of poloidal beta at the pedestal top. This is the dependence that would be expected from theory if KBMs were responsible for limiting the pedestal width. The EPED model has been tested across an extended data on C-Mod, reproducing pedestal height and width reasonably well, and extending the tested range of EPED to within a factor of 3 of the absolute pedestal pressure targeted for ITER. In addition, C-Mod offers access to two regimes, enhanced D-alpha (EDA) H-mode and I- mode, that have high pedestals but in which large ELM activity is naturally suppressed and, instead, particle and impurity transport are regulated continuously. Significant progress has been made in both measuring and modeling pedestal fluctuations, transport and stability in these regimes. Pedestals of EDA H-mode and I-mode discharges are found to be ideal MHD stable, consistent with the general absence of ELM activity. Like ELITE, the BOUT++ code finds the EDA pedestal to be stable to ideal modes. However, it does identify finite growth rates for edge modes when realistic values of resistivity and diamagnetism are included. The result is consistent with the interpretation of the quasi-coherent mode (QCM), which is omnipresent in the EDA pedestal

  2. High resolution measurements of neutral density and ionization rate in the main chamber of the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R.L. E-mail: boivin@psfc.mit.edu; Goetz, J.; Hubbard, A.; Hughes, J.W.; Irby, J.; LaBombard, B.; Marmar, E.; Mossessian, D.; Terry, J.L

    2001-03-01

    Recent theoretical and experimental work has focused on the importance of neutrals in the edge dynamics of Alcator C-Mod plasmas. Two new high resolution detectors have been installed on the Alcator C-Mod tokamak to measure the neutral density and ionization rate near the edge of the discharge in the main chamber. The detectors consist of a 20 channel photodiode array which views the plasma tangentially 12.5 cm below the outer midplane, and 10 cm above the inner midplane, with a nominal radial resolution of 2 and 3 mm, respectively. The spectral bandwidth is limited by a filter with a very narrow band around the neutral deuterium Lyman alpha wavelength (1215 Angstrom). The local emissivity is then obtained via a standard Abel inversion of the absolutely calibrated brightness profile. Employing well-known branching ratios, and using measured local electron density and temperature, we therefore, infer the neutral density and ionization rate with similar radial resolution. We have observed that both Lyman alpha emissivity and ionization rate are usually peaked near the separatrix with a full width, half-maximum between 1 and 2 cm. The neutral density was found to drop rapidly with decreasing minor radius, from 2-3x10{sup 17}/m{sup 3} (5-10 mm outside the separatrix) to 1-2x10{sup 16}/m{sup 3} (5-10 mm inside the separatrix) for a line averaged density of 2.0x10{sup 20}/m{sup 3}. Variations in ionization rate and neutral density from low to high confinement mode (L and H mode) are also discussed.

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

    International Nuclear Information System (INIS)

    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-bare>1·1020[m-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 and 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 (Te = 5keV,Bt = 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.

  4. Axisymmetric Control in Alcator C-Mod

    Science.gov (United States)

    Tinios, Gerasimos

    1995-01-01

    This thesis investigates the degree to which linear axisymmetric modeling of the response of a tokamak plasma can reproduce observed experimental behavior. The emphasis is on the vertical instability. The motivation for this work lies in the fact that, once dependable models have been developed, modern control theory methods can be used to design feedback laws for more effective and efficient tokamak control. The models are tested against experimental data from the Alcator C-Mod tokamak. A linear model for each subsystem of the closed-loop system constituting an Alcator C-Mod discharge under feedback control has been constructed. A non-rigid, approximately flux-conserving, perturbed equilibrium plasma response model is used in the comparison to experiment. A detailed toroidally symmetric model of the vacuum vessel and the supporting superstructure is used. Modeling of the power supplies feeding the active coils has been included. Experiments have been conducted with vertically unstable plasmas where the feedback was turned off and the plasma response was observed in an open -loop configuration. The closed-loop behavior has been examined by injecting step perturbations into the desired vertical position of the plasma. The agreement between theory and experiment in the open-loop configuration was very satisfactory, proving that the perturbed equilibrium plasma response model and a toroidally symmetric electromagnetic model of the vacuum vessel and the structure can be trusted for the purpose of calculations for control law design. When the power supplies and the feedback computer hardware are added to the system, however, as they are in the closed-loop configuration, they introduce nonlinearities that make it difficult to explain observed behavior with linear theory. Nonlinear simulation of the time evolution of the closed-loop experiments was able to account for the discrepancies between linear theory and experiment. (Copies available exclusively from MIT Libraries

  5. Molybdenum emission from impurity-induced m= 1 snake-modes on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Bitter, M.; Gates, D.; Hill, K.; Pablant, N. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Granetz, R.; Reinke, M.; Podpaly, Y.; Rice, J. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sugiyama, L. [MIT - Laboratory for Nuclear Science, Cambridge, Massachusetts 02139 (United States)

    2012-10-15

    A suite of novel high-resolution spectroscopic imaging diagnostics has facilitated the identification and localization of molybdenum impurities as the main species during the formation and lifetime of m= 1 impurity-induced snake-modes on Alcator C-Mod. Such measurements made it possible to infer, for the first time, the perturbed radiated power density profiles from which the impurity density can be deduced.

  6. First results from Alcator-C-MOD

    Energy Technology Data Exchange (ETDEWEB)

    Hutchinson, I.H.; Boivin, R.; Bombarda, F.; Bonoli, P.; Fairfax, S.; Fiore, C.; Goetz, J.; Golovato, S.; Granetz, R.; Greenwald, M.; Horne, S.; Hubbard, A.; Irby, J.; LaBombard, B.; Lipschultz, B.; Marmar, E.; McCracken, G.; Porkolab, M.; Rice, J.; Snipes, J.; Takase, Y.; Terry, J.; Wolfe, S.; Christensen, C.; Garnier, D.; Graf, M.; Hsu, T.; Luke, T.; May, M.; Niemczewski, A.; Tinios, G.; Schachter, J.; Urbahn, J. (Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States))

    1994-05-01

    Early operation of the Alcator-C-MOD tokamak [I.H. Hutchinson, [ital Proceedings] [ital of] [ital IEEE] 13[ital th] [ital Symposium] [ital on] [ital Fusion] [ital Engineering], Knoxville, TN, edited by M. Lubell, M. Nestor, and S. Vaughan (Institute of Electrical and Electronic Engineers, New York, 1990), Vol. 1, p. 13] is surveyed. Reliable operation, with plasma current up to 1 MA, has been obtained, despite the massive conducting superstructure and the associated error fields. However, vertical disruptions are not slowed by the long vessel time constant. With pellet fueling, peak densities up to 9[times]10[sup 20] m[sup [minus]3] have been attained and snakes'' are often seen. Initial characterization of divertor and scrape-off layer is presented and indicates approximately Bohm diffusion. The edge plasma shows a wealth of marfe-like phenomena, including a transition to detachment from the divertor plates with accompanying radiative divertor regions. Energy confinement generally appears to exceed the expectations of neo-Alcator scaling. A transition to Ohmic H mode has been observed. Ion cyclotron heating experiments have demonstrated good power coupling, in agreement with theory.

  7. A Spatially Resolving X-ray Crystal Spectrometer for Measurement of Ion-temperature and Rotation-velocity Profiles on the AlcatorC-Mod Tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-03-24

    A new spatially resolving x-ray crystal spectrometer capable of measuring continuous spatial profiles of high resolution spectra (λ/dλ > 6000) of He-like and H-like Ar Kα lines with good spatial (~1 cm) and temporal (~10 ms) resolutions has been installed on the Alcator C-Mod tokamak. Two spherically bent crystals image the spectra onto four two-dimensional Pilatus II pixel detectors. Tomographic inversion enables inference of local line emissivity, ion temperature (Ti), and toroidal plasma rotation velocity (vφ) from the line Doppler widths and shifts. The data analysis techniqu

  8. Alcator C-MOD proposal addendum

    International Nuclear Information System (INIS)

    Since the design concept and overall purpose of the Alcator C-MOD device are similar to that proposed in October 1985, we have chosen in this document only to highlight areas where changes or additions have been made. Chapters in the Addendum correspond to those in the Proposal, except Chapter 9 which describes a number of toroidal improvement concepts which are being considered for inclusion in the Alcator C-MOD experimental program. A description of the redesign and a discussion of the objectives of the experimental program are given

  9. Impurity Screening in Ohmic and H-Mode Plasmas in the Alcator C-Mod Tokamak^*

    Science.gov (United States)

    McCracken, G. M.

    1996-11-01

    The impurity density in the confined plasma is determined not only by the impurity production rate but also by screening, i.e. the balance between the perpendicular and parallel transport in the SOL. The relative importance of screening has been studied by injecting gaseous recycling (Ne, Ar) and non-recycling impurities (N, C) into various poloidal positions of the SOL and divertor in C-Mod. The density of the non-recycling impurities in the core is a function of the poloidal position of injection, while the screening of recycling impurities is not. In both cases screening is significantly worse ( ~3x) during divertor detachment. For a given injection rate of N2 gas into H-mode discharges, the number of impurities in the core is typically a factor of 3 greater than for ohmic discharges. Optical imaging of low charge states of the injected non-recycling impurities using a camera and spectral filters shows a directed plume, indicating flow of impurities towards the divertor target at all positions studied, even at the inboard midplane. This implies that the friction force due to plasma flow dominates the parallel ion temperature gradient force. The spatial distribution of low charge states in the divertor has been studied using a multichord visible spectrometer, and the distribution of the nitrogen radiation has been been studied using a 20 chord bolometer array. The results show that for attached discharges the nitrogen radiation is predominantly in the SOL below the X-point. The position of the radiation is not strongly dependent on the position of nitrogen injection. The screening has been compared with calculations using the DIVIMP Monte Carlo code [1] and with an analytical model of the impurity transport. Work supported by U.S. DOE Contract No. DE-AC02-78ET51013. În collaboration with B Lipschultz, B LaBombard, J A Goetz, R Granetz, D Jablonski, H Ohkawa, J Terry, MIT, S Lisgo, P C Stangeby, University of Toronto, ^1P.C. Stangeby and D. Elder, J. Nucl. Mater

  10. The physics and engineering of Alcator C-Mod

    International Nuclear Information System (INIS)

    Alcator C-Mod is a new tokamak under construction at M.I.T. that promises to play an important and flexible role in the international fusion research effort. The physics and engineering features of the tokamak are described, giving an overview of the machine and plasma configurations. On the basis of empirical scaling laws, we predict the plasma confinement performance to be near DT equivalent breakeven. The planned experimental program is addressed to many of the vital physics questions still uncertain in high-performance tokamak plasma behaviour as well as to the investigation of innovative approaches to tokamak improvement. 17 refs., 17 figs., 3 tabs

  11. Lower hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

    Science.gov (United States)

    Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Terry, J. L.; Whyte, D. G.; Baek, S. G.; Edlund, E.; Hubbard, A. E.; Hughes, J. W.; Kuang, A. Q.; Reinke, M. L.; Shiraiwa, S.; Wallace, G. M.; Walk, J. R.

    2016-05-01

    For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt ( t 1.0 × 10 20 (m-3)). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivate the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch.

  12. Alcator C-MOD final safety analysis

    International Nuclear Information System (INIS)

    This document is designed to address the safety issues involved with the Alcator C-Mod project. This report will begin with a brief description of the experimental objectives which will be followed by information concerning the site. The Alcator C-Mod experiment is a pulsed fusion experiment in which a plasma formed from small amounts of hydrogen or deuterium gas is confined in a magnetic field for short periods (∼1 s). No radioactive fuels or fissile materials are used in the device, so that no criticality hazard exists and no credible nuclear accident can occur. During deuterium operation, the production of a small number of neutrons from a short pulse could result in a small amount of short- and intermediate-lived radioactive isotopes being produced inside the experimental cell. This report will demonstrate that this does not pose an additional hazard to the general population. The health and safety hazards resulting from Alcator C-Mod occur to the workers on the experiment, each of which is described in its own chapter with the steps taken to minimize the risk to employees. These hazards include fire, chemicals and cryogenics, air quality, electrical, electromagnetic radiation, ionizing radiation, and mechanical and natural phenomena. None of these hazards is unique to the facility, and methods of protection from them are well defined and are discussed in the chapter which describes each hazard. The quality assurance program, critical to ensuring the safety aspects of the program, will also be described

  13. Planning for US ion cyclotron heating research relevant to the Compact Ignition Tokamak and Alcator C-Mod

    International Nuclear Information System (INIS)

    Ion cyclotron heating (ICH) has been chosen as the primary method for providing auxiliary heating power to the plasma in the Compact Ignition Tokamak (CIT). Sustained progress in ion cyclotron range of frequencies (ICRF) heating experiments, together with supporting technology development, continues to justify selection of this technique as the preferred one for heating CIT to ignition. However, the CIT requirements are sufficiently different from existing achievements that continued experimentation and development are needed to meet the goals of the CIT experiment with a high degree of reliability. The purpose of this report is fourfold: (1) to review briefly the physics and technology research and development (R and D) needs for ICH on CIT, (2) to review the status of and planned programs for ICH on US and international machines, (3) to propose a unified ''mainline'' R and D program specifically geared to testing components for CIT, and (4) to assess the needs for experiments including C-Mod, the Tokamak Fusion Test Reactor (TFTR), and DIII-D to provide earlier information and improved probability of success for CIT ICH. 4 refs., 4 figs., 5 tabs

  14. Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique

  15. Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jablonski, D.F.

    1996-05-01

    A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique.

  16. ICRF heating antennas for Alcator C-Mod

    International Nuclear Information System (INIS)

    Initial ICRF coupling studies were performed successfully on the Alcator C-Mod tokamak with all-metallic first wall, using a radially movable single current strap (monopole) antenna. The plasma loading is 10-30 times the vacuum loading, which is sufficient to inject 2 MW of RF power through one two-strap (dipole) antenna. The observed loading is consistent with predictions of full-wave calculations. Up to 1 MW of RF power (antenna power density of > or approx. 10 MW/m2) has been injected into the C-Mod plasma with no increase in the fractional radiated power Prad/Pin. Definite signs of both ion and electron heating were observed at power levels of 0.4 MW and higher (PRF/POH > or approx. 0.3) in the D(H) minority heating regime. High power heating experiments with 4 MW of source power and two dipole antennas will begin in early 1994. (author)

  17. Transport experiments in Alcator-C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Greenwald, M.; Boivin, R.L.; Bonoli, P.; Christensen, C.; Fiore, C.; Garnier, D.; Goetz, J.; Golovato, S.; Graf, M.; Granetz, R.; Horne, S.; Hsu, T.; Hubbard, A.; Hutchinson, I.; Irby, J.; Kurz, C.; LaBombard, B.; Lipschultz, B.; Luke, T.; Marmar, E.; McCracken, G.; Niemczewski, A.; O`Shea, P.; Porkolab, M.; Rice, J.; Reardon, J.; Schachter, J.; Snipes, J.; Stek, P.; Takase, Y.; Terry, J.; Umansky, M.; Watterson, R.; Wolfe, S. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Bombarda, F. [ENEA-Frascati, Frascati (Italy); May, M. [Johns Hopkins University, Baltimore, Maryland 21218 (United States); Welch, B. [University of Maryland, College Park, Maryland 20742 (United States)

    1995-06-01

    A series of transport experiments has been carried out in Alcator-C-Mod. [Phys Plasmas {bold 1}, 1511 (1994)]. Data from both Ohmic and ICRF (ion cyclotron range of frequencies) heated plasmas can be fitted with an L-mode (low mode) scaling law. The Ohmic {tau}{sub {ital E}}`s show no scaling with density in any regime and can reach values of 2--3 times neo-Alcator. Impurity confinement has been studied with the laser blow-off technique with {tau}{sub {ital I}} showing nearly linear scaling with plasma current. Ohmic and ICRF H modes are obtained over a wide range of discharge parameters, extending the range in the international database for {ital nB}, by almost a factor of 10. The power threshold for ELM-free (edge localized mode) discharges is in rough agreement with the scaling {ital P}/{ital S}=0.044{ital nB}. Energy diffusivities of Ohmic and ICRF heated plasmas have been measured from local analysis of plasma profiles and power fluxes. The same analysis produces a value for plasma resistivity which lies between the Spitzer and neoclassical calculations. Analysis of plasma transients have yielded values for particle diffusivity and convection velocity.

  18. Wide-frequency range, dynamic matching network and power system for the “Shoelace” radio frequency antenna on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    A wide-frequency range (50–300 kHz) power system has been implemented for use with a new RF antenna – the “Shoelace” antenna – built to drive coherent plasma fluctuations in the edge of the Alcator C-Mod tokamak. A custom, dynamically tunable matching network allows two commercial 1 kW, 50-Ω RF amplifiers to drive the low-impedance, inductive load presented by the antenna. This is accomplished by a discretely variable L-match network, with 81 independently selected steps available for each of the series and parallel legs of the matching configuration. A compact programmable logic device provides a control system that measures the frequency with better than 1 kHz accuracy and transitions to the correct tuning state in less than 1 ms. At least 85% of source power is dissipated in the antenna across the operational frequency range, with a minimum frequency slew rate of 1 MHz/s; the best performance is achieved in the narrower band from 80 to 150 kHz which is of interest in typical experiments. The RF frequency can be run with open-loop control, following a pre-programmed analog waveform, or phase-locked to track a plasma fluctuation diagnostic signal in real time with programmable phase delay; the amplitude control is always open-loop. The control waveforms and phase delay are programmed remotely. These tools have enabled first-of-a-kind measurements of the tokamak edge plasma system response in the frequency range and at the wave number at which coherent fluctuations regulate heat and particle transport through the plasma boundary

  19. Wide-frequency range, dynamic matching network and power system for the "Shoelace" radio frequency antenna on the Alcator C-Mod tokamak.

    Science.gov (United States)

    Golfinopoulos, Theodore; LaBombard, Brian; Burke, William; Parker, Ronald R; Parkin, William; Woskov, Paul

    2014-04-01

    A wide-frequency range (50-300 kHz) power system has been implemented for use with a new RF antenna - the "Shoelace" antenna - built to drive coherent plasma fluctuations in the edge of the Alcator C-Mod tokamak. A custom, dynamically tunable matching network allows two commercial 1 kW, 50-Ω RF amplifiers to drive the low-impedance, inductive load presented by the antenna. This is accomplished by a discretely variable L-match network, with 81 independently selected steps available for each of the series and parallel legs of the matching configuration. A compact programmable logic device provides a control system that measures the frequency with better than 1 kHz accuracy and transitions to the correct tuning state in less than 1 ms. At least 85% of source power is dissipated in the antenna across the operational frequency range, with a minimum frequency slew rate of 1 MHz/s; the best performance is achieved in the narrower band from 80 to 150 kHz which is of interest in typical experiments. The RF frequency can be run with open-loop control, following a pre-programmed analog waveform, or phase-locked to track a plasma fluctuation diagnostic signal in real time with programmable phase delay; the amplitude control is always open-loop. The control waveforms and phase delay are programmed remotely. These tools have enabled first-of-a-kind measurements of the tokamak edge plasma system response in the frequency range and at the wave number at which coherent fluctuations regulate heat and particle transport through the plasma boundary.

  20. Comparison of tungsten nano-tendrils grown in Alcator C-Mod and linear plasma devices

    NARCIS (Netherlands)

    Wright, G. M.; D. Brunner,; Baldwin, M. J.; Bystrov, K.; Doerner, R. P.; Labombard, B.; Lipschultz, B.; De Temmerman, G.; J.L. Terry,; Whyte, D. G.; Woller, K.B.

    2013-01-01

    Growth of tungsten nano-tendrils (“fuzz”) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a la

  1. Edge Minority Heating Experiment in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben; J.L. Terry; P. Bonoli; R. Budny; C.S. Chang; C. Fiore; G. Schilling; S. Wukitch; J. Hughes; Y. Lin; R. Perkins; M. Porkolab; the Alcator C-Mod Team

    2005-03-25

    An attempt was made to control global plasma confinement in the Alcator C-Mod tokamak by applying ion cyclotron resonance heating (ICRH) power to the plasma edge in order to deliberately create a minority ion tail loss. In theory, an edge fast ion loss could modify the edge electric field and so stabilize the edge turbulence, which might then reduce the H-mode power threshold or improve the H-mode barrier. However, the experimental result was that edge minority heating resulted in no improvement in the edge plasma parameters or global stored energy, at least at power levels of radio-frequency power is less than or equal to 5.5 MW. A preliminary analysis of these results is presented and some ideas for improvement are discussed.

  2. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  3. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak.

    Science.gov (United States)

    Brunner, D; Burke, W; Kuang, A Q; LaBombard, B; Lipschultz, B; Wolfe, S

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  4. Disruptions and halo currents in Alcator C-Mod

    Science.gov (United States)

    Granetz, R. S.; Hutchinson, I. H.; Sorci, J.; Irby, J. H.; La Bombard, B.; Gwinn, D.

    1996-05-01

    Disruptions in Alcator C-Mod can generate large eddy currents in the highly conducting vacuum vessel and internal structures, including a significant poloidal component due to halo currents. In order to understand better the stresses arising from the resulting J*B forces, Alcator C-Mod has been fitted with a comprehensive set of sensors to measure the spatial distribution and temporal behaviour of the halo currents. It is found that they are toroidally asymmetric, with a typical peaking factor of 2. The asymmetric pattern usually rotates toroidally at a few kilohertz, thus ruling out first wall non-uniformities as the cause of the asymmetry. Analysis of the information compiled in the C-Mod disruption database indicates that the maximum halo current during a disruption scales roughly as either Ip2/Bphi or Ip/q95, but that there is a large amount of variation that is not yet understood

  5. Alcator C-Mod ICRF antenna and matching circuit

    International Nuclear Information System (INIS)

    Alcator C-Mod will be a compact, high field, high density, divertor tokamak. Two FMIT transmitters will supply 4 MW of power in 1 sec pulses at 80 MHz for ICRF heating. Fast wave minority heating experiments are planned in D(3He) at 8 T and D(H) at 5.5 T. The first antenna will have a single current strap inside a box structure, which will be movable radially. The antenna will be inserted through a side port, making the rf power density on the antenna surface ∼2 kW/cm2 at 2 MW. The antenna will be center-tapped for mechanical strength and have a double layer Faraday screen tilted along the field lines. The antenna geometry was chosen to maximize power coupling assuming voltage-limited operation. A wide antenna with slotted box sides appears the best design, and 10 Ω of loading is required to couple 2 MW of power at a voltage limit of 40 kV. Matching is achieved by choice of the drive point to a resonant circuit formed by the antenna and a loop of transmission line outside of the vacuum and by tuning elements in the transmission line to the transmitter. 6 refs., 4 figs

  6. Overview of the Alcator C-Mod Research Program

    Science.gov (United States)

    Marmar, E.; Bader, A.; Bakhtiari, M.; Barnard, H.; Beck, W.; Bespamyatnov, I.; Binus, A.; Bonoli, P.; Bose, B.; Bitter, M.; Cziegler, I.; Dekow, G.; Dominguez, A.; Duval, B.; Edlund, E.; Ernst, D.; Ferrara, M.; Fiore, C.; Fredian, T.; Graf, A.; Granetz, R.; Greenwald, M.; Grulke, O.; Gwinn, D.; Harrison, S.; Harvey, R.; Hender, T. C.; Hosea, J.; Hill, K.; Howard, N.; Howell, D. F.; Hubbard, A.; Hughes, J. W.; Hutchinson, I.; Ince-Cushman, A.; Irby, J.; Izzo, V.; Kanojia, A.; Kessel, C.; Ko, J. S.; Koert, P.; La Bombard, B.; Lau, C.; Lin, L.; Lin, Y.; Lipschultz, B.; Liptac, J.; Ma, Y.; Marr, K.; May, M.; McDermott, R.; Meneghini, O.; Mikkelsen, D.; Ochoukov, R.; Parker, R.; Phillips, C. K.; Phillips, P.; Podpaly, Y.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Scott, S.; Schmidt, A.; Sears, J.; Shiraiwa, S.; Sips, A.; Smick, N.; Snipes, J.; Stillerman, J.; Takase, Y.; Terry, D.; Terry, J.; Tsujii, N.; Valeo, E.; Vieira, R.; Wallace, G.; Whyte, D.; Wilson, J. R.; Wolfe, S.; Wright, G.; Wright, J.; Wukitch, S.; Wurden, G.; Xu, P.; Zhurovich, K.; Zaks, J.; Zweben, S.

    2009-10-01

    This paper summarizes highlights of research results from the Alcator C-Mod tokamak covering the period 2006-2008. Active flow drive, using mode converted ion cyclotron waves, has been observed for the first time in a tokamak plasma, using a mix of D and 3He ion species; toroidal and poloidal flows are driven near the location of the mode conversion layer. ICRF induced edge sheaths are implicated in both the erosion of thin boron coatings and the generation of metallic impurities. Lower hybrid range of frequencies (LHRF) microwaves have been used for efficient current drive, current profile modification and toroidal flow drive. In addition, LHRF has been used to modify the H-mode pedestal, increasing temperature, decreasing density and lowering the pedestal collisionality. Studies of hydrogen isotope retention in solid metallic plasma facing components reveal significantly higher retention than expected from ex situ laboratory studies; a model to explain the results, based on plasma/neutral induced lattice damage, has been developed and tested. During gas-puff mitigation of disruptions, induced MHD instabilities cause the magnetic field to become stochastic, resulting in reduction of halo currents, spreading of plasma power loading and loss of runaway electrons before they cause damage. Detailed pedestal rotation profile measurements have been used to infer Er profiles, and correlation with global H-mode confinement. An improved L-mode regime, obtained at q95 <= 3 with ion drift away from the active X-point, shows very good energy confinement with a strong temperature pedestal, a weak density pedestal, and no evidence of particle or impurity accumulation, without the need for ELMs or any additional edge density regulation mechanism.

  7. ICRF heating experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Y.; Bonoli, P.T.; Hubbard, A.; Mazurenko, A.; OShea, P.J.; Porkolab, M.; Reardon, J.; Wukitch, S.; Boivin, R. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Bombarda, F. [Associazione EURATOM-ENEA sulla Fusione, Frascati 00044 (Italy); Fiore, C.; Garnier, D.; Goetz, J.A.; Granetz, R.; Greenwald, M. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Hartmann, D. [Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Hutchinson, I.H.; Irby, J.; LaBombard, B.; Lipschultz, B.; Marmar, E. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); May, M. [Department of Physics, The John Hopkins University, Baltimore, Maryland 21218 (United States); Rice, J.; Rost, J.C.; Schachter, J.; Snipes, J.A.; Stek, P.; Terry, J. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Welch, B. [Institute for Plasma Research, University of Maryland, College Park, Maryland 20742 (United States); Wolfe, S. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States)

    1997-04-01

    Routine high power operation of the ICRF heating system (up to 3.5 MW at 80 MHz) has enabled studies of enhanced confinement modes as well as high heat flux divertor experiments in the Alcator C-Mod tokamak with toroidal magnetic fields of up to 8 T. H-mode was routinely observed when the edge temperature exceeded a threshold value. Boronization has reduced the radiated power to approximately 30{percent} of the input power, which had little effect on confinement of L-mode plasmas, but had a large impact on the performance of H-mode plasmas. It has become possible to achieve quasi-steady-state H-modes with H{approx_equal}2 and {beta}{sub N}{approx_equal}1.5 simultaneously with P{sub rad}/P{sub in}{approx_equal}0.3 and Z{sub eff}{approx_equal}1.5. PEP mode can be obtained with central ICRF heating combined with core fuelling by pellet injection. Because of the high central density, ion heating becomes the dominant heating channel during PEP mode. For direct electron heating schemes, such as heating by the mode converted ion Bernstein wave, the electron heating profile can be measured using the break-in-slope analysis of the electron temperature at rf power transitions. Mode conversion heating produced highly localized electron heating profiles, both on-axis and off-axis. Recent developments in full-wave codes have improved the agreement between the observed experimental results and the theoretically calculated power absorption profiles and power partition between ions and electrons. {copyright} {ital 1997 American Institute of Physics.}

  8. Overview of Alcator C-Mod Research

    Science.gov (United States)

    White, Anne

    2015-11-01

    Research on C-Mod supports next-step-devices: RF heating, current and flow drive, divertor/PMI physics, non-ELMing regimes with enhanced confinement, and disruption mitigation/runaway dynamics. Disruption mitigation experiments in MHD-unstable plasmas show MGI works equally well with and without locked modes. The L-I-mode threshold is found to be independent of magnetic field, opening an expanded operating range at high field. The toroidal and radial structure of power deposition of RF waves into the edge plasma has been systematically quantified, through the use of a unique set of fast time resolution edge diagnostics. Progress in understanding multi-channel core transport has been significant. Full-physics, ITG/TEM/ETG gyrokinetic simulations show that nonlinear cross-scale coupling enhances both ion and electron heat flux to match experiments, explaining the origin of electron heat flux and stiffness. Dynamic, passive measurements of the core rotation velocity profiles with X-ray imaging crystal spectroscopy show the direction of intrinsic rotation reversals depends on central safety factor, not on the magnetic shear. Design studies for ADX and SPARC are establishing the engineering, economics and physics for a fusion energy development path leveraging new superconducting magnet technologies. This work is supported by the US DOE under DE- FC02-99ER54512-CMOD.

  9. Overview of the Alcator C-MOD research programme

    Science.gov (United States)

    Scott, S.; Bader, A.; Bakhtiari, M.; Basse, N.; Beck, W.; Biewer, T.; Bernabei, S.; Bonoli, P.; Bose, B.; Bravenec, R.; Bespamyatnov, I.; Childs, R.; Cziegler, I.; Doerner, R.; Edlund, E.; Ernst, D.; Fasoli, A.; Ferrara, M.; Fiore, C.; Fredian, T.; Graf, A.; Graves, T.; Granetz, R.; Greenough, N.; Greenwald, M.; Grimes, M.; Grulke, O.; Gwinn, D.; Harvey, R.; Harrison, S.; Hender, T. C.; Hosea, J.; Howell, D. F.; Hubbard, A. E.; Hughes, J. W.; Hutchinson, I.; Ince-Cushman, A.; Irby, J.; Jernigan, T.; Johnson, D.; Ko, J.; Koert, P.; La Bombard, B.; Kanojia, A.; Lin, L.; Lin, Y.; Lipschultz, B.; Liptac, J.; Lynn, A.; MacGibbon, P.; Marmar, E.; Marr, K.; May, M.; Mikkelsen, D. R.; McDermott, R.; Parisot, A.; Parker, R.; Phillips, C. K.; Phillips, P.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Sampsell, M.; Schilling, G.; Schmidt, A.; Smick, N.; Smirnov, A.; Snipes, J.; Stotler, D.; Stillerman, J.; Tang, V.; Terry, D.; Terry, J.; Ulrickson, M.; Vieira, R.; Wallace, G.; Whyte, D.; Wilson, J. R.; Wright, G.; Wright, J.; Wolfe, S.; Wukitch, S.; Wurden, G.; Yuh, H.; Zhurovich, K.; Zaks, J.; Zweben, S.

    2007-10-01

    Alcator C-MOD has compared plasma performance with plasma-facing components (PFCs) coated with boron to all-metal PFCs to assess projections of energy confinement from current experiments to next-generation burning tokamak plasmas. Low-Z coatings reduce metallic impurity influx and diminish radiative losses leading to higher H-mode pedestal pressure that improves global energy confinement through profile stiffness. RF sheath rectification along flux tubes that intersect the RF antenna is found to be a major cause of localized boron erosion and impurity generation. Initial lower hybrid current drive (LHCD) experiments (PLH < 900 kW) in preparation for future advanced-tokamak studies have demonstrated fully non-inductive current drive at Ip ~ 1.0 MA with good efficiency, Idrive = 0.4 PLH/neoR (MA, MW, 1020 m-3,m). The potential to mitigate disruptions in ITER through massive gas-jet impurity puffing has been extended to significantly higher plasma pressures and shorter disruption times. The fraction of total plasma energy radiated increases with the Z of the impurity gas, reaching 90% for krypton. A positive major-radius scaling of the error field threshold for locked modes (Bth/B ~ R0.68±0.19) is inferred from its measured variation with BT that implies a favourable threshold value for ITER. A phase contrast imaging diagnostic has been used to study the structure of Alfvén cascades and turbulent density fluctuations in plasmas with an internal transport barrier. Understanding the mechanisms responsible for regulating the H-mode pedestal height is also crucial for projecting performance in ITER. Modelling of H-mode edge fuelling indicates high self-screening to neutrals in the pedestal and scrape-off layer (SOL), and reproduces experimental density pedestal response to changes in neutral source, including a weak variation of pedestal height and constant width. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as I_p^2

  10. Initial ICRF coupling and heating experiments on Alcator C-Mod

    International Nuclear Information System (INIS)

    Initial ICRF coupling studies were performed successfully on the Alcator C-Mod tokamak with all-metallic first wall, using a radially movable single-current-strap (monopole) antenna. The plasma loading is 10-40 times the vacuum loading, which is sufficient to inject 2MW of r.f. power through one two-strap (dipole) antenna. The observed loading is consistent with predictions of full-wave calculations. Up to 1MW of r.f. power (antenna power density of XXXX10MWm-2) has been injected into the C-Mod plasma without serious impurity problems. Definite signs of both ion and electron heating were observed at power levels of 0.4MW and higher (PRF/POHXXXX0.3) in the D(H) minority heating regime. High-power heating experiments with 4MW of source power and two dipole antennas began in May 1994. ((orig.))

  11. Modeling of Alcator C-Mod Divertor Baffling Experiments

    Energy Technology Data Exchange (ETDEWEB)

    D. P. Stotler; C. S. Pitcher; C. J. Boswell; T. K. Chung; B. LaBombard; B. Lipschultz; J. L. Terry; R. J. Kanzleiter

    2000-11-29

    A specific Alcator C-Mod discharge from the series of divertor baffling experiments is simulated with the DEGAS 2 Monte Carlo neutral transport code. A simple two-point plasma model is used to describe the plasma variation between Langmuir probe locations. A range of conductances for the bypass between the divertor plenum and the main chamber are considered. The experimentally observed insensitivity of the neutral current flowing through the bypass and of the D alpha emissions to the magnitude of the conductance is reproduced. The current of atoms in this regime is being limited by atomic physics processes and not the bypass conductance. The simulated trends in the divertor pressure, bypass current, and D alpha emission agree only qualitatively with the experimental measurements, however. Possible explanations for the quantitative differences are discussed.

  12. Correlation ECE diagnostic in Alcator C-Mod

    International Nuclear Information System (INIS)

    Correlation ECE (CECE) is a diagnostic technique that allows measurement of small amplitude electron temperature, Te, fluctuations through standard cross-correlation analysis methods. In Alcator C-Mod, a new CECE diagnostic has been installed[Sung RSI 2012], and interesting phenomena have been observed in various plasma conditions. We find that local Te fluctuations near the edge (ρ ~ 0:8) decrease across the linearto- saturated ohmic confinement transition, with fluctuations decreasing with increasing plasma density [Sung NF 2013], which occurs simultaneously with rotation reversals [Rice NF 2011]. Te fluctuations are also reduced across core rotation reversals with an increase of plasma density in RF heated L-mode plasmas, which implies that the same physics related to the reduction of Te fluctuations may be applied to both ohmic and RF heated L-mode plasmas. In I-mode plasmas, we observe the reduction of core Te fluctuations, which indicates changes of turbulence occur not only in the pedestal region but also in the core across the L/I transition [White NF 2014]. The present CECE diagnostic system in C-Mod and these experimental results are described in this paper

  13. Implementation of LHCD Experiments on Alcator C-Mod

    Science.gov (United States)

    Parker, R.; Basse, N.; Beck, W.; Bernabei, S.; Childs, R.; Ellis, R.; Fredd, E.; Greenough, N.; Grimes, M.; Gwinn, D.; Hosea, J.; Irby, J.; Koert, P.; Kung, C. C.; Labombard, B.; Liptac, J.; Loesser, G. D.; Marmar, E.; Schilling, G.; Terry, D.; Terry, J.; Vieira, R.; Wallace, G.; Wilson, J. R.; Zaks, J.

    2005-09-01

    An antenna-transmitter system for driving current in the LHRF has been installed in Alcator C-Mod. The antenna is a grill consisting of 4 poloidal rows of waveguides, each with 24 guides in the toroidal direction. Power is supplied by 12 klystrons capable of 250 kW operation at a frequency of 4.6 GHz. Thus the total source power is 3 MW, with about 1.5 MW available to be coupled to the plasma. Power supply and heat throughput limits in C-Mod limit the pulse length to 5 s, which however represents several current redistribution times. With 90° phasing, the n∥ spectrum is sharply peaked at 2.3 and the range 1.5 < n∥ < 3.5 can be accessed dynamically by varying the phase of the klystrons. The system is in the commissioning phase with klystron power limited to ˜20 kW and pulse length to 10 ms. Early results from plasma operation are discussed.

  14. Non-axisymmetric Field Effects on Alcator C-Mod

    Science.gov (United States)

    Wolfe, S.; Hutchinson, I.; Granetz, R.; Rice, J.; Hubbard, A.; Irby, J.; Vieira, R.; Cochran, W.; Gwinn, D.; Rosati, J.; Lynn, A.

    2003-10-01

    A set of coils capable of producing non-axisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode (m) spectra has been installed on Alcator C-Mod. This coilset has been used to suppress locked modes during low density or high current operation and also to induce locked modes in normally stable configurations in order to study error field effects. Locked modes are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement. The inferred value of the threshold perturbation for producing a locked mode is of order B_21/B_T ˜ 10-4, where B_21 is the helically resonant m/n=2/1 field evaluated at the q=2 surface. This value is comparable to extrapolations based on experiments on JET and DIII-D, but is inconsistent with stronger BT and size scaling inferred from Compass-D results(R. J. Buttery, et al., 17th Fusion Energy Conference, Oct. 1998, Yokohama (IAEA-CN-69) EX8/5). The C-Mod result therefore has favorable implications for the locked mode threshold in ITER.

  15. Correlation ECE diagnostic in Alcator C-Mod

    Directory of Open Access Journals (Sweden)

    Sung C.

    2015-01-01

    electron temperature, Te, fluctuations through standard cross-correlation analysis methods. In Alcator C-Mod, a new CECE diagnostic has been installed[Sung RSI 2012], and interesting phenomena have been observed in various plasma conditions. We find that local Te fluctuations near the edge (ρ ~ 0:8 decrease across the linearto- saturated ohmic confinement transition, with fluctuations decreasing with increasing plasma density[Sung NF 2013], which occurs simultaneously with rotation reversals[Rice NF 2011]. Te fluctuations are also reduced across core rotation reversals with an increase of plasma density in RF heated L-mode plasmas, which implies that the same physics related to the reduction of Te fluctuations may be applied to both ohmic and RF heated L-mode plasmas. In I-mode plasmas, we observe the reduction of core Te fluctuations, which indicates changes of turbulence occur not only in the pedestal region but also in the core across the L/I transition[White NF 2014]. The present CECE diagnostic system in C-Mod and these experimental results are described in this paper.

  16. Long Term Retention of Deuterium and Tritium in Alcator C-Mod

    International Nuclear Information System (INIS)

    We estimate the total in-vessel deuterium retention in Alcator C-Mod from a run campaign of about 1090 plasmas. The estimate is based on measurements of deuterium retained on 22 molybdenum tiles from the inner wall and divertor. The areal density of deuterium on the tiles was measured by nuclear reaction analysis. From these data, the in-vessel deuterium inventory is estimated to be about 0.1 gram, assuming the deuterium coverage is toroidally symmetric. Most of the retained deuterium is on the walls of the main plasma chamber, only about 2.5% of the deuterium is in the divertor. The D coverage is consistent with a layer saturated by implantation with ions and charge-exchange neutrals from the plasma. This contrasts with tokamaks with carbon plasma-facing components (PFC's) where long-term retention of tritium and deuterium is large and mainly in the divertor due to codeposition with carbon eroded by the plasma. The low deuterium retention in the C-Mod divertor is mainly due to the absence of carbon PFC's in C-Mod and the low erosion rate of Mo

  17. Extended pulse-length operation of Alcator C-Mod

    Science.gov (United States)

    Wolfe, S.; Irby, J.; Terry, J.; Labombard, B.; Wukitch, S.; Marmar, E.; Cochran, W.; Dekow, G.; Gwinn, D.; Maqueda, R.

    2001-10-01

    The C-Mod Advanced Tokamak program depends on the unique capability of the C-Mod facility to operate with plasma pulse lengths corresponding to multiple skin times with high performance parameters. Specifically, pulse lengths with current and toroidal field flattops of order 5 seconds, with toroidal field of 4 tesla, are proposed. In the case of the AT program, these plasmas would have current sustained non-inductively, i.e. by a combination of RF (lower hybrid) current drive and pressure-driven current. Experiments during the 2001 experimental campaign will extend the plasma pulse length to the maximum possible with only inductive current drive. The purpose of these experiments is to test and demonstrate the long-pulse capability of the coils, power system, control system, etc., and to test power and particle handling performance under long pulse conditions. In support of the latter goal, we will benchmark divertor surface heating during medium-power operation and assess the effectiveness of X-point sweeping and N2 puffing for dissipating the divertor heat loads. Results of these experiments will be presented.

  18. High confinement dissipative divertor operation on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Goetz, J.A.; LaBombard, B.; Lipschultz, B.; Pitcher, C.S.; Terry, J.L.; Boswell, C.; Gangadhara, S.; Pappas, D.; Weaver, J.; Welch, B.; Boivin, R.L.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hubbard, A.; Hutchinson, I.; Irby, J.; Marmar, E.; Mossessian, D.; Porkolab, M.; Rice, J.; Rowan, W.L.; Schilling, G.; Snipes, J.; Takase, Y.; Wolfe, S.; Wukitch, S. [Plasma Science Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1999-05-01

    Alcator C-Mod [I. H. Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)] has operated a High-confinement-mode (H-mode) plasma together with a dissipative divertor and low core Z{sub eff}. The initially attached plasma is characterized by steady-state enhancement factor, H{sub ITER89P} [P. N. Yushmanov {ital et al.}, Nucl. Fusion {bold 30}, 1999 (1990)], of 1.9, central Z{sub eff} of 1.1, and a radiative fraction of {approximately}50{percent}. Feedback control of a nitrogen gas puff is used to increase radiative losses in both the core/edge and divertor plasmas in almost equal amounts. Simultaneously, the core plasma maintains H{sub ITER89P} of 1.6 and Z{sub eff} of 1.4 in this nearly 100{percent} radiative state. The power and particle flux to the divertor plates have been reduced to very low levels while the core plasma is relatively unchanged by the dissipative nature of the divertor. {copyright} {ital 1999 American Institute of Physics.}

  19. Study of the Effects of Neutrals in Alcator C-Mod Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R L; Boswell, C; Goetz, J A; Hubbard, A E; Irby, J; LaBombard, B; Marmar, E S; Mossessian, D; Owen, L W; Pitcher, C S; Terry, J L

    1999-06-14

    Recently, much effort has been dedicated to understanding the bifurcation involved in the transition from a low to high confinement regime. While several theories have been brought forward, many factors remain to be elucidated, one of which involves the role played by neutral particles in the evolution of a transport barrier near the edge of the plasma. Alcator C-Mod is especially well suited for the study of neutral particle effects, mainly because of its high plasma and neutral densities, and closed divertor geometry. Alcator C-Mod employs ICRF as auxiiiary heating for obtaining a high confinement regime, although ohmic H-modes are routinely obtained as well. The neutrals can enter the edge dynamics through the particle, momentum and energy balance. In the particle balance, the source of neutrals has to be evaluated vis-8-vis the formation of the edge density pedestal. It is widely believed that plasma rotation is an important factor in reducing transport. In this case, neutrals could act as a momentum sink, through the charge-exchange process. That same process can also modify the energy balance of the plasma near the edge by increasing the cross-field heat flux. These effects are quite difficult to measure experimentally, in large part because neutral particle diagnosis is not an easy task, and because of the inherent 3-dimensional aspect of the problem. Consequently, the neutral's spatial and energy distributions are usually not well known. In Alcator C-Mod, we recently implemented a series of diagnostics for the purpose of measuring these distributions. They include measurements of the neutral pressure at many locations around the tokamak, and spatially resolved measurements of Lyman-a and charge-exchange power emission. A high-resolution multichord (20 channels) tangential view of neutral deuterium emission (Lyman-a) has been recently installed near the midplane. The viewing area covers approximately 4 cm across the separatrix, with a nominal 2 mm

  20. Neutral particle analysis of ICRF heated discharges on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, W D; Boivin, R L; Bonoli, P T; Fiore, C L; Hubbard, A; Irby, J; Nelson-Melby, E; Porkolab, M; Wukitch, S J [NW17-121 MIT-PSFC Massachusetts Institute of Technology, 175 Albany Street, Cambridge, MA 02139 (United States)

    2003-08-01

    A neutral particle analyser (NPA) has been used to make measurements of the ion distribution in the Alcator C-Mod tokamak. We have used the analyser to measure the energy distribution of majority deuterons and minority hydrogen (up to 20 keV) at 0.4{<=}r/a{<=}0.6, during minority ion cyclotron range of frequencies (ICRF) heating. These energy spectra were also simulated by using the predicted minority ion distribution from a combined Fokker Planck ICRF code in a model for the neutral particle flux. The model predictions for neutral flux were found to be in qualitative agreement with the measured energy spectra. The energy spectra for two different ICRF antenna configurations were also measured and found to be very similar.

  1. Divertor heat flux footprints in EDA H-mode discharges on Alcator C-Mod

    International Nuclear Information System (INIS)

    The physics that sets the width of the power exhaust channel in a tokamak scrape-off layer and its scaling with engineering parameters is of fundamental importance for reactor design, yet it remains to be understood. An extensive array of divertor heat flux diagnostics was recently commissioned in Alcator C-Mod with the aim of improving our understanding. Initial results are reported from EDA H-mode discharges in which plasma current, input power, toroidal field and magnetic topology were varied. The integral width of the outer divertor heat flux footprint is found to lie in the range of 3-5 mm mapped to the mid-plane. Widths are insensitive to single versus double-null topology and the magnitude of toroidal field. Pedestal physics appears to largely determine these widths; a dependence of width on plasma thermal energy is noted, yielding a reduction in width as plasma current is increased for the best EDA H-modes.

  2. Divertor heat flux footprints in EDA H-mode discharges on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    LaBombard, B., E-mail: labombard@psfc.mit.edu [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Terry, J.L.; Hughes, J.W.; Brunner, D.; Payne, J.; Reinke, M.L.; Lin, Y.; Wukitch, S. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2011-08-01

    The physics that sets the width of the power exhaust channel in a tokamak scrape-off layer and its scaling with engineering parameters is of fundamental importance for reactor design, yet it remains to be understood. An extensive array of divertor heat flux diagnostics was recently commissioned in Alcator C-Mod with the aim of improving our understanding. Initial results are reported from EDA H-mode discharges in which plasma current, input power, toroidal field and magnetic topology were varied. The integral width of the outer divertor heat flux footprint is found to lie in the range of 3-5 mm mapped to the mid-plane. Widths are insensitive to single versus double-null topology and the magnitude of toroidal field. Pedestal physics appears to largely determine these widths; a dependence of width on plasma thermal energy is noted, yielding a reduction in width as plasma current is increased for the best EDA H-modes.

  3. Neutral Transport Simulations of Gas Puff Imaging Experiments on Alcator C-Mod

    International Nuclear Information System (INIS)

    Visible imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results.Visibl e imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results

  4. Measurement of particle transport coefficients on Alcator C-Mod

    International Nuclear Information System (INIS)

    The goal of this thesis was to study the behavior of the plasma transport during the divertor detachment in order to explain the central electron density rise. The measurement of particle transport coefficients requires sophisticated diagnostic tools. A two color interferometer system was developed and installed on Alcator C-Mod to measure the electron density with high spatial (∼ 2 cm) and high temporal (≤ 1.0 ms) resolution. The system consists of 10 CO2 (10.6 μm) and 4 HeNe (.6328 μm) chords that are used to measure the line integrated density to within 0.08 CO2 degrees or 2.3 x 1016m-2 theoretically. Using the two color interferometer, a series of gas puffing experiments were conducted. The density was varied above and below the threshold density for detachment at a constant magnetic field and plasma current. Using a gas modulation technique, the particle diffusion, D, and the convective velocity, V, were determined. Profiles were inverted using a SVD inversion and the transport coefficients were extracted with a time regression analysis and a transport simulation analysis. Results from each analysis were in good agreement. Measured profiles of the coefficients increased with the radius and the values were consistent with measurements from other experiments. The values exceeded neoclassical predictions by a factor of 10. The profiles also exhibited an inverse dependence with plasma density. The scaling of both attached and detached plasmas agreed well with this inverse scaling. This result and the lack of change in the energy and impurity transport indicate that there was no change in the underlying transport processes after detachment

  5. Measurement of particle transport coefficients on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Luke, T.C.T.

    1994-10-01

    The goal of this thesis was to study the behavior of the plasma transport during the divertor detachment in order to explain the central electron density rise. The measurement of particle transport coefficients requires sophisticated diagnostic tools. A two color interferometer system was developed and installed on Alcator C-Mod to measure the electron density with high spatial ({approx} 2 cm) and high temporal ({le} 1.0 ms) resolution. The system consists of 10 CO{sub 2} (10.6 {mu}m) and 4 HeNe (.6328 {mu}m) chords that are used to measure the line integrated density to within 0.08 CO{sub 2} degrees or 2.3 {times} 10{sup 16}m{sup {minus}2} theoretically. Using the two color interferometer, a series of gas puffing experiments were conducted. The density was varied above and below the threshold density for detachment at a constant magnetic field and plasma current. Using a gas modulation technique, the particle diffusion, D, and the convective velocity, V, were determined. Profiles were inverted using a SVD inversion and the transport coefficients were extracted with a time regression analysis and a transport simulation analysis. Results from each analysis were in good agreement. Measured profiles of the coefficients increased with the radius and the values were consistent with measurements from other experiments. The values exceeded neoclassical predictions by a factor of 10. The profiles also exhibited an inverse dependence with plasma density. The scaling of both attached and detached plasmas agreed well with this inverse scaling. This result and the lack of change in the energy and impurity transport indicate that there was no change in the underlying transport processes after detachment.

  6. Blob sizes and velocities in the Alcator C-Mod scrape-off layer

    DEFF Research Database (Denmark)

    Kube, R.; Garcia, O.E.; LaBombard, B.;

    A new blob-tracking algorithm for the GPI diagnostic installed in the outboard-midplane of Alcator C-Mod is developed. I t tracks large-amplitude fluctuations propagating through the scrape-off layer and calculates blob sizes and velocities. We compare the results of this method to a blob velocity...

  7. Mode conversion electron heating in Alcator C-Mod: Theory and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Bonoli, P. T. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Brambilla, M. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Nelson-Melby, E. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Phillips, C. K. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Porkolab, M. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Schilling, G. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Taylor, G. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Wukitch, S. J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Boivin, R. L. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Boswell, C. J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)] (and others)

    2000-05-01

    Localized electron heating [full width at half maximum of {delta}(r/a){approx_equal}0.2] by mode converted ion Bernstein waves (IBW) has been observed in the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. These experiments were performed in D({sup 3}He) plasmas at high magnetic field (B{sub 0}=7.9 T), high-plasma density (n{sub e0}{>=}1.5x10{sup 20} m{sup -3}), and for 0.05{<=}n{sub He-3}/n{sub e}{<=}0.30. Electron heating profiles of the mode converted IBW were measured using a break in slope analysis of the electron temperature versus time in the presence of rf (radio frequency) modulation. The peak position of electron heating was found to be well-correlated with {sup 3}He concentration, in agreement with the predictions of cold plasma theory. Recently, a toroidal full-wave ion cyclotron range of frequencies (ICRF) code TORIC [M. Brambilla, Nucl. Fusion 38, 1805 (1998)] was modified to include the effects of IBW electron Landau damping at (k{sub (perpendicular} {sub sign)}{rho}{sub i}){sup 2}>>1, This model was used in combination with a 1D (one-dimensional) integral wave equation code METS [D. N. Smithe et al., Radio Frequency Power in Plasmas, AIP Conf. Proc. 403 (1997), p. 367] to analyze these experiments. Model predictions were found to be in qualitative and in some instances quantitative agreement with experimental measurements. A model for mode conversion current drive (MCCD) has also been developed which combines a toroidal full wave code with an adjoint evaluation of the ICRF current drive efficiency. Predictions for off-axis MCCD in C-Mod have been made using this model and will be described. (c) 2000 American Institute of Physics.

  8. Characterization of enhanced D{alpha} high-confinement modes in Alcator {ital C}-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Greenwald, M.; Boivin, R.; Bonoli, P.; Budny, R.; Fiore, C.; Goetz, J.; Granetz, R.; Hubbard, A.; Hutchinson, I.; Irby, J.; LaBombard, B.; Lin, Y.; Lipschultz, B.; Marmar, E.; Mazurenko, A.; Mossessian, D.; Sunn Pedersen, T.; Pitcher, C.S.; Porkolab, M.; Rice, J.; Rowan, W.; Snipes, J.; Schilling, G.; Takase, Y.; Terry, J.; Wolfe, S.; Weaver, J.; Welch, B.; Wukitch, S. [MIT-Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States)

    1999-05-01

    Regimes of high-confinement mode have been studied in the Alcator {ital C}-Mod tokamak [Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)]. Plasmas with no edge localized modes (ELM-free) have been compared in detail to a new regime, enhanced D{alpha} (EDA). EDA discharges have only slightly lower energy confinement than comparable ELM-free ones, but show markedly reduced impurity confinement. Thus EDA discharges do not accumulate impurities and typically have a lower fraction of radiated power. The edge gradients in EDA seem to be relaxed by a continuous process rather than an intermittent one as is the case for standard ELMy discharges and thus do not present the first wall with large periodic heat loads. This process is probably related to fluctuations seen in the plasma edge. EDA plasmas are more likely at low plasma current (q{gt}3.7), for moderate plasma shaping, (triangularity {approximately}0.35{endash}0.55), and for high neutral pressures. As observed in soft x-ray emission, the pedestal width is found to scale with the same parameters that determine the EDA/ELM-free boundary. {copyright} {ital 1999 American Institute of Physics.}

  9. Measurements of the high confinement mode pedestal region on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, A.E.; Boivin, R.L.; Granetz, R.S.; Greenwald, M.; Hutchinson, I.H.; Irby, J.H.; In, Y.; Kesner, J.; LaBombard, B.; Lin, Y.; Rice, J.E.; Sunn Pedersen, T.; Snipes, J.A.; Stek, P.C.; Takase, Y.; Wolfe, S.M.; Wukitch, S. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1998-05-01

    Measurements of the steep transport barrier at the edge of the Alcator C-Mod tokamak [I. H. Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)] are presented. The parameters at the top of this barrier are in the range T{sub e}=150{endash}750 eV and n{sub e}=0.5{minus}3.3{times}10{sup 20} m{sup {minus}3}, depending on the confinement regime. Type III edge localized modes (ELMs) have an upper temperature limit. T{sub e} pedestal profiles show a barrier width {Delta}{sub T}{approx_equal}8 mm. Soft x-ray emissivity profiles are narrower, with {Delta}=2{endash}4 mm. Edge currents are calculated to alter the ideal stability boundary favorably, leading to ideally stable pedestal profiles. High frequency, broadband, edge density fluctuations are sometimes observed in H-mode (high-confinement mode) and are associated with enhanced particle transport. Coherent magnetic fluctuations localized near the pedestal are also seen. {copyright} {ital 1998 American Institute of Physics.}

  10. Analysis of 4-strap ICRF Antenna Performance in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    G. Schilling; S.J. Wukitch; R.L. Boivin; J.A. Goetz; J.C. Hosea; J.H. Irby; Y. Lin; A. Parisot; M. Porkolab; J.R. Wilson; the Alcator C-Mod Team

    2003-07-31

    A 4-strap ICRF antenna was designed and fabricated for plasma heating and current drive in the Alcator C-Mod tokamak. Initial upgrades were carried out in 2000 and 2001, which eliminated surface arcing between the metallic protection tiles and reduced plasma-wall interactions at the antenna front surface. A boron nitride septum was added at the antenna midplane to intersect electric fields resulting from radio-frequency sheath rectification, which eliminated antenna corner heating at high power levels. The current feeds to the radiating straps were reoriented from an E||B to E parallel B geometry, avoiding the empirically observed {approx}15 kV/cm field limit and raising antenna voltage holding capability. Further modifications were carried out in 2002 and 2003. These included changes to the antenna current strap, the boron nitride tile mounting geometry, and shielding the BN-metal interface from the plasma. The antenna heating efficiency, power, and voltage characteristics under these various configurations will be presented.

  11. Upgrades to the 4-strap ICRF Antenna in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    G. Schilling; J.C. Hosea; J.R. Wilson; W. Beck; R.L. Boivin; P.T. Bonoli; D. Gwinn; W.E. Lee; E. Nelson-Melby; M. Porkolab; R. Vieira; S.J. Wukitch; and J.A. Goetz

    2001-06-12

    A 4-strap ICRF antenna suitable for plasma heating and current drive has been designed and fabricated for the Alcator C-Mod tokamak. Initial operation in plasma was limited by high metallic impurity injection resulting from front surface arcing between protection tiles and from current straps to Faraday shields. Antenna modifications were made in February 2000, resulting in impurity reduction, but low-heating efficiency was observed when the antenna was operated in its 4-strap rather than a 2-strap configuration. Further modifications were made in July 2000, with the installation of BN plasma-facing tiles and radio- frequency bypassing of the antenna backplane edges and ends to reduce potential leakage coupling to plasma surface modes. Good heating efficiency was now observed in both heating configurations, but coupled power was limited to 2.5 MW in H-mode, 3 MW in L-mode, by plasma-wall interactions. Additional modifications were started in February 2001 and will be completed by this meeting. All the above upgrades and their effect on antenna performance will be presented.

  12. Effects of neutral particles on edge dynamics in Alcator C-Mod plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Boivin, R. L. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Goetz, J. A. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hubbard, A. E. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hughes, J. W. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hutchinson, I. H. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Irby, J. H. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); LaBombard, B. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Marmar, E. S. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Mossessian, D. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Pitcher, C. S. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)] (and others)

    2000-05-01

    Neutral particle densities and energy losses have been measured in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Their effect on the formation and evolution of the edge barrier which accompanies the enhanced confinement regime are discussed. The neutrals can enter the edge dynamics through the particle, momentum, and energy balance. Neutral densities of up to 5x10{sup 16} m{sup -3} have been measured in the edge barrier region. Neutrals enter the local dynamics around most of the periphery, not just at the X-point. High resolution measurements of the ionization profile have been obtained for the region near the separatrix. The profile shifts inside the separatrix as the plasma is making a transition from low-to high-mode confinement (H-mode) regimes, partly accounting for the dramatic rise in edge density. The measured neutral density is large enough to affect the bulk ion momentum by charge exchange, and thereby introduces a negative radial electric field at the edge. At the same time, significant edge heat flux, carried by the neutrals, contributes to the measured power loss. At very high edge densities, this loss mechanism could contribute to quenching H-modes. (c) 2000 American Institute of Physics.

  13. Upgrades to the 4-strap ICRF antenna in Alcator C-Mod

    Science.gov (United States)

    Schilling, G.; Hosea, J. C.; Wilson, J. R.; Beck, W.; Boivin, R. L.; Bonoli, P. T.; Gwinn, D.; Lee, W. D.; Nelson-Melby, E.; Porkolab, M.; Vieira, R.; Wukitch, S. J.; Goetz, J. A.

    2001-10-01

    A 4-strap ICRF antenna suitable for plasma heating and current drive has been designed and fabricated for the Alcator C-Mod tokamak. Initial operation in plasma was limited by high metallic impurity injection resulting from front surface arcing between protection tiles and from current straps to Faraday shields. Antenna modifications were made in 2/2000, resulting in impurity reduction, but low heating efficiency was observed when the antenna was operated in its 4-strap rather than a 2-strap configuration. Further modifications were made in 7/2000, with the installation of BN plasma-facing tiles and radiofrequency bypassing of the antenna backplane edges and ends to reduce potential leakage coupling to plasma surface modes. Good heating efficiency was now observed in both heating configurations, but coupled power was limited to 2.5 MW in H-mode, 3 MW in L-mode, by plasma-wall interactions. Additional modifications were started in 2/2001 and will be completed by this meeting. All the above upgrades and their effect on antenna performance will be presented.

  14. The multi-spectral line-polarization MSE system on Alcator C-Mod

    Science.gov (United States)

    Mumgaard, R. T.; Scott, S. D.; Khoury, M.

    2016-11-01

    A multi-spectral line-polarization motional Stark effect (MSE-MSLP) diagnostic has been developed for the Alcator C-Mod tokamak wherein the Stokes vector is measured in multiple wavelength bands simultaneously on the same sightline to enable better polarized background subtraction. A ten-sightline, four wavelength MSE-MSLP detector system was designed, constructed, and qualified. This system consists of a high-throughput polychromator for each sightline designed to provide large étendue and precise spectral filtering in a cost-effective manner. Each polychromator utilizes four narrow bandpass interference filters and four custom large diameter avalanche photodiode detectors. Two filters collect light to the red and blue of the MSE emission spectrum while the remaining two filters collect the beam pi and sigma emission generated at the same viewing volume. The filter wavelengths are temperature tuned using custom ovens in an automated manner. All system functions are remote controllable and the system can be easily retrofitted to existing single-wavelength line-polarization MSE systems.

  15. BOUT++ Simulations of Edge Turbulence in Alcator C-Mod's EDA H-Mode

    Science.gov (United States)

    Davis, E. M.; Porkolab, M.; Hughes, J. W.; Labombard, B.; Snyder, P. B.; Xu, X. Q.

    2013-10-01

    Energy confinement in tokamaks is believed to be strongly controlled by plasma transport in the pedestal. The pedestal of Alcator C-Mod's Enhanced Dα (EDA) H-mode (ν* > 1) is regulated by a quasi-coherent mode (QCM), an edge fluctuation believed to reduce particle confinement and allow steady-state H-mode operation. ELITE calculations indicate that EDA H-modes sit well below the ideal peeling-ballooning instability threshold, in contrast with ELMy H-modes. Here, we use a 3-field reduced MHD model in BOUT++ to study the effects of nonideal and nonlinear physics on EDA H-modes. In particular, incorporation of realistic pedestal resistivity is found to drive resistive ballooning modes (RBMs) and increase linear growth rates above the corresponding ideal rates. These RBMs may ultimately be responsible for constraining the EDA pedestal gradient. However, recent high-fidelity mirror Langmuir probe measurements indicate that the QCM is an electron drift-Alfvén wave - not a RBM. Inclusion of the parallel pressure gradient term in the 3-field reduced MHD Ohm's law and various higher field fluid models are implemented in an effort to capture this drift wave-like response. This work was performed under the auspices of the USDoE under awards DE-FG02-94-ER54235, DE-AC52-07NA27344, DE-AC52-07NA27344, and NNSA SSGF.

  16. The development of an Omegratron plasma ion mass spectrometer for Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, E.E. Jr.

    1993-05-01

    A new diagnostic device, the Omegatron Probe, has been developed to investigate relative impurity levels and impurity charge state distribution in the Alcator C-Mod Tokamak edge plasma. The Omegatron probe consists of two principal components, a ``front-end`` of independently biased grids, arranged in a gridded energy analyzer fashion and a large collection cavity. Particles enter the probe in a thin ``ribbon`` through a knife-edge slit. The grids provide a means to measure and control the parallel energy distribution of the ions. In the collection cavity, an oscillating electric field is applied perpendicularly to the ambient magnetic field. Ions whose cyclotron frequencies are resonant with this electric field oscillation will gain perpendicular energy and be collected. In this way, the probe can be operated in two modes: first, by fixing the potentials on the grids and sweeping frequencies to obtain a `` Z/m spectrum`` of ion species and second, by fixing the frequency and sweeping the grid potentials to obtain the distribution function of an individual impurity species. The Omegatron probe performed successfully in tests on a Hollow Cathode Discharge (HCD) linear plasma column. It obtained measurements of T{sub e} {approx} 5 eV, T{sub i} (H{sup +}) {approx} 2.0 {plus_minus} 0.2 eV, n{sub 0} {approx} 9 {times} 10{sup 15} m{sup {minus}3}, RMS potential fluctuation levels of {approximately} 0.5 {plus_minus} 0.05 {plus_minus} T{sub e}, and obtained ``Z/m`` spectra for the plasma ions (H{sup +}, H{sub 2}{sup +}, He{sup +}). Additional experiments confirmed the theoretical scalings of the f/{delta}f resolution with the applied electric field and magnetic field strengths. The instrument yielded an absolute level of resolution, f/{delta}f, of approximately 2.5 to 3 times the theoretical values. Finally, the results from the HCD are used to project operation on Alcator C-Mod.

  17. Alcator C-Mod ion cyclotron antenna performance

    International Nuclear Information System (INIS)

    Ion cyclotron range of frequency (ICRF) heating is expected to be an important auxiliary heating source for ITER and fusion reactors. One of the keys to successful ICRF heating is the antenna performance and a number of issues can limit the antenna performance including poor voltage and power handling, impurity production, strong RF plasma edge interactions, poor RF coupling, and localized heating of the antenna structure. High power density antenna operation, with all metal protection tiles and plasma facing components (PFC), present significant challenges to ICRF antenna operation. High performance plasmas can be achieved with all metal PFC's after boronization. In Ohmic H-mode discharges, the plasma performance degradation occurs at a rate 3-4 times slower than RF heated discharges with the similar input energy (discharge integrated). The erosion process also appears to be accelerated for weaker single pass absorption heating scenario, D(3He) on C-Mod. The C-Mod H minority single pass absorption is stronger and is similar to that expected in ITER implying similar RF induced erosion in ITER as on C-Mod. Since Faraday screen-less antenna operation has a number of advantages; the J antenna was operated without a Faraday screen. The voltage and power handling were unaffected by the screen removal. However, the heating effectiveness was 15-20% less and the influx of Cu was identified as the likely cause of the decreased performance. On C-Mod, high density discharges can yield neutral pressures at which antenna operation is prohibited. This neutral pressure limit may be related to phenomena associated with antenna ELM (edge localized mode) interactions. Experiments showed that multipactor can cause a glow discharge at neutral pressures two orders of magnitude below the Paschen breakdown limit. In the presence of a 0.1 T B-field, measurements on the C-Mod antennas showed the presence of a glow discharge at a neutral pressure similar to the observed operational neutral

  18. Current Profile Measurements from Moderate to Strong Lower Hybrid Single-Pass Damping on Alcator C-Mod

    Science.gov (United States)

    Mumgaard, R. T.; Wallace, G. M.; Scott, S. D.; Shiraiwa, S.; Faust, I.; Parker, R. R.

    2015-11-01

    Lower Hybrid Current Drive (LHCD) is an effective tool to non-inductively drive up to 100% of the plasma current on Alcator C-Mod. Measurements with an upgraded MSE diagnostic show that the fast-electron current profile is broader than the Ohmic current profile but still located the plasma core in agreement with strongly centrally peaked fast electron bremsstrahlung (FEB) measurements. Scans in a regime of high current drive efficiency across a range of density, LHCD power, launched n||, and plasma current show the driven current profile, FEB profile shapes, and current drive efficiency are sensitive only to total plasma current. Simulations using ray-tracing Fokker Planck codes show that the rays make 1-3 bounces through the plasma edge to bridge the spectral gap. Although in agreement with the total current, the simulations qualitatively disagree with experiment regarding current and FEB profiles as well as sensitivity to power and density. Simulations at higher plasma temperature and current predict stronger single-pass damping and preliminary experiments show increased current drive efficiency. Experiments to determine if the profile discrepancies persist when the ray bounces play a reduced role are planned, including companion experiments in D and He resulting in different edge plasma conditions. This work was performed on the Alcator C-Mod tokamak, a DoE Office of Science user facility, and is supported by USDoE awards DE-FC02-99ER54512 and DE-AC02-09CH11466.

  19. Status of the neutron diagnostic experiment for Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Fiore, C.L.; Boivin, R.; Granetz, R.S.; Fuller, T.; Kurz, C. (MIT Plasma Fusion Center, Cambridge, Massachusetts 02139 (United States))

    1992-10-01

    The Alcator C-Mod experiment is expected to produce neutrons at a rate of up to 1 {times} 10{sup 16} n/s during its highest performance operation. The global neutron detection system must provide accurate measurement of this production over the full operating regime of the experiment, 10{sup 11}--10{sup 16} n/s. This is accomplished using a series of uranium fission detectors enriched in U{sup 235} of staggered sensitivity. These are distributed in two moderator assemblies. A BF{sub 3} detector is added to each moderator to cover the lowest operational range. Preliminary calibration of the system has been accomplished using a 65.4 {mu}g Cf{sup 252} source placed inside the Alcator C-Mod vacuum vessel. The system design and calibration is discussed herein.

  20. Measurements of relativistic emission from runaway electrons in Alcator C-Mod: spectrum, polarization, and spatial structure

    Science.gov (United States)

    Granetz, Robert; Mumgaard, Robert

    2014-10-01

    At low densities, runaway electrons (RE's) can be generated during the flattop of Alcator C-Mod discharges with highly relativistic energies, γ >> 1 , allowing careful study under steady conditions. These RE's emit light in a narrow forward-peaked cone which is detected with a number of diagnostics, including spectrometers, a video imaging camera, and polarimetry (using the MSE system), in addition to the standard hard x-ray detectors. These measurements of the relativistic emission can provide information about the RE energy distribution, pitch angle distribution, and spatial distribution. Unlike most other tokamaks, C-Mod's high magnetic field shifts the peak of the continuum emission into the visible, due to the smaller gyroradius and higher gyro-frequency, allowing for excellent spectral coverage with standard spectrometers, and thus detailed comparison to theoretical predictions of synchrotron and bremsstrahlung spectra. Additionally, camera images occasionally show highly structured formations. Profiles of the polarization fraction and polarization angle show radial structure, including a jump of 90° outboard of the magnetic axis, in qualitative agreement with recent theoretical calculations for relativistic electrons in a tokamak field. This work is supported by the U.S. Department of Energy.

  1. Simulations of beam-emission spectroscopy on Alcator C-Mod (abstract)

    Energy Technology Data Exchange (ETDEWEB)

    Sampsell, M. B.; Bravenec, R. V.; Rowan, W. L.; Eisner, E. C.; Patterson, D. M.; Bretz, N. L.; Boivin, R. L.; Hubbard, A. E.; Irby, J. H.; Marmar, E. S. (and others)

    2001-01-01

    An eight-channel beam-emission-spectroscopy (BES)1 system has been installed on the Alcator C-Mod tokamak, intended for use with a diagnostic neutral hydrogen beam (DNB). Capable of localized measurements from the plasma edge to the plasma core, the BES diagnostic collects light from the first Balmer transition (H{sub {alpha}}) resultant from beam/plasma collisions. The H{sub {alpha}} line splits into several components whose central wavelengths depend on the viewing geometry, the magnetic field, and the beam energy. This is due to the Doppler shifts from viewing the beam off perpendicular, the different velocities of the three mass components of the beam (H, H{sub 2}, H{sub 3}), and the large motional Stark effect. Optimal signal-to-noise requires collecting these components while attenuating all other emission: primarily bremsstrahlung and D{sub {alpha}} radiation (from plasma D{sup 0}/e{sup -} collisions). Tunable bandpass filters are thus required. A BES simulation code has been developed that calculates the brightnesses (bremsstrahlung, D{sub {alpha}}, H{sub {alpha}}) versus wavelength using plasma profile data from the C-Mod MDSplus database,2 a computation of the beam penetration, the viewing and DNB geometries, and bandpass filter characteristics. The model was first used to estimate signal levels and choose the optimal BES bandpass filters; its ultimate purpose is to determine the shot-to-shot tuning requirements of the filters for different discharge conditions. Comparisons of measured and predicted background bremsstrahlung and D{sub {alpha}} brightnesses are presented, as are first measurements and calculations of the beam emission. The code is written in the IDL programming language3 utilizing the ''widget'' graphical user interface. Designed for geometrical and spectral flexibility, it can be modified to simulate other beam diagnostics such as motional-Stark-effect plasma current measurements and charge-exchange recombination

  2. Survey of ICRF heating experiments and enhanced performance modes in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Y.; Boivin, R.L.; Bonoli, P.T. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Plasma Fusion Center] [and others

    1996-12-01

    Results of ICRF heating experiments in Alcator C-Mod during the November 1994 to June 1995 campaign are summarized. Efficient heating of high-density (nbar{sub e} ``tokamaks. PEP modes with highly peaked density and ion temperature profiles and highly enhanced fusion reactivity were obtained with Li pellet injection followed by on-axis ICRF heating at both B{sub T} = 5.3 T (H minority heating) and 8 T ({sup 3}He minority heating). In H-{sup 3}He plasmas at B{sub T} = 6.5 T highly localized direct electron heating by the mode converted ion Bernstein wave was observed. Nearly complete absorption by electrons in a small volume resulted in an extremely high electron heating power density of P{sub eO} ``

  3. The design and performance of a twenty barrel hydrogen pellet injector for Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Urbahn, J.A.

    1994-05-01

    A twenty barrel hydrogen pellet injector has been designed, built and tested both in the laboratory and on the Alcator C-Mod Tokamak at MIT. The injector functions by firing pellets of frozen hydrogen or deuterium deep into the plasma discharge for the purpose of fueling the plasma, modifying the density profile and increasing the global energy confinement time. The design goals of the injector are: (1) Operational flexibility, (2) High reliability, (3) Remote operation with minimal maintenance. These requirements have lead to a single stage, pipe gun design with twenty barrels. Pellets are formed by in- situ condensation of the fuel gas, thus avoiding moving parts at cryogenic temperatures. The injector is the first to dispense with the need for cryogenic fluids and instead uses a closed cycle refrigerator to cool the thermal system components. The twenty barrels of the injector produce pellets of four different size groups and allow for a high degree of flexibility in fueling experiments. Operation of the injector is under PLC control allowing for remote operation, interlocked safety features and automated pellet manufacturing. The injector has been extrusively tested and shown to produce pellets reliably with velocities up to 1400 m/sec. During the period from September to November of 1993, the injector was successfully used to fire pellets into over fifty plasma discharges. Experimental results include data on the pellet penetration into the plasma using an advanced pellet tracking diagnostic with improved time and spatial response. Data from the tracker indicates pellet penetrations were between 30 and 86 percent of the plasma minor radius.

  4. Pedestal structure and stability in H-mode and I-mode: a comparative study on Alcator C-Mod

    International Nuclear Information System (INIS)

    New experimental data from the Alcator C-Mod tokamak are used to benchmark predictive modelling of the edge pedestal in various high-confinement regimes, contributing to greater confidence in projection of pedestal height and width in ITER and reactors. ELMy H-modes operate near stability limits for ideal peeling–ballooning modes, as shown by calculations with the ELITE code. Experimental pedestal width in ELMy H-mode scales as the square root of βpol at the pedestal top, i.e. the dependence expected from theory if kinetic ballooning modes (KBMs) were responsible for limiting the pedestal width. A search for KBMs in experiment has revealed a short-wavelength electromagnetic fluctuation in the pedestal that is a candidate driver for inter-edge localized mode (ELM) pedestal regulation. A predictive pedestal model (EPED) has been tested on an extended set of ELMy H-modes from C-Mod, reproducing pedestal height and width reasonably well across the data set, and extending the tested range of EPED to the highest absolute pressures available on any existing tokamak and to within a factor of three of the pedestal pressure targeted for ITER. In addition, C-Mod offers access to two regimes, enhanced D-alpha (EDA) H-mode and I-mode, that have high pedestals, but in which large ELM activity is naturally suppressed and, instead, particle and impurity transport are regulated continuously. Pedestals of EDA H-mode and I-mode discharges are found to be ideal magnetohydrodynamic (MHD) stable with ELITE, consistent with the general absence of ELM activity. Invocation of alternative physics mechanisms may be required to make EPED-like predictions of pedestals in these kinds of intrinsically ELM-suppressed regimes, which would be very beneficial to operation in burning plasma devices. (paper)

  5. ECE Temperature Fluctuations associated with EDA H-Mode discharges in Alcator C-Mod

    Science.gov (United States)

    Phillips, P. E.; Lynn, A. G.

    2006-10-01

    Alcator C-Mod exhibits an ELM-free H-mode with ``enhanced,,lpha'' emission accompanied by a quasi-coherent mode (QCM) edge relaxation mechanism. This steady state H-mode lowers the peak heat load to the diverters which is advantageous for reactor operations. A high-resolution heterodyne electron-cyclotron-emission (ECE) radiometer with 32 channels (δR˜7mm) and a bandwidth up to 1MHz covering the full radius of C-Mod has observed spatial resolved temperature fluctuations that are highly correlated with the edge QCM mode. The QCM mode is also directly observed by the edge ECE channels though the changes in optical depth due to the large density fluctuations in the QCM (˜30%). Details of these measurements will be presented in this poster.

  6. Novel energy resolving x-ray pinhole camera on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Pablant, N. A.; Delgado-Aparicio, L.; Bitter, M.; Ellis, R.; Hill, K. W. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Brandstetter, S.; Eikenberry, E.; Hofer, P.; Schneebeli, M. [Dectris Ltd., Baden (Switzerland)

    2012-10-15

    A new energy resolving x-ray pinhole camera has been recently installed on Alcator C-Mod. This diagnostic is capable of 1D or 2D imaging with a spatial resolution of Almost-Equal-To 1 cm, an energy resolution of Almost-Equal-To 1 keV in the range of 3.5-15 keV and a maximum time resolution of 5 ms. A novel use of a Pilatus 2 hybrid-pixel x-ray detector [P. Kraft et al., J. Synchrotron Rad. 16, 368 (2009)] is employed in which the lower energy threshold of individual pixels is adjusted, allowing regions of a single detector to be sensitive to different x-ray energy ranges. Development of this new detector calibration technique was done as a collaboration between PPPL and Dectris Ltd. The calibration procedure is described, and the energy resolution of the detector is characterized. Initial data from this installation on Alcator C-Mod is presented. This diagnostic provides line-integrated measurements of impurity emission which can be used to determine impurity concentrations as well as the electron energy distribution.

  7. Design of a New Optical System for Alcator C-Mod Motional Stark Effect Diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Jinseok; Scott, Steve; Manfred, Bitter; Lerner, Lerner

    2009-11-12

    The motional Stark effect (MSE) diagnostic on Alcator C-Mod uses an in-vessel optical system (five lenses and three mirrors) to relay polarized light to an external polarimeter because port access limitations on Alcator C-Mod preclude a direct view of the diagnostic beam. The system experiences unacceptable, spurious drifts of order several degrees in measured pitch angle over the course of a run day. Recent experiments illuminated the MSE diagnostic with polarized light of fixed orientation as heat was applied to various optical elements. A large change in measured angle was observed as two particular lenses were heated, indicating that thermal-stress-induced birefringence is a likely cause of the spurious variability. Several new optical designs have been evaluated to eliminate the affected in-vessel lenses and to replace the focusing they provide with curved mirrors; however, ray tracing calculations imply that this method is not feasible. A new approach is under consideration that utilizes in situ calibrations with in-vessel reference polarized light sources. 2008 American Institute of Physics.

  8. Second Harmonics of Reversed Shear TAE in Alcator C-Mod Geometry

    Science.gov (United States)

    Chen, Eugene; Berk, Herbert; Breizman, Boris; Zheng, Linjin

    2009-11-01

    Experiments on Alcator C-Mod, operating with reversed magnetic shear, reveal Toroidal Alfven Eigenmodes (TAE) together with signals at twice the mode frequency. The double frequency signals can be viewed as second harmonic sidebands driven by quadratic non-linear terms in the MHD equations, in analogy with a corresponding theory for Alfven Cascades [1]. However, these nonlinear sidebands have not yet been quantified by any of the existing codes. In this work, we extend AEGIS code [2] to capture nonlinear effects iteratively by treating the nonlinear terms as a driving source in the linear MHD solver. We first compute the TAE mode structure for realistic geometry and q-profile and then use it to find the spatial structure of the second harmonic density perturbation, which can be directly compared with PCI measurements at Alcator C-Mod. [1] H. Smith, B. N. Breizman, M. Lisak and D. Anderson, Physics of Plasmas 13 042504 (2006) [2] L. J. Zheng and M. Kotschenreuther, Journal of Computational Physics 211 (2006) 748-766

  9. Study and optimization of boronization in Alcator C-Mod using the Surface Science Station (S{sup 3})

    Energy Technology Data Exchange (ETDEWEB)

    Ochoukov, Roman, E-mail: ochoukov@psfc.mit.edu [Plasma Science and Fusion Center MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Whyte, Dennis; Lipschultz, Bruce; LaBombard, Brian [Plasma Science and Fusion Center MIT, NW17, 175 Albany Street, Cambridge, MA 02139 (United States); Gierse, Niels [Institute of Energy and Climate Research - Plasma Physics, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Partner in the Trilateral Euregio Cluster, Juelich (Germany); Physikalisches Institut, Universitaet zu Koeln, D-50937 Cologne (Germany); Harrison, Soren [Fusion Research Technologies, 519 Somerville Avenue 243, Somerville, MA 02143 (United States)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Boron deposition profiles measured on Alcator C-Mod during boronization. Black-Right-Pointing-Pointer Boron deposition profile is consistent with ionic deposition. Black-Right-Pointing-Pointer Monte Carlo simulation of boron deposition agrees with experiment assuming warm ({approx}1-10 eV) boron{sup +1} ions. - Abstract: A Surface Science Station (S{sup 3}) on the Alcator C-Mod tokamak is used to study and optimize the location and rate of boron film deposition in situ during electron cyclotron (EC) discharge plasmas using 2.45 GHz radio-frequency (RF) heating and a mixture of helium and diborane (B{sub 2}D{sub 6}) gasses. The radial profile of boron deposition is measured with a pair of quartz microbalances (QMB) on S{sup 3}, the faces of which can be rotated 360 Degree-Sign including orientations parallel and perpendicular to the toroidal magnetic field B{sub T} {approx}0.1 T. The plasma electron density is measured with a Langmuir probe, also on S{sup 3} in the vicinity of the QMBs, and typical values are {approx}1 Multiplication-Sign 10{sup 16} m{sup -3}. A maximum boron deposition rate of 0.82 {mu}g/cm{sup 2}/min is obtained, which corresponds to 3.5 nm/min if the film density is that of solid boron. These deposition rates are sufficient for boron film applications between tokamak discharges. However the deposition does not peak at the EC resonance as previously assumed. Rather, deposition peaks near the upper hybrid (UH) resonance, {approx}5 cm outboard of the EC resonance. This has implications for RF absorption, with the RF waves being no longer damped on the electrons at the EC resonance. The previously inferred radial locations of critical erosion zones in Alcator C-Mod also need to be re-evaluated. The boron deposition profile versus major radius follows the ion flux/density profile, implying that the boron deposition is primarily ionic. The application of a vertical magnetic field (B{sub V} {approx}0.01 T) was

  10. Multispecies density peaking in gyrokinetic turbulence simulations of low collisionality Alcator C-Mod plasmas

    Science.gov (United States)

    Mikkelsen, D. R.; Bitter, M.; Delgado-Aparicio, L.; Hill, K. W.; Greenwald, M.; Howard, N. T.; Hughes, J. W.; Rice, J. E.; Reinke, M. L.; Podpaly, Y.; Ma, Y.; Candy, J.; Waltz, R. E.

    2015-06-01

    Peaked density profiles in low-collisionality AUG and JET H-mode plasmas are probably caused by a turbulently driven particle pinch, and Alcator C-Mod experiments confirmed that collisionality is a critical parameter. Density peaking in reactors could produce a number of important effects, some beneficial, such as enhanced fusion power and transport of fuel ions from the edge to the core, while others are undesirable, such as lower beta limits, reduced radiation from the plasma edge, and consequently higher divertor heat loads. Fundamental understanding of the pinch will enable planning to optimize these impacts. We show that density peaking is predicted by nonlinear gyrokinetic turbulence simulations based on measured profile data from low collisionality H-mode plasma in Alcator C-Mod. Multiple ion species are included to determine whether hydrogenic density peaking has an isotope dependence or is influenced by typical levels of low-Z impurities, and whether impurity density peaking depends on the species. We find that the deuterium density profile is slightly more peaked than that of hydrogen, and that experimentally relevant levels of boron have no appreciable effect on hydrogenic density peaking. The ratio of density at r/a = 0.44 to that at r/a = 0.74 is 1.2 for the majority D and minority H ions (and for electrons), and increases with impurity Z: 1.1 for helium, 1.15 for boron, 1.3 for neon, 1.4 for argon, and 1.5 for molybdenum. The ion temperature profile is varied to match better the predicted heat flux with the experimental transport analysis, but the resulting factor of two change in heat transport has only a weak effect on the predicted density peaking.

  11. Spectroscopic measurement of impurity transport coefficients and penetration efficiencies in Alcator C-Mod plasmas

    Science.gov (United States)

    Graf, M. A.; Rice, J. E.; Terry, J. L.; Marmar, E. S.; Goetz, J. A.; McCracken, G. M.; Bombarda, F.; May, M. J.

    1995-01-01

    Impurity transport coefficients and the penetration efficiencies of intrinsic and injected impurities through the separatrix of diverted Alcator C-Mod discharges have been measured using x-ray and vacuum ultraviolet (VUV) spectroscopic diagnostics. The dominant low Z intrinsic impurity in C-Mod is carbon which is found to be present in concentrations of less than 0.5%. Molybdenum, from the plasma facing components, is the dominant high Z impurity and is typically found in concentrations of about 0.02%. Trace amounts of medium and high Z nonrecycling impurities can be injected at the midplane using the laser blow-off technique and calibrated amounts of recycling, gaseous impurities can be introduced through fast valves either at the midplane or at various locations in the divertor chamber. A five chord crystal x-ray spectrometer array with high spectral resolution is used to provide spatial profiles of high charge state impurities. An absolutely calibrated, grazing incidence VUV spectrograph with high time resolution and a broad spectral range allows for the simultaneous measurement of many impurity lines. Various filtered soft x-ray diode arrays allow for spatial reconstructions of plasma emissivity. The observed brightnesses and emissivities from a number of impurity lines are used together with the mist transport code and a collisional-radiative atomic physics model to determine charge state density profiles and impurity transport coefficients. Comparisons of the deduced impurity content with the measured Zeff and total radiated power of the plasma are made.

  12. Upgraded PMI diagnostic capabilities using Accelerator-based In-situ Materials Surveillance (AIMS) on Alcator C-Mod

    Science.gov (United States)

    Kesler, Leigh; Barnard, Harold; Hartwig, Zachary; Sorbom, Brandon; Lanza, Richard; Terry, David; Vieira, Rui; Whyte, Dennis

    2014-10-01

    The AIMS diagnostic was developed to rapidly and non-invasively characterize in-situ plasma material interactions (PMI) in a tokamak. Recent improvements are described which significantly expand this measurement capability on Alcator C-Mod. The detection time at each wall location is reduced from about 10 min to 30 s, via improved hardware and detection geometry. Detectors are in an augmented re-entrant tube to maximize the solid angle between detectors and diagnostic locations. Spatial range is expanded by using beam dynamics simulation to design upgraded B-field power supplies to provide maximal poloidal access, including a ~20° toroidal range in the divertor. Measurement accuracy is improved with angular and energy resolved cross section measurements obtained using a separate 0.9 MeV deuteron ion accelerator. Future improvements include the installation of recessed scintillator tiles as beam targets for calibration of the diagnostic. Additionally, implanted depth marker tiles will enable AIMS to observe the in-situ erosion and deposition of high-Z plasma-facing materials. This work is supported by U.S. DOE Grant No. DE-FG02-94ER54235 and Cooperative Agreement No. DE-FC02-99ER54512.

  13. In-situ erosion and deposition measurements of plasma-facing surfaces in Alcator C-Mod

    Science.gov (United States)

    Barnard, Harold S.

    2014-10-01

    The Accelerator Based In-situ Materials Surveillance (AIMS) diagnostic was recently developed to demonstrate the novel application of ion beam analysis (IBA) to in-vessel studies of plasma materials interactions in Alcator C-Mod. The AIMS diagnostic injects a 900 keV deuterium ion beam into the tokamak's vacuum vessel between plasma discharges while magnetic fields are used to steer the ion beam to plasma facing component (PFC) surfaces. Spectroscopic analysis of neutrons and gamma rays from the induced nuclear reactions provides a quantitative, spatially resolved map of the PFC surface composition that includes boron (B) and deuterium (D) content. Since AIMS is sensitive to low-Z elements and C-Mod regularly boronizes PFCs, the evolution of B and D on PFCs can be used to directly study erosion, deposition, and fuel retention in response to plasma operations and wall conditioning processes. AIMS analysis of 18 lower single null I-mode discharges show a net boron deposition rate of 6 +/- 2 nm/s on the inner wall while subsequent inner wall limited discharges and a disruption did not show significant changes in B. Measurements of D content showed relative changes of >2.5 following a similar trend. This suggests high D retention rates and net B deposition rates of ~18 cm/year of plasma exposure are possible and depend strongly on the plasma conditions. Ex-situ IBA was also performed on the same PFCs after removal from C-Mod, successfully validating the AIMS technique. These IBA measurements also show that the B content on the inner wall varied toroidally and poloidally from 0 to 3000 nm, demonstrating the importance of the spatial resolution provided by AIMS and the sensitivity of PFCs to B-field alignment. AIMS upgrades are underway for operation in 2014 and we anticipate new measurements correlating the evolution of PFC surfaces to plasma configuration, RF heating, and current drive scenarios. This work is supported by U.S. DOE Grant No. DE-FG02-94ER54235 and

  14. Design and operation of a novel divertor cryopumping system in Alcator C-Mod

    Science.gov (United States)

    Labombard, B.; Beck, B.; Bosco, J.; Childs, R.; Gwinn, D.; Irby, J.; Leccacorvi, R.; Marazita, S.; Mucic, N.; Pierson, S.; Rokhman, Y.; Titus, P.; Vieira, R.; Zaks, J.; Zhukovsky, A.

    2007-11-01

    C-Mod's recently installed upper-divertor cryopump is unique among the world's tokamaks, employing an array of gas-pumping slots that penetrate the upper divertor target. This geometry enables the use of a single toroidal loop of liquid helium, operating in an efficient heat transfer regime with low or no helium flow. A system pumping speed of 9,600 l/sec for D2 gas has been achieved, matching that of a full-scale prototype system. Neutral pressures in the pumping slots during upper-null plasmas (USN) are found to meet or exceed pressures in the lower divertor's private flux region during lower-null (LSN) -- evidence that the pumping-slot geometry is performing as intended. Very high steady-state pumping throughputs (exceeding ˜140 torr-l/s) have been demonstrated in USN. Reliable and efficient operation of the pump has been established, synchronized with the C-Mod shot cycle and consuming 60 to 90 liters of liquid helium during a full day of operation.

  15. Upgrade to the Gas Puff Imaging Diagnostic that Views Alcator C-Mod's Inboard Edge

    Science.gov (United States)

    Sierchio, J. M.; Terry, J. L.

    2012-10-01

    We describe an upgrade of Alcator C-Mod's Gas Puff Imaging system which views the inboard plasma edge and SOL along lines-of-sight that are approximately parallel to the local magnetic field. The views are arranged in a 2D (R,Z) array with ˜2.8 cm radial coverage and ˜2.4 cm poloidal coverage. 23 of 54 available views were coupled via fibers to individual interference filters and PIN photodiode detectors. We are in the process of upgrading the system in order to increase the sensitivity of the system by replacing the PIN photodiodes with a 4x8 array of Avalanche Photo-Diodes (APD). Light from 30 views is coupled to the single-chip APD array through a single interference filter. We expect an improvement in signal-to-noise ratio of more than 10x. The frequency response of the system will increase from ˜400 kHz to 1MHz. The dynamic range of the new system is manipulated by changing the high-voltages on the APDs. Test results of the detectors' channel-to-channel cross-talk, frequency response, and gain curves will be presented, along with schematics of the experimental setup. The upgraded system allows for more study of inboard edge fluctuations, including whether the quasi-coherent fluctuations observed in the outboard edge also exist inboard.

  16. The dynamics and structure of edge-localized-modes in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Terry, J.L. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States)]. E-mail: terry@psfc.mit.edu; Cziegler, I. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); Hubbard, A.E. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); Snipes, J.A. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); Hughes, J.W. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); Greenwald, M.J. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); LaBombard, B. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); Lin, Y. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States); Phillips, P. [Fusion Research Center, University of Texas-Austin, Austin, TX 78712 (United States); Wukitch, S. [Plasma Science and Fusion Center, MIT, 175 Albany St., Cambridge, MA 02139 (United States)

    2007-06-15

    Characteristics of discrete ELMs produced in Alcator C-Mod discharges of low edge collisionality (0.2 < {nu} {sup *} < 1) and large lower triangularity ({delta} {sub lower} {approx} 0.75) are examined. The energy lost per ELM from the H-mode pedestal is {approx}10% of the pedestal energy. These ELMs exhibit relatively long-lived precursor oscillations, often with two modes of intermediate toroidal mode number present. At the ELM 'crash' multiple plasma filament structures are expelled into the scrape-off-layer. A short-lived high frequency ({approx}0.5 MHz) magnetic oscillation is initiated at the 'crash'. The initial ELM filaments are large perturbations to the SOL with radial extents of 0.5-1 cm and typical radial propagation velocities of 1 km/s. Velocities of up to 8 km/s have been seen. The poloidal extent of the initial filaments is >4.5 cm. The initial filaments are followed (at intervals of {approx}100 {mu}s) by multiple, less perturbing secondary filaments.

  17. Investigation of performance limiting phenomena in a variable phase ICRF antenna in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Wukitch, S J [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Boivin, R L [General Atomics, San Diego, CA 92186 (United States); Bonoli, P T [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Goetz, J A [University of Wisconsin, Madison, WI 53706 (United States); Irby, J [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Hutchinson, I [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Lin, Y [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Parisot, A [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Porkolab, M [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Marmar, E [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Schilling, G [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Wilson, J R [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2004-09-01

    High power density, phased antenna operation can often be limited by antenna voltage handling and/or impurity and density production. Using a pair of two-strap antennas for comparison, the performance of a four-strap, fast wave antenna is assessed for a variety of configurations and antenna phases in Alcator C-Mod. To obtain robust voltage handling, the antenna was reconfigured to eliminate regions where the RF E-field is parallel to B or to reduce the RF E-field to <1.0 MV m{sup -1}. To limit impurity generation, BN tiles were used to replace the original Mo tiles, a BN clad septum was inserted to limit field line connection length, and BN-metal interfaces were shielded from the plasma. With these modifications, the antenna heating efficiency and impurity generation are nearly identical to those of the two-strap antennas and independent of antenna phase in L-mode discharges. This antenna has achieved 11 MW m{sup -2} in both heating and current drive phases in both L-mode and H-mode discharges.

  18. Measurement of LHCD edge power deposition through modulation techniques on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Baek, S. G.; Chilenksi, M. A.; Hubbard, A.; Hughes, J. W.; Terry, J. L.; Shiraiwa, S.; Walk, J. R.; Wallace, G. M.; Whyte, D. G. [MIT Plasma Science and Fusion Center, Cambridge, MA USA (United States); Edlund, E. [Princeton Plasma Physics Laboratory, Princeton, NJ USA (United States)

    2015-12-10

    The efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 · 1020 [m{sup −3}]) no measurable current is driven. While a number of causes have been suggested, no specific mechanism has been shown to be responsible for the loss of current drive at high density. Fast modulation of the LH power was used to isolate and quantify the LHCD deposition within the plasma. Measurements from these plasmas provide unique evidence for determining a root cause. Modulation of LH power in steady plasmas exhibited no correlated change in the core temperature. A correlated, prompt response in the edge suggests that the loss in efficiency is related to a edge absorption mechanism. This follows previous results which found the generation of n{sub ||}-independent SOL currents. Multiple Langmuir probe array measurements of the conducted heat conclude that the lost power is deposited near the last closed flux surface. The heat flux induced by LH waves onto the outer divertor is calculated. Changes in the neutral pressure, ionization and hard X-ray emission at high density highlight the importance of the active divertor in the loss of efficiency. Results of this study implicate a mechanism which may occur over multiple passes, leading to power absorption near the LCFS.

  19. Robotic calibration of the motional Stark effect diagnostic on Alcator C-Mod

    Science.gov (United States)

    Mumgaard, Robert T.; Scott, Steven D.; Ko, Jinseok

    2014-05-01

    The capability to calibrate diagnostics, such as the Motional Stark Effect (MSE) diagnostic, without using plasma or beam-into-gas discharges will become increasingly important on next step fusion facilities due to machine availability and operational constraints. A robotic calibration system consisting of a motorized three-axis positioning system and a polarization light source capable of generating arbitrary polarization states with a linear polarization angle accuracy of deployed on Alcator C-Mod. The polarization response of the complex diagnostic is shown to be fully captured using a Fourier expansion of the detector signals in terms of even harmonics of the input polarization angle. The system's high precision robotic control of position and orientation allow it to be used also to calibrate the geometry of the instrument's view. Combined with careful measurements of the narrow bandpass spectral filters, this system fully calibrates the diagnostic without any plasma discharges. The system's high repeatability, flexibility, and speed has been exploited to quantify several systematics in the MSE diagnostic response, providing a more complete understanding of the diagnostic performance.

  20. Impurity screening studies in the ALCATOR C-Mod tokamak

    International Nuclear Information System (INIS)

    Screening experiments have been undertaken in both limited and diverted discharges with a range of gaseous impurities in ohmic discharges. Measurements have been made as a function of plasma density and of the poloidal position of gas injection. It has been found that for recycling impurities such as neon and argon the number of impurities in the core plasma is proportional to the number injected. For non-recycling impurities (carbon and nitrogen) the number in the core is a function of the rate of injection. For discharges limited on the inner wall the screening is a function of the poloidal position of injection, with the injection at the inner wall giving the poorest screening. In diverted discharges with recycling impurities the position of injection does not significantly affect the screening. For non-recycling impurities the screening is typically a factor of 3 better when impurities are injected from the divertor rather than from the outside midplane. However, the best screening occurs when the impurities are injected at the inner midplane. Screening is typically a factor of 10 better for diverted than for limited discharges. Impurity transport has been modelled using the Monte Carlo code DIVIMP with a background plasma derived from experimental measurements of plasma parameters at the target and in the scrape-off layer (SOL). It is found that the code can reproduce the experimental measurements within a factor of 2. (author). 12 refs, 3 figs

  1. Disruption Mitigation Experiments with Two Gas Jets on Alcator C-Mod

    International Nuclear Information System (INIS)

    Full text: Massive gas injection (MGI) disruption mitigation experiments have shown that this technique can quickly convert a large fraction of plasma thermal and magnetic energy into radiated power. To date, gas has been injected from a single spatial location, and bolometric measurements have shown that the resulting radiated power is often toroidally asymmetric, which could cause melting of beryllium first wall surfaces in ITER. Therefore, the ITER MGI system proposes multiple gas jets distributed around the torus. On Alcator C-Mod, a 2nd gas jet has been installed 154 degrees around the torus from the existing gas jet. The hardware components of both gas jets are nominally identical. A pair of AXUV photodiode arrays viewing the plasma midplane is used to measure the n = 1 component of the toroidal asymmetry, and a toroidally-distributed set of six individual detectors, each viewing a collimated slice of the plasma, provides toroidal resolution higher than n = 1 and is used to calculate the toroidal peaking factor (TPF) of the radiated power. Experiments have begun to characterise the effect of using two jets on the radiation TPF, varying the relative timing between the firing of the gas jets from shot-to-shot. It is observed that the TPF depends on the phase of the disruption. During the pre-TQ, when the gas is cooling the plasma edge, the TPF varies reproducibly with the relative gas jet timing. However, during the thermal quench (TQ) and current quench (CQ), when most of the plasma energy is radiated, the results are more complicated. At large positive delay times (i.e., jet 2 fires well after jet 1) the TPF is seen to be variable and often high, agreeing with earlier results using only this jet. However, at large negative delay times (jet 2 fires well before jet 1), the TPF is significantly lower and more reproducible. This may indicate that slight differences in hardware and/or geometry between the two gas jet systems are important. The growth of n = 1 MHD

  2. EMC3-EIRENE modeling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

    2015-08-15

    Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

  3. DIVIMP modeling of impurity flows and screening in Alcator C-Mod

    International Nuclear Information System (INIS)

    We report on impurity transport modeling using the DIVIMP code which is able to qualitatively reproduce the poloidal variation of non-recycling impurity penetration factor (PFNR) found in C-Mod experiments: a lower PFNR is computed at the inboard (3.6%) and divertor target locations (0.7%) than at the outboard (11%). By artificially increasing the modeled inner SOL plasma flow to correspond to measured values, a better quantitative agreement between modeled and measured PFs is achieved. We have also roughly reproduced the observed penetration factor for recycling impurities both in time dependence and magnitude. The model has shown that under attached conditions, the majority of recycling impurity ions flow into the confined plasma through the outboard side separatrix. For detached conditions the impurity influx across the separatrix is more concentrated near the divertor

  4. ICRF Impurity Behavior with Boron Coated Molybdenum Tiles in Alcator C-Mod

    International Nuclear Information System (INIS)

    Full text: Although ion cyclotron range of frequency (ICRF) heating is considered an excellent candidate for bulk heating, minimizing impurity production associated with ICRF operation, particularly with metallic plasma facing components (PFC), remains one of the primary challenges for ICRF utilization. In C-Mod and present experiments, boronization, an in-situ applied boron film, is utilized to control impurities and its effectiveness has a limited lifetime. In C-Mod, the lifetime has been observed to be proportional to integrated injected RF Joules and the degradation is faster than in equivalent ohmic heated discharges the ICRF is enhancing the erosion rate of the boron film. In an effort to identify important erosion and impurity source locations, we have vacuum plasma sprayed ∼ 100 microns of boron on molybdenum tiles from the outer divertor shelf, main plasma limiters, and the RF antennas. We have also modified the shape of the main plasma limiter and increased our spectroscopic monitoring diagnostics of the main plasma limiter. Finally, we have installed a set of probes to monitor the plasma potential and RF fields on field lines connected an antenna. For ICRF heated H-modes, the core molybdenum levels was significantly reduced and remained at low levels for increased integrated injected RF Joules. The core molybdenum levels also no longer scales with RF power in L-mode in contrast with previous results with boronization and molybdenum plasma facing components. Initial Post campaign analysis of the boron coating will also be presented. Boronization and impurity, typically nitrogen or neon, seeded discharges enabled high plasma and ICRF antenna performance. The boronization suggests that other impurity sources are important but are yet to be identified. Impurity seeding had two important effects: reduced core molybdenum levels and suppressed antenna faults due to arcs and injections from antenna structure. The lower core molybdenum level is surprising since

  5. H-mode edge stability of Alcator C-mod plasmas

    International Nuclear Information System (INIS)

    For steady state H-mode operation, a relaxation mechanism is required to limit build-up of the edge gradient and impurity content. C-Mod sees two such mechanisms - EDA and grassy ELMs, but not large type I ELMs. In EDA the edge relaxation is provided by an edge localized quasi coherent electromagnetic mode that exists at moderate pedestal temperature T3.5 and does not limit the build up of the edge pressure gradient. The mode is not observed in the ideal MHD stability analysis, but is recorded in the nonlinear real geometry fluctuations modeling based on fluid equations and is thus tentatively identified as a resistive ballooning mode. At high edge pressure gradients and temperatures the mode is replaced by broadband fluctuations (f< 50 kHz) and small irregular ELMs are observed. Based on ideal MHD calculations that include the effects of edge bootstrap current, these ELMs are identified as medium n (10 < n < 50) coupled peeling/ballooning modes. The stability thresholds, its dependence on the plasma shape and the modes structure are studied experimentally and with the linear MHD stability code ELITE. (author)

  6. Diamagnetic measurements on the Alcator C tokamak

    International Nuclear Information System (INIS)

    A procedure for determining the total thermal energy content of a magnetically confined plasma from a measurement of the plasma magnetization has been successfully implemented on the Alcator C tokamak. When a plasma is confined by a magnetic field, the kinetic pressure of the plasma is supported by an interaction between the confining magnetic field and drift currents which flow in the plasma. These drift currents induce an additional magnetic field which can be measured by means of appropriately positioned pickup coils. From a measurement of this magnetic field and of the confining magnetic field, one can calculate the spatially averaged plasma pressure, which is related to the thermal energy content of the plasma by the equation of state of the plasma. The theory on which this measurement is based is described in detail. The fields and currents which flow in the plasma are related to the confining magnetic field and the plasma pressure by requiring that the plasma be in equilibrium, i.e., by balancing the forces due to pressure gradients against those due to magnetic interactions. The apparatus used to make this measurement is described and some example data analyses are carried out

  7. Interferometría láser heterodina para diagnóstico en plasmas de fusión. Experimentos y medidas realizados en el Stellarator TJ-II y el Tokamak C-MOD

    OpenAIRE

    Acedo Gallardo, Pablo

    2000-01-01

    La presente tesis describe la concepción, el diseño, la instalación y la obtención de primeros resultados de dos interferómetros láser heterodinos con dos longitudes de onda para la medida de densidades electrónicas en dos máquinas de fusión: el Stellarator T J-II (Centro de Investigaciones Energéticas, Mediambientales y Tecnológicas, Madrid) y el Tokamak Alcator C-Mod (Plasma Science and Fusion Center, MIT, EEUU). Este diagnóstico interferométrico de la densidad electrónica ba...

  8. Reduced-model (SOLT) simulations of an EDA H-mode shot at Alcator C-Mod

    Science.gov (United States)

    Russell, D. A.; D'Ippolito, D. A.; Myra, J. R.; Labombard, B.; Terry, J. L.; Zweben, S. J.

    2011-10-01

    Reduced-model scrape-off layer turbulence (SOLT) simulations of an Enhanced D-Alpha (EDA) H-mode observed at C-Mod were conducted to explore observed variations in scrape-off-layer (SOL) width. The amplitude of a mean poloidal flow was varied to control the level of turbulence in the simulation and to reproduce the observed heat flux across the separatrix. SOL width decreased with increasing input power and with increasing separatrix temperature in both experiment and simulation, consistent with the strong temperature dependence of collision-limited parallel heat flux. A persistent quasi-coherent mode (QCM) dominates the SOLT turbulence. The wavelength of the SOLT QCM is comparable to that of the QCM consistently observed on C-Mod during EDA operation. The SOLT QCM consists of a quasi-stationary string of vortices, located just inside the separatrix, poloidally convected by the mean flow and occasionally emitting blobs into the SOL. The mode frequency is dominated by the Doppler shift of this convected pattern. Analysis reveals underlying drift-interchange and Kelvin-Helmholtz instabilities. Supported by USDOE under DE-FG02-97ER54392, DE-AC02-09CH11466, DE-FC02-99ER54512 and S009625-F.

  9. Theoretical studies of lower hybrid current drive in the Alcator C tokamak

    International Nuclear Information System (INIS)

    Theoretical studies of lower hybrid current drive in the Alcator C tokamak using a simulation model is presented, the model incorporates a 1-D radial transport code to solve for the time evolution of the bulk plasma quantities

  10. Heat-flux footprints for I-mode and EDA H-mode plasmas on Alcator C-Mod

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Hughes, J. W.; Reinke, M. L.; Whyte, D. G.

    2013-07-01

    IR thermography is used to measure the heat flux footprints on C-Mod's outer target in I-mode and EDA H-mode plasmas. The footprint profiles are fit to a function with a simple physical interpretation. The fit parameter that is sensitive to the power decay length into the SOL, λSOL, is ˜1-3× larger in I-modes than in H-modes at similar plasma current, which is the dominant dependence for the H-mode λSOL. In contrast, the fit parameter sensitive to transport into the private-flux-zone along the divertor leg is somewhat smaller in I-mode than in H-mode, but otherwise displays no obvious dependence on Ip, Bt, or stored energy. A third measure of the footprint width, the "integral width", is not significantly different between H- and I-modes. Also discussed are significant differences in the global power flows of the H-modes with "favorable"∇B drift direction and those of the I-modes with "unfavorable"∇B drift direction.

  11. Radiofrequency-heated enhanced confinement modes in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Y.; Boivin, R.L.; Bombarda, F.; Bonoli, P.T.; Christensen, C.; Fiore, C.; Garnier, D.; Goetz, J.A.; Golovato, S.N.; Granetz, R.; Greenwald, M.; Horne, S.F.; Hubbard, A.; Hutchinson, I.H.; Irby, J.; LaBombard, B.; Lipschultz, B.; Marmar, E.; May, M.; Mazurenko, A.; McCracken, G.; OShea, P.; Porkolab, M.; Reardon, J.; Rice, J.; Rost, C.; Schachter, J.; Snipes, J.A.; Stek, P.; Terry, J.; Watterson, R.; Welch, B.; Wolfe, S. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1997-05-01

    Enhanced confinement modes up to a toroidal field of B{sub T}=8T have been studied with up to 3.5 MW of radiofrequency (rf) heating power in the ion cyclotron range of frequencies (ICRF) at 80 MHz. H-mode is observed when the edge temperature exceeds a threshold value. The high confinement mode (H-mode) with higher confinement enhancement factors (H) and longer duration became possible after boronization by reducing the radiated power from the main plasma. A quasi-steady state with high confinement (H=2.0), high normalized beta ({beta}{sub N}=1.5), low radiated power fraction (P{sub rad}{sup main}/P{sub loss}=0.3), and low effective charge (Z{sub eff}=1.5) has been obtained in Enhanced D{sub {alpha}} H-mode. This type of H-mode has enhanced levels of continuous D{sub {alpha}} emission and very little or no edge localized mode (ELM) activity, and reduced core particle confinement time relative to ELM-free H-mode. The pellet enhanced performance (PEP) mode is obtained by combining core fueling with pellet injection and core heating. A highly peaked pressure profile with a central value of 8 atmospheres was observed. The steep pressure gradient drives off-axis bootstrap current, resulting in a shear reversed safety factor (q) profile. Suppression of sawteeth appears to be important in maintaining the highly peaked pressure profile. Lithium pellets were found to be more effective than deuterium pellets in raising q{sub 0}. {copyright} {ital 1997 American Institute of Physics.}

  12. Soft x-ray tomography on the Alcator C tokamak

    International Nuclear Information System (INIS)

    A soft x-ray tomography experiment has been performed on the Alcator C tokamak. An 80-chord array of detectors consisting of miniature PIN photodiodes was used to obtain tomographic reconstructions of the soft x-ray emissivity function's poloidal cross-section. The detectors are located around the periphery of the plasma at one toroidal location (top and bottom ports) and are capable of yielding useful information over a wide range of plasma operating parameters and conditions. The reconstruction algorithm employed makes no assumption whatsoever about plasma rotation, position, or symmetry. Its performance was tested, and it was found to work well and to be fairly insensitive to estimated levels of random and systematic errors in the data

  13. Ion Bernstein wave experiments on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Ion Bernstein wave experiments are carried out on the Alcator C tokamak to study wave excitation, propagation, absorption, and plasma heating due to wave power absorption. It is shown that ion Bernstein wave power is coupled into the plasma and follows the expected dispersion relation. The antenna loading is maximized when the hydrogen second harmonic layer is positioned just behind the antenna. Plasma heating results at three values of the toroidal magnetic field are presented. Central ion temperature increases of ΔT/sub i//Ti /approx lt/ 0.1 and density increases Δn/n 6s/sup /minus/1/ for plasmas within the density range 0.6 /times/ 1020m/sup /minus/3/ ≤ /bar n//sub e/ ≤ 4 /times/ 1020m/sup /minus/3/ and magnetic fields 2.4 ≥ ω/Ω/sub H/ ≥ 1.1. The density increases is usually accompanied by an improvement in the global particle confinement time relative to the Ohmic value. The ion heating rate is measured to be ΔT/sub i//P/sub rf/ ≅ 2-4.5 eV/kW at low densities. At higher densities /bar n//sub e/ ≤ 1.5 /times/ 1020m/sup /minus/3/ the ion heating rate dramatically decreases. It is shown that the decrease in the ion heating rate can be explained by the combined effects of wave scattering through the edge turbulence and the decreasing on energy confinement of these discharges with density. The effect of observed edge turbulence is shown to cause a broadening of the rf power deposition profile with increasing density. It is shown that the inferred value of the Ohmic ion thermal conduction, when compared to the Chang-Hinton neoclassical prediction, exhibits an increasing anomaly with increasing plasma density

  14. Observation of Lower-Hybrid Current Drive at High Densities in the Alcator C Tokamak

    Science.gov (United States)

    Porkolab, M.; Schuss, J. J.; Lloyd, B.; Takase, Y.; Texter, S.; Bonoli, P.; Fiore, C.; Gandy, R.; Gwinn, D.; Lipschultz, B.; Marmar, E.; Pappas, D.; Parker, R.; Pribyl, P.

    1984-07-01

    A quasi-steady-state lower-hybrid current-drive operation is demonstrated in the Alcator C tokamak at densities up to n―e~=1×1014 cm-3. The current-drive efficiency is measured experimentally over a wide range of densities and magnetic fields. The radial distribution of high-energy x rays indicates that the current-carrying electrons peak near the plasma axis.

  15. Energy Confinement of High-Density Pellet-Fueled Plasmas in the Alcator C Tokamak

    Science.gov (United States)

    Greenwald, M.; Gwinn, D.; Milora, S.; Parker, J.; Parker, R.; Wolfe, S.; Besen, M.; Camacho, F.; Fairfax, S.; Fiore, C.; Foord, M.; Gandy, R.; Gomez, C.; Granetz, R.; Labombard, B.; Lipschultz, B.; Lloyd, B.; Marmar, E.; McCool, S.; Pappas, D.; Petrasso, R.; Pribyl, P.; Rice, J.; Schuresko, D.; Takase, Y.; Terry, J.; Watterson, R.

    1984-07-01

    A series of pellet-fueling experiments has been carried out on the Alcator C tokamak. High-speed hydrogen pellets penetrate to within a few centimeters of the magnetic axis, raise the plasma density, and produce peaked density profiles. Energy confinement is observed to increase over similar discharges fueled only by gas puffing. In this manner record values of electron density, plasma pressure, and Lawson number (n τ) have been achieved.

  16. Preparing the Alcator C bolometer system for use on MTX (Microwave Tokamak Experiment)

    Science.gov (United States)

    Marinak, Marty

    1988-02-01

    The Alcator C bolometer array has been modified to be compatible with electron cyclotron heating on the Microwave Tokamak Experiment. Fine wire mesh screens are mounted on the front of the bolometer collimator tubes to attenuate microwave heating of the bolometers. Structural changes eliminate openings in the seams of the bolometer housing, which represent pathways for microwaves to enter the system. This paper outlines the operational principles of the bolometer system, discusses the measured and predicted performance characteristics of the bolometer array, and includes a concise guide to the operation of the bolometer controller.

  17. High gain free electron laser for heating and current drive in the ALCATOR-C tokamak

    International Nuclear Information System (INIS)

    The free electron laser (FEL) particle simulation code, FRED, has been used to examine the design of an FEL for amplifying radiation in the one to two millimeter wavelength range for use in electron heating and current drive in a tokamak device such as ALCATOR-C. As a desired design goal a peak output power of 8 GW, with a minimum input power in the 1 to 100 watt range has been used. The effects of electron beam current, energy and brightness, laser frequency and input power as well as wiggler wavelength and overall wiggler length on the performance of the FEL have been examined

  18. Lower hybrid heating in the Alcator A tokamak

    Science.gov (United States)

    Schuss, J. J.; Porkolab, M.; Takase, Y.; Cope, D.; Fairfax, S.; Greenwald, M.; Gwinn, D.; Hutchinson, I. H.; Kusse, B.; Marmar, E.

    1981-04-01

    The results of the moderate-power (P less than 100 kW) Alcator A lower-hybrid-heating experiment are presented. In this experiment, RF power densities of up to 8 kW per sq cm were achieved for 40-ms pulses. Both electron and ion heating were observed as the plasma density was varied. No impurity influx or density rise was observed because of the RF pulse. The ion heating, however, as evidenced by the formation of an energetic ion tail in the plasma center, occurred at a lower plasma density than expected from linear plasma wave theory. Furthermore, the electron heating observed at lower densities was not expected from linear waveguide-plasma coupling theory. Contrary to expectations, the ion heating was found to be independent of the waveguide phasing. These results, together with RF probe measurements in the edge plasma, suggest that the lower hybrid waves launched by the array may undergo strong scattering from parametric instabilities or density fluctuations near the plasma edge.

  19. RF heating and current drive experiments on the Alcator C and Versator II tokamaks

    International Nuclear Information System (INIS)

    Lower hybrid heating and current drive experiments on the Alcator C tokamak (R = 0.64 m, a = 0.165 m, molybdenum limiters) were performed at a frequency of 4.6 GHz with net injected rf powers up to P/sub rf/ ≤ 1.5 MW. Recent experiments have focused on energy confinement studies in lower hybrid current-driven (LHCD) and LHRF heated Ohmic discharges, and sawtooth stabilization in combined LHCD-OH driven discharges at densities n-bar/sub e/ ≤ 1.4 x 1020 m-3. Ion Bernstein wave heating experiments were also carried out in Alcator C at a frequency of f = 183 MHz at power levels P/sub rf/ ≤ 200 kW. Significant heating (ΔT/sub i/ ≤ 400 eV) was observed at ω/ω/sub CH/ ≅ 1.5, 2.5 and ω/ω/sub CD/ ≅ 2.5 at densities n-bar/sub e/ ≅ 1 x 1020 m-3. In the Versator II tokamak, particle confinement improvement (by factors of ≅2) was observed in the presence of 2.45 GHz lower hybrid current drive

  20. RF heating and current drive experiments on the Alcator C and Versator II tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Porkolab, M.; Bonoli, P.; Chen, K.I.; Fiore, C.; Granetz, R.; Griffin, D.; Gwinn, D.; Knowlton, S.; Lipschultz, B.; Luckhardt, S.C.

    1986-11-01

    Lower hybrid heating and current drive experiments on the Alcator C tokamak (R = 0.64 m, a = 0.165 m, molybdenum limiters) were performed at a frequency of 4.6 GHz with net injected rf powers up to P/sub rf/ less than or equal to 1.5 MW. Recent experiments have focused on energy confinement studies in lower hybrid current-driven (LHCD) and LHRF heated Ohmic discharges, and sawtooth stabilization in combined LHCD-OH driven discharges at densities n-bar/sub e/ less than or equal to 1.4 x 10/sup 20/ m/sup -3/. Ion Bernstein wave heating experiments were also carried out in Alcator C at a frequency of f = 183 MHz at power levels P/sub rf/ less than or equal to 200 kW. Significant heating (..delta..T/sub i/ less than or equal to 400 eV) was observed at ..omega../..omega../sub CH/ approx. = 1.5, 2.5 and ..omega../..omega../sub CD/ approx. = 2.5 at densities n-bar/sub e/ approx. = 1 x 10/sup 20/ m/sup -3/. In the Versator II tokamak, particle confinement improvement (by factors of approx. =2) was observed in the presence of 2.45 GHz lower hybrid current drive.

  1. Molybdenum density profiles on C-Mod using FAC generated cooling curves

    Science.gov (United States)

    Reinke, M.

    2005-10-01

    For tokamaks with high-Z plasma facing components, maintaining a low impurity content is necessary to produce high quality, repeatable discharges. A GENeral Impurity Emissivity (GENIE) method is outlined for determining impurity profiles using experimental spectroscopy data, an impurity transport code, and the atomic physics package, Flexible Atomic Code (FAC). Modular programming is emphasized in order to make the method extendable to arbitrary impurities, diagnostic sets and tokamaks. Development of GENIE is ongoing, but a necessary first step is to verify FAC. A testing stage of GENIE that ignores transport is demonstrated and the results are validated against the published molybdenum cooling-curve generated using HULLAC. Bolometry and Thomson scattering data are used to determine molybdenum density profiles on Alcator C-Mod using the Mo cooling-curve. Instances where this method fails are shown as well to illustrate the need for a more advanced version of GENIE that generates and uses charge state distributions that assume transport.

  2. Experimental study of visible and ultraviolet impurity emission from the Alcator C tokamak

    International Nuclear Information System (INIS)

    Densities of carbon, oxygen, and silicon in Alcator C tokamak plasmas have been computed from spectroscopic measurements of the absolute brightnesses of visible and ultraviolet emission lines in combination with a one dimensional numerical calculation which models the charge state and emissivity profiles. Profiles of all the charge states of a particular impurity were calculated by utilizing independent measurements of plasma density and temperature and solving the coupled system of transport and rate equations connecting the ionization states. These profiles were then used to calculate emissivity and brightness profiles by solving the matrix equation relating the level populations through atomic processes such as electron impact excitation, de-excitation, spontaneous emission and cascades from upper levels. Good agreement was found between predicted impurity line brightnesses and experimentally measured brightnesses of different charge states. Three different types of limiter materials, molybdenum, graphite and SiC coated graphite have been used on Alcator C. It was determined that the principal impurities in the plasma, under most conditions, depends upon the type of limiter being used. However, the sources of the impurities are both the wall and the limiters, since it was observed that the wall becomes coated with limiter material due to plasma discharges

  3. Poloidal asymmetries in the limiter shadow plasma of the Alcator C tokamak. Volume 1

    International Nuclear Information System (INIS)

    This thesis investigates conditions which exist in the limiter shadow plasma of the Alcator C tokamak. The understanding of this edge plasma region is approached from both experimental and theoretical points of view. First, a general overview of edge plasma physical processes is presented. Simple edge plasma models and conditions which can theoretically result in a poloidally asymmetric edge plasma are discussed. A review of data obtained from previous diagnostics in the Alcator C edge plasma is then used to motivate the development of a new edge plasma diagnostic system (DENSEPACK) to experimentally investigate poloidal asymmetries in this region. The bulk of this thesis focuses on the marked poloidal asymmetries detected by this poloidal probe array and possible mechanisms which might support such asymmetries on a magnetic flux surface. In processing the probe data, some important considerations on fitting Langmuir probe characteristics are identified. The remainder of this thesis catalogues edge versus central plasma parameter dependences. Regression analysis techniques are applied to characterize edge density for various central plasma parameters. Edge plasma conditions during lower hybrid radio frequency heating and pellet injection are also discussed

  4. Lower-hybrid-heating experiments on the Alcator C and the Versator II Tokamaks

    Science.gov (United States)

    Porkolab, M.; Schuss, J. J.; Takase, Y.; Texter, S.; Fiore, C. L.; Gandy, R.; Greenwald, M. J.; Gwinn, D. A.; Lipschultz, B.; Marmar, E. S.

    Initial results from lower hybrid wave heating experiments carried out on the MIT Alcator-C and Versator II Tokamak are reported. In the Alcator-C experiments a 4 waveguide array, with internally brazed ceramic windows was used to inject 160 kW of microwave power at 4.6 GHz into the plasma with nO less than or equal to 1 x 10(15) cm(+3), and BO less than or equal to 12 T. The RF coupling studies show optimal coupling when the local density at the waveguide mouth is 25 to 50 times overdense. Heating experiments show an ion tail formation in hydrogen discharge peaking at a density of anti-n approx. = 2.7 x 10(14) cm(+3) at B = 8.9 T, and bulk ion heating at a density of anti n approx. = 1.5 x 10(14) c(+3) at B approx. = 11 T. Evidence of RF current enhancement has been observed at a density of n approx. = 3 x 10(13) cm (+3). Doppler broadening of the OVII and NVI lines shows a (RADICAL)T/sub i/= 50 eV rise in the bulk ion temperature. A significant RF produced ion tail is also observed by charge exchange analysis. A toroidal ray tracing code and a 1-D transport code to study the heating density bands and heating efficiencies were successfully combined.

  5. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  6. Direct detection of lower hybrid wave using a reflectometer on Alcator C-Moda)

    Science.gov (United States)

    Shiraiwa, S.; Baek, S.; Dominguez, A.; Marmar, E.; Parker, R.; Kramer, G. J.

    2010-10-01

    The possibility of directly detecting a density perturbation produced by lower hybrid (LH) waves using a reflectometer is presented. We investigate the microwave scattering of reflectometer probe beams by a model density fluctuation produced by short wavelength LH waves in an Alcator C-Mod experimental condition. In the O-mode case, the maximum response of phase measurement is found to occur when the density perturbation is approximately centimeters in front of the antenna, where Bragg scattering condition is satisfied. In the X-mode case, the phase measurement is predicted to be more sensitive to the density fluctuation close to the cut-off layer. A feasibility test was carried out using a 50 GHz O-mode reflectometer on the Alcator C-Mod tokamak, and positive results including the detection of 4.6 GHz pump wave and parametric decay instabilities were obtained.

  7. Diagnosis of mildly relativistic electron velocity distributions by electron cyclotron emission in the Alcator C tokamak

    International Nuclear Information System (INIS)

    Mildly relativistic electron velocity distributions are diagnosed from measurements of the first few electron cyclotron emission harmonics in the Alcator C tokamak. The approach employs a vertical viewing chord through the center of the tokamak plasma terminating at a compact, high-performance viewing dump. The cyclotron emission spectra obtained in this way are dominated by frequency downshifts due to the relativistic mass increase, which discriminates the electrons by their total energy. In this way a one-to-one correspondence between the energy and the emission frequency is accomplished in the absence of harmonic superpositions. The distribution, described by f/sub p/, the line-averaged phase space density, and Λ, the anisotropy factor, is determined from the ratio of the optically thin harmonics or polarizations. Diagnosis of spectra in the second and the third harmonic range of frequencies obtained during lower hybrid heating, current drive, and low density ohmic discharges are carried out, using different methods depending on the degree of harmonic superposition present in the spectrum and the availability of more than one ratio measurement. Discussions of transient phenomena, the radiation temperature measurement from the optically thick first harmonic, and the measurements compared to the angular hard x-ray diagnostic results illuminate the capabilities of the vertically viewing electron cyclotron emission diagnostic

  8. Lower hybrid current drive experiments on the MIT Alcator C and Versator II tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Knowlton, S.; McDermott, S.; Porkolab, M.; Takase, Y.; Texter, S.; Bonoli, P.; Fiore, C.; McCool, S.; Mayberry, M.; Chen, K.I.

    1985-08-01

    Energy confinement studies in lower hybrid RF driven discharges at 4.6 GHz have been carried out on the Alcator C tokamak. The electron temperature profile is measured by a five point Thomson scattering system and the ion temperature by charge-exchange analysis. The energy content of the bulk plasma is found to be similar for RF-driven and ohmic discharges of identical current and density. In the parameter range anti n/sub e/ = 3 - 7 x 10/sup 13/ cm/sup -3/, B = 7 - 11 T, I/sub p/ = 100 - 200 kA, q (a) > 8, the RF power needed to sustain the discharge is significantly greater than the ohmic power required to maintain a similar plasma. The gross energy confinement time is lower in the RF-driven discharges than in the ohmic ones by a factor of 1.5 to 4, depending on plasma conditions. The frequency scaling of the density limit for current drive is reported from the Versator II tokamak. The steady-state current drive density limit of anti n/sub e/ = 6 x 10/sup 12/ cm/sup -3/ at 800 MHz. has been raised to a density of at least anti n/sub e/ = 1.0 x 10/sup 13/ cm/sup -3/ at the same toroidal field by operations at a frequency of 2.45 GHz. Superthermal electron effects during RF injection are observed up to a density of anti n/sub e/ = 2.5 x 10/sup 13/ cm/sup -3/.

  9. Measurement of the current density profile in the Alcator C tokamak using lithium pellets

    International Nuclear Information System (INIS)

    High-speed lithium pellets have been injected into Alcator C tokamak plasmas in order to measure the internal magnetic field, and thus current density profiles. In the pellet ablation cloud, intense visible line radiation from the Li+ ion (λ∼5485 A, 1s2s 3S-1s2p 3P) is polarized due to the Zeeman effect, and measurement of the polarization angle yields the direction of the total local magnetic field. A ''snap shot'' of the q profile is obtained as the pellet penetrates from the edge into the center of the discharge, in a time of about 300 μs. The spatial resolution of the measurement is about 1 cm. At a toroidal field of BT=10 T, the emission in the unshifted π component of the Zeeman triplet is more than 80% polarized, and q profiles have been obtained. The pellets are perturbative (left-angle Δne right-angle/left-angle ne right-angle ∼1), but the total pellet penetration time is at least a factor of 1000 smaller than the classical skin time. It can thus be anticipated that the current density profile should not be perturbed significantly during the time of the measurement. With some relatively straightforward modifications and refinements, precision approaching 10% for the measurement of q profiles should be achievable. The technique appears viable, using Li, as long as the toroidal field is approx-gt 4 T

  10. Soft x-ray tomography diagnostic for the Alcator C tokamak

    International Nuclear Information System (INIS)

    A soft x-ray tomography diagnostic has been built and operated on the Alcator C tokamak with the use of an 80-channel array of detectors. The detector system consists of miniature PIN photodiodes which are located around the periphery of the plasma at one toroidal location (top and bottom ports) in order to image the poloidal cross section of the plasma's soft x-ray emissivity. The compact size and inexpensive cost of these detectors allow for the large number of views required to avoid any assumptions about plasma rotation, position, or symmetry in the reconstruction algorithm. The use of logarithmic amplifiers in the electronic circuitry makes it possible to record conveniently data from a wide range of plasma operating parameters. This diagnostic has been used to study plasma position and shape, magnetohydrodynamic (MHD) phenomena, and the effects of impurities. Data from a pellet-fueled discharge exhibiting large-scale MHD activity are presented. The system has recently been expanded to 128 detectors with the addition of an array on a side port in order to improve its coverage of the plasma and the resolution of the reconstructions

  11. A fast neutron spectrometer for D-D fusion neutron measurements at the Alcator C tokamak

    Science.gov (United States)

    Fisher, W. A.; Chen, S. H.; Gwinn, D.; Parker, R. R.

    1984-01-01

    A neutron spectrometer using a high pressure 3He ionization chamber has been designed and used to measure the neutron spectrum from an ohmically heated deuterium plasma. The resolution of the spectrometer at 2.45 MeV is determine to be 46 keV full width at half-maximum (fwhm). Particular attention has been paid to optimizing the detector shielding and collimation to reject thermal and epithermal neutrons scattered from the tokamak structure. As a result, measurements indicate that the ratio of the number of counts in the 2.45 MeV peak to the total number of detected neutron events is {1}/{67}. For the 8 μs amplifier time constant used, a count rate as high as 44 counts per second has been achieved in the thermonuclear peak. The observed spectra have been compared with calculated spectra using the MCNP Monte Carlo Neutral Particle Transport code and they show good agreement. There is little evidence of neutrons produced from photoneutron reactions or electrodisintegration. It has been possible to confirm that the shape of the thermonuclear peak is consistent with the Gaussian shape predicted and that the ion temperature as determined from the line width is consistent with other Alcator C ion temperature diagnostics, and follows the trends predicted by the theory of Doppler line broadening.

  12. Visible continuum measurements on the Alcator C Tokamak: Changes in particle transport during pellet fuelled discharges

    International Nuclear Information System (INIS)

    A spatially resolving visible light detector system is used to measure continuum radiation near 5360A on the Alcator C Tokamak. For the typically hot plasmas studied, the continuum emission is found to be dominated by bremsstrahlung radiation near this wavelength region. Accurate determinations of Z/sub eff/ are obtained from continuum measurements using independently determined temperature and density measurements. Density profiles during high density, clean pellet fueled discharges, are also determined and are used to study the changes in particle transport after injection. For discharges with sufficiently large pellet density increases, density profiles are found to become more peaked following the injection. In these cases, the profiles are found to remain peaked for the remainder of the discharge, or until a ''giant'' sawtooth or minor disruption abruptly returns the profiles to a flatter pre-pellet condition. Analysis of density profiles after pellet injection yields information about the radial diffusion and convection velocity of the plasma particles. The peakedness in the density profiles, observed after pellet injection, is attributable mostly to increases in inward convection. It is concluded that neoclassical fluxes are too small to account for these changes. 70 refs., 55 figs

  13. Impurity generation during intense lower hybrid heating experiments on the Alcator C tokamak

    Science.gov (United States)

    Marmar, E.; Foord, M.; Labombard, B.; Lipschultz, B.; Moreno, J.; Rice, J.; Terry, J.; Lloyd, B.; Porkolab, M.; Schuss, J.; Takase, Y.; Texter, S.; Fiore, C.; Gandy, R.; Granetz, R.; Greenwald, M.; Gwinn, D.; McCool, S.; Pappas, D.; Parker, R. R.; Pribyl, P.; Watterson, R.; Wolfe, S. M.

    1984-05-01

    Experiments are underway on the Alcator C tokamak with over 1 MW of RF power injected into the plasma at a frequency of 4.6 GHz to study both heating and current drive effects. During these studies, impurity generation from limiter structures has been observed. The RF induced impurity influx is a strongly nonlinear function of net injected power. For PRF < 500 kW, only small effects are seen. As PRF approaches 1 MW, however, sharp increases in impurity influxes and Zeff are observed. Three different limiter materials have been used during these studies: molybdenum, graphite, and silicon-carbide coated graphite. In each case, the materials of the limiter structure are seen to dominate the increased impurity influx. In a typical case, with P RF = 1.0 MW, overlinene = 1.3 × 10 14cm-3, and the SiC coated limiters, Zeff is seen to increase from 1.5 before the RF pulse to about 4 during the heating. At the same time, central Te increases from 2000 to 3000 eV and central Ti from 1200 to 1800 eV. Similar effects are seen in both H 2 and D 2 working gas discharges. The contribution to impurity generation of nonthermal electrons, which are produced by the RF, is under investigation. Changes in edge plasma temperature and density, as well as the possibility that the particle transport is affected by the RF, are also being examined. Results of the experiments with the three different limiter materials are compared, and contributions of impurity radiation to the overall power balance are estimated.

  14. Fast wave ion cyclotron resonance heating experiments on the Alcator C tokamak

    International Nuclear Information System (INIS)

    Minority regime fast wave ICRF heating experiments have been conducted on the Alcator C tokamak at rf power levels sufficient to produce significant changes in plasma properties, and in particular to investigate the scaling to high density of the rf heating efficiency. Up to 450 kW of rf power at frequency f = 180 MHz, was injected into plasmas composed of deuterium majority and hydrogen minority ion species at magnetic field B0 = 12 T, density 0.8 ≤ /bar n/sub e// ≤ 5 /times/ 1020 m-3, ion temperature T/sub D/(0) /approximately/ 1 keV, electron temperature T/sub e/(0) /approximately/ 1.5--2.5 keV, and minority concentration 0.25 /approx lt/ /eta/sub H// ≤ 8%. Deuterium heating ΔT/sub D/(0) = 400 eV was observed at /bar n/sub e// = 1 /times/ 1020 m-3, with smaller temperature increases at higher density. However, there was no significant change in electron temperature and the minority temperatures were insufficient to account for the launched rf power. Minority concentration scans indicated most efficient deuterium heating at the lowest possible concentration, in apparent contradiction with theory. Incremental heating /tau/sub inc// /equivalent to/ ΔW/ΔP up to 5 ms was independent of density, in spite of theoretical predictions of favorable density scaling of rf absorption and in stark contrast to Ohmic confinement times /tau/sub E// /equivalent to/ W/P. After accounting for mode conversion and minority losses due to toroidal field ripple, unconfined orbits, asymmetric drag, neoclassical and sawtooth transport, and charge-exchange, it was found that the losses as well as the net power deposition on deuterium do scale very favorably with density. Nevertheless, when the net rf and Ohmic powers deposited on deuterium are compared, they are found to be equally efficient at heating the deuterium. 139 refs

  15. Simulated plasma facing component measurements for an in situ surface diagnostic on Alcator C-Moda)

    Science.gov (United States)

    Hartwig, Z. S.; Whyte, D. G.

    2010-10-01

    The ideal in situ plasma facing component (PFC) diagnostic for magnetic fusion devices would perform surface element and isotope composition measurements on a shot-to-shot (˜10 min) time scale with ˜1 μm depth and ˜1 cm spatial resolution over large areas of PFCs. To this end, the experimental adaptation of the customary laboratory surface diagnostic—nuclear scattering of MeV ions—to the Alcator C-Mod tokamak is being guided by ACRONYM, a Geant4 synthetic diagnostic. The diagnostic technique and ACRONYM are described, and synthetic measurements of film thickness for boron-coated PFCs are presented.

  16. Spectral measurements of fluctuating ω/sub pe/ radiation from Alcator C tokamak

    International Nuclear Information System (INIS)

    High resolution spectral measurements have been made of the fluctuating electron plasma frequency (ω/sub pe/) radiation from Alcator C. Three techniques have been used in making the measurements. Features as narrow as 350 kHz have been observed (Δf/f approx. = 6 x 10-6), impling that a highly coherent process is responsible for the emission

  17. Two dimensional radiated power diagnostics on Alcator C-Moda)

    Science.gov (United States)

    Reinke, M. L.; Hutchinson, I. H.

    2008-10-01

    The radiated power diagnostics for the Alcator C-Mod tokamak have been upgraded to measure two dimensional structure of the photon emissivity profile in order to investigate poloidal asymmetries in the core radiation. Commonly utilized unbiased absolute extreme ultraviolet (AXUV) diode arrays view the plasma along five different horizontal planes. The layout of the diagnostic set is shown and the results from calibrations and recent experiments are discussed. Data showing a significant, 30%-40%, inboard/outboard emissivity asymmetry during ELM-free H-mode are presented. The ability to use AXUV diode arrays to measure absolute radiated power is explored by comparing diode and resistive bolometer-based emissivity profiles for highly radiative L-mode plasmas seeded with argon. Emissivity profiles match in the core but disagree radially outward resulting in an underprediction of Prad of nearly 50% by the diodes compared to Prad determined using resistive bolometers.

  18. Production of internal transport barriers via self-generated mean flows in Alcator C-Moda)

    Science.gov (United States)

    Fiore, C. L.; Ernst, D. R.; Podpaly, Y. A.; Mikkelsen, D.; Howard, N. T.; Lee, Jungpyo; Reinke, M. L.; Rice, J. E.; Hughes, J. W.; Ma, Y.; Rowan, W. L.; Bespamyatnov, I.

    2012-05-01

    New results suggest that changes observed in the intrinsic toroidal rotation influence the internal transport barrier (ITB) formation in the Alcator C-Mod tokamak [E. S. Marmar and Alcator C-Mod group, Fusion Sci. Technol. 51, 261 (2007)]. These arise when the resonance for ion cyclotron range of frequencies (ICRF) minority heating is positioned off-axis at or outside of the plasma half-radius. These ITBs form in a reactor relevant regime, without particle or momentum injection, with Ti ≈ Te, and with monotonic q profiles (qmin 1.5 × 105 rad/s) in the region where the ITB foot is observed. Gyrokinetic analyses indicate that this spontaneous shearing rate is comparable to the linear ion temperature gradient (ITG) growth rate at the ITB location and is sufficient to reduce the turbulent particle and energy transport. New and detailed measurement of the ion temperature demonstrates that the radial profile flattens as the ICRF resonance position moves off axis, decreasing the drive for the ITG the instability as well. These results are the first evidence that intrinsic rotation can affect confinement in ITB plasmas.

  19. Energy confinement studies of lower hybrid current driven discharges in the Alcator C tokamak

    International Nuclear Information System (INIS)

    The energy confinement properties of purely RF-driven plasmas on Alcator C are being investigated by experimental measurements of the bulk electron and ion temperature profiles, and by numerical modelling of the lower hybrid wave propagation and thermal energy transport. Power balance studies are performed on plasmas with parameters n-bar/sub e/ = 3-7 x 1013 cm-3, B = 7-11 T, I/sub p/ = 100-200 kA, and q(a)> or =8, and with RF powers up to 1 MW at 4.6 GHz. The temperature mesurements from RF-driven discharges are compared with those from similar ohmic discharges (identical current and density). The gross energy confinement time, defined by tau/sub E/03/2 (Σ∫n/sub j/T/sub j/ dV)/P/sub i//sub n/, is lower in the RF-driven discharges than in the ohmic ones by a factor of 1.5 to 4, depending on the plasma conditions and RF power. While tau/sub E/ in ohmic discharges increases with density and is independent of the toroidal field, the confinement time in the RF-driven discharges decreases with RF power, is independent of density in the range n-bar/sub e/ = 3-7 x 1013cm-3, and increases with the toroidal magnetic field. The confinement degrades slightly with increasing current in both RF-driven and ohmic discharges. The code used to simulate these results employs a lower hybrid ray tracing package, a Fokker-Planck code for the evolution of the fast electron tail, and enhanced thermal transport models for auxiliary-heated plasmas. In the code simulations, more than 80% of the injected RF power is absorbed by electron Landau damping in the inner half of the plasma column, the remainder being dissipated by collisions near the plasma edge

  20. A new interferometry-based electron density fluctuation diagnostic on Alcator C-Moda)

    Science.gov (United States)

    Kasten, C. P.; Irby, J. H.; Murray, R.; White, A. E.; Pace, D. C.

    2012-10-01

    The two-color interferometry diagnostic on the Alcator C-Mod tokamak has been upgraded to measure fluctuations in the electron density and density gradient for turbulence and transport studies. Diagnostic features and capabilities are described. In differential mode, fast phase demodulation electronics detect the relative phase change between ten adjacent, radially-separated (ΔR = 1.2 cm, adjustable), vertical-viewing chords, which allows for measurement of the line-integrated electron density gradient. The system can be configured to detect the absolute phase shift of each chord by comparison to a local oscillator, measuring the line-integrated density. Each chord is sensitive to density fluctuations with kR < 20.3 cm-1 and is digitized at up to 10 MS/s, resolving aspects of ion temperature gradient-driven modes and other long-wavelength turbulence. Data from C-Mod discharges is presented, including observations of the quasi-coherent mode in enhanced D-alpha H-mode plasmas and the weakly coherent mode in I-mode.

  1. Observation of Energetic Particle Driven Modes Relevant to Advanced Tokamak Regimes

    Energy Technology Data Exchange (ETDEWEB)

    R. Nazikian; B. Alper; H.L. Berk; D. Borba; C. Boswell; R.V. Budny; K.H. Burrell; C.Z. Cheng; E.J. Doyle; E. Edlund; R.J. Fonck; A. Fukuyama; N.N. Gorelenkov; C.M. Greenfield; D.J. Gupta; M. Ishikawa; R.J. Jayakumar; G.J. Kramer; Y. Kusama; R.J. La Haye; G.R. McKee; W.A. Peebles; S.D. Pinches; M. Porkolab; J. Rapp; T.L. Rhodes; S.E. Sharapov; K. Shinohara; J.A. Snipes; W.M. Solomon; E.J. Strait; M. Takechi; M.A. Van Zeeland; W.P. West; K.L. Wong; S. Wukitch; L. Zeng

    2004-10-21

    Measurements of high-frequency oscillations in JET [Joint European Torus], JT-60U, Alcator C-Mod, DIII-D, and TFTR [Tokamak Fusion Test Reactor] plasmas are contributing to a new understanding of fast ion-driven instabilities relevant to Advanced Tokamak (AT) regimes. A model based on the transition from a cylindrical-like frequency-chirping mode to the Toroidal Alfven Eigenmode (TAE) has successfully encompassed many of the characteristics seen in experiments. In a surprising development, the use of internal density fluctuation diagnostics has revealed many more modes than has been detected on edge magnetic probes. A corollary discovery is the observation of modes excited by fast particles traveling well below the Alfven velocity. These observations open up new opportunities for investigating a ''sea of Alfven Eigenmodes'' in present-scale experiments, and highlight the need for core fluctuation and fast ion measurements in a future burning-plasma experiment.

  2. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    International Nuclear Information System (INIS)

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in vii DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  3. Study of electron temperature evolution during sawtoothing and pellet injection using thermal electron cyclotron emission in the Alcator C tokamak

    International Nuclear Information System (INIS)

    A study of the electron temperature evolution has been performed using thermal electron cyclotron emission. A six channel far infrared polychromator was used to monitor the radiation eminating from six radial locations. The time resolution was <3 μs. Three events were studied, the sawtooth disruption, propagation of the sawtooth generated heatpulse and the electron temperature response to pellet injection. The sawtooth disruption in Alcator takes place in 20 to 50 μs, the energy mixing radius is approx. 8 cm or a/2. It is shown that this is inconsistent with single resonant surface Kadomtsev reconnection. Various forms of scalings for the sawtooth period and amplitude were compared. The electron heatpulse propagation has been used to estimate chi e(the electron thermal diffusivity). The fast temperature relaxation observed during pellet injection has also been studied. Electron temperature profile reconstructions have shown that the profile shape can recover to its pre-injection form in a time scale of 200 μs to 3 ms depending on pellet size

  4. Experimental studies of edge turbulence and confinement in Alcator C-Moda)

    Science.gov (United States)

    Cziegler, I.; Terry, J. L.; Hughes, J. W.; LaBombard, B.

    2010-05-01

    The steep gradient edge region and scrape-off-layer (SOL) on the low-field-side of Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] tokamak plasmas are studied using gas-puff-imaging diagnostics. In L-mode plasmas, the region extending ˜2 cm inside the magnetic separatrix has fluctuations showing a broad, turbulent spectrum, propagating in the electron diamagnetic drift direction, whereas features in the open field line region propagate in the ion diamagnetic drift direction. This structure is robust against toroidal field strength, poloidal null-point geometry, plasma current, and plasma density. Global parameter dependence of spectral and spatial structure of the turbulence inside the separatrix is explored and characterized, and both the intensity and spectral distributions are found to depend strongly on the plasma density normalized to the tokamak density limit. In H-mode discharges the fluctuations at and inside the magnetic separatrix show fundamentally different trends compared to L-mode, with the electron diamagnetic direction propagating turbulence greatly reduced in ELM-free [F. Wagner et al., Proceedings of the Thirteenth Conference on Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1982), Vol. I, p. 277], and completely dominated by the modelike structure of the quasicoherent mode in enhanced D-alpha regimes [A. E. Hubbard, R. L. Boivin, R. S. Granetz et al., Phys. Plasmas 8, 2033 (2001)], while the normalized SOL turbulence is largely unaffected.

  5. Fluctuating zonal flows in the I-mode regime in Alcator C-Moda)

    Science.gov (United States)

    Cziegler, I.; Diamond, P. H.; Fedorczak, N.; Manz, P.; Tynan, G. R.; Xu, M.; Churchill, R. M.; Hubbard, A. E.; Lipschultz, B.; Sierchio, J. M.; Terry, J. L.; Theiler, C.

    2013-05-01

    Velocity fields and density fluctuations of edge turbulence are studied in I-mode [F. Ryter et al., Plasma Phys. Controlled Fusion 40, 725 (1998)] plasmas of the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] tokamak, which are characterized by a strong thermal transport barrier in the edge while providing little or no barrier to the transport of both bulk and impurity particles. Although previous work showed no clear geodesic-acoustic modes (GAM) on C-Mod, using a newly implemented, gas-puff-imaging based time-delay-estimate velocity inference algorithm, GAM are now shown to be ubiquitous in all I-mode discharges examined to date, with the time histories of the GAM and the I-mode specific [D. Whyte et al., Nucl. Fusion 50, 105005 (2010)] Weakly Coherent Mode (WCM, f = 100-300 kHz, Δf/f≈0.5, and kθ≈1.3 cm-1) closely following each other through the entire duration of the regime. Thus, the I-mode presents an example of a plasma state in which zero frequency zonal flows and GAM continuously coexist. Using two-field (density-velocity and radial-poloidal velocity) bispectral methods, the GAM are shown to be coupled to the WCM and to be responsible for its broad frequency structure. The effective nonlinear growth rate of the GAM is estimated, and its comparison to the collisional damping rate seems to suggest a new view on I-mode threshold physics.

  6. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

    2010-10-01

    Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

  7. H-mode pedestal and threshold studies over an expanded operating space on Alcator C-Moda)

    Science.gov (United States)

    Hubbard, A. E.; Hughes, J. W.; Bespamyatnov, I. O.; Biewer, T.; Cziegler, I.; LaBombard, B.; Lin, Y.; McDermott, R.; Rice, J. E.; Rowan, W. L.; Snipes, J. A.; Terry, J. L.; Wolfe, S. M.; Wukitch, S.

    2007-05-01

    This paper reports on studies of the edge transport barrier and transition threshold of the high confinement (H) mode of operation on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)], over a wide range of toroidal field (2.6-7.86T) and plasma current (0.4-1.7MA). The H-mode power threshold and edge temperature at the transition increase with field. Barrier widths, pressure limits, and confinement are nearly independent of field at constant current, but the operational space at high B shifts toward higher temperature and lower density and collisionality. Experiments with reversed field and current show that scrape-off-layer flows in the high-field side depend primarily on configuration. In configurations with the B ×∇B drift away from the active X-point, these flows lead to more countercurrent core rotation, which apparently contributes to higher H-mode thresholds. In the unfavorable case, edge temperature thresholds are higher, and slow evolution of profiles indicates a reduction in thermal transport prior to the transition in particle confinement. Pedestal temperatures in this case are also higher than in the favorable configuration. Both high-field and reversed-field results suggest that parameters at the L-H transition are influencing the evolution and parameters of the H-mode pedestal.

  8. Evaluation of Optimized ICRF and LHRF Antennas in Alcator C-Mod

    International Nuclear Information System (INIS)

    Full text: Ion cyclotron range of frequency heating (ICRF) and lower hybrid range of frequency current drive (LHCD) are expected to be key heating and current drive actuators for future fusion reactors and devices. However, impurity contamination associated with ICRF antenna operation remains a major challenge, particularly in devices with metallic plasma facing components. For LHCD, maximizing coupled power to the plasma remains a challenge, particularly to maintain low reflection coefficient over range of plasma conditions. Here, we report on an experimental investigation to test whether a field aligned (FA) ICRF antenna can reduce the impurity contamination and SOL modification associated with antenna operation. We also report on results from a new limiter for the LH coupler designed to reduce reflection coefficients across a wider range of plasma conditions. The unique feature of the so-called FA-antenna is that the current straps and antenna box structure are perpendicular to the total magnetic field. This alignment allows integrated E|| (electric field along a magnetic field line) to be minimized through symmetry. Using finite element method and a cold plasma model, the FA-antenna has been found to have lower integrated E|| relative to the previous antenna geometry. Initial results indicate that the impurity contamination associated with the FA-antenna is lower relative to our standard ICRF antennas. Configured as a 2-strap antenna, the antenna has lower core impurity contamination and lower impurity source at the antenna at high power density (∼ 15 MW/m2). An array of core and boundary plasma diagnostics are presently being used to characterize the impurity behavior and impact on the SOL transport and SOL density profiles; the latest results will be presented. For LHCD, reflection coefficients are very sensitive to the local density and its profile in front of the LHCD coupler. Previously the local LH coupler protection limiter was fixed to the outer wall of the vacuum vessel. The new limiter is mounted on the coupler and protrudes 0.25 mm beyond the coupler for plasma heat flux protection. The protection tiles allow the LH launcher to be moved closer to the plasma than previously possible. Initial high power (Pnet ∼ 700 kW) results show lower reflection coefficients were achieved (Γ2 ∼ 0.1) as compared to the old configuration (Γ2 ∼ 0.2). (author)

  9. Investigation of the origin of neutrals in the main chamber of Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lipschultz, B. [Massachusetts Institute of Technology Plasma Science and Fusion Center Cambridge, MA (United States)]. E-mail: Blip@psfc.mit.edu; LaBombard, B. [Massachusetts Institute of Technology Plasma Science and Fusion Center Cambridge, MA (United States); Pitcher, C.S. [Plasma Surface Engineering Inc, Toronto (Canada); Boivin, R

    2002-06-01

    A series of experiments are described which are aimed at quantifying the relative contribution of divertor leakage and radial ion transport on neutral pressures surrounding the core plasma. Evidence is presented implying that cross-field transport competes with, or dominates, parallel transport in such a way that plasma exists far out in the scrape-off layer (SOL) in the shadow of limiters and recycles on main chamber surfaces. The transition from L- to H-mode core confinement does not affect the far SOL characteristics nor the scaling of pressures around the plasma. Variations in magnetic equilibrium are used to vary the divertor pressures independent of the core plasma. Based on the analysis of such experiments using a simple neutral flow model we estimate that neutrals escaping from leaks in the lower (closed) divertor during lower x-point operation contribute a smaller fraction ({approx}10-30%) of the midplane pressure than main chamber recycling. The inferred leakage is much larger from the upper (open) divertor during upper x-point operation. Most neutrals escaping from either divertor do not directly travel to the midplane. Instead, they are redirected, most likely by some combination of ionization and/or collisions (elastic, charge exchange). (author)

  10. Studies of EDA H-mode in Alcator C-Mod

    International Nuclear Information System (INIS)

    Studies of the enhanced Dα H-mode (EDA) have been extended to include ohmic plasmas. No clear difference in the EDA/ELMfree boundary or in other phenomenology are seen between ohmic and ICRF-heated plasmas, suggesting that neither the effect of ion tails nor direct RF/edge plasma interaction plays a role in EDA. Edge safety factor (q95) is the principal variable which determines which regime a discharge will be in. When q95 is greater than 4.0 for standard-shaped plasmas, the discharge is almost always EDA, while when it is less than 3.5, the plasma is almost always ELMfree. New edge diagnostics have allowed measurement of pedestal profiles with resolution of the order of 1 mm. Sudden changes in profile widths are not seen when the plasma makes a transition from EDA to ELMfree; however, the widths do vary with the same parameters that determine the EDA/ELMfree boundary. Strong edge-density fluctuations are observed to accompany EDA and may be responsible for the change in particle transport which is observed. The fluctuations have a quasi-coherent component whose frequency varies inversely with the pedestal width as measured by a visible continuum diagnostic. (author)

  11. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs.

  12. Full wave simulations of fast wave efficiency and power losses in the scrape-off layer of tokamak plasmas in mid/high harmonic and minority heating regimes

    Energy Technology Data Exchange (ETDEWEB)

    Bertelli, N.; Jaeger, E. F.; Hosea, J. C.; Phillips, C. K.; Berry, L.; Bonoli, P. T.; Gerhardt, S. P.; Green, D.; LeBlanc, B.; Perkins, R. J.; Qin, C. M.; Pinsker, R. I.; Prater, R.; Ryan, P. M.; Taylor, G.; Valeo, E. J.; Wilson, J. R.; Wright, J. C.; Zhang, X. J.

    2015-12-17

    Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves (HHFW), have found strong interaction between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 2D and 3D AORSA results for the National Spherical Torus eXperiment (NSTX) have shown a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is removed from in front of the antenna by increasing the edge density. Here, full wave simulations have been extended for 'conventional' tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results in HHFW regime show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for C-Mod and EAST, which operate in the minority heating regime.

  13. Investigation of lower hybrid physics through power modulation experiments on Alcator C-Moda)

    Science.gov (United States)

    Schmidt, A.; Bonoli, P. T.; Meneghini, O.; Parker, R. R.; Porkolab, M.; Shiraiwa, S.; Wallace, G.; Wright, J. C.; Harvey, R. W.; Wilson, J. R.

    2011-05-01

    Lower hybrid current drive (LHCD) is an attractive tool for off-axis current profile control in magnetically confined tokamak plasmas and burning plasmas (ITER), because of its high current drive efficiency. The LHCD system on Alcator C-Mod operates at 4.6 GHz, with ~ 1 MW of coupled power, and can produce a wide range of launched parallel refractive index (n||) spectra. A 32 chord, perpendicularly viewing hard x-ray camera has been used to measure the spatial and energy distribution of fast electrons generated by lower hybrid (LH) waves. Square-wave modulation of LH power on a time scale much faster than the current relaxation time does not significantly alter the poloidal magnetic field inside the plasma and thus allows for realistic modeling and consistent plasma conditions for different n|| spectra. Inverted hard x-ray profiles show clear changes in LH-driven fast electron location with differing n||. Boxcar binning of hard x-rays during LH power modulation allows for ~ 1 ms time resolution which is sufficient to resolve the build-up, steady-state, and slowing-down phases of fast electrons. Ray-tracing/Fokker-Planck modeling in combination with a synthetic hard x-ray diagnostic shows quantitative agreement with the x-ray data for high n|| cases. The time histories of hollow x-ray profiles have been used to measure off-axis fast electron transport in the outer half of the plasma, which is found to be small on a slowing down time scale.

  14. Nonaxisymmetric field effects on Alcator C-Moda)

    Science.gov (United States)

    Wolfe, S. M.; Hutchinson, I. H.; Granetz, R. S.; Rice, J.; Hubbard, A.; Lynn, A.; Phillips, P.; Hender, T. C.; Howell, D. F.; La Haye, R. J.; Scoville, J. T.

    2005-05-01

    A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n =1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B21/BT), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10-4. This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on "as-built" coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2/1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T.

  15. Edge energy transport barrier and turbulence in the I-mode regime on Alcator C-Moda)

    Science.gov (United States)

    Hubbard, A. E.; Whyte, D. G.; Churchill, R. M.; Cziegler, I.; Dominguez, A.; Golfinopoulos, T.; Hughes, J. W.; Rice, J. E.; Bespamyatnov, I.; Greenwald, M. J.; Howard, N.; Lipschultz, B.; Marmar, E. S.; Reinke, M. L.; Rowan, W. L.; Terry, J. L.

    2011-05-01

    We report extended studies of the I-mode regime [Whyte et al., Nucl. Fusion 50, 105005 (2010)] obtained in the Alcator C-Mod tokamak [Marmar et al., Fusion Sci. Technol. 51(3), 3261 (2007)]. This regime, usually accessed with unfavorable ion B × ∇B drift, features an edge thermal transport barrier without a strong particle transport barrier. Steady I-modes have now been obtained with favorable B × ∇B drift, by using specific plasma shapes, as well as with unfavorable drift over a wider range of shapes and plasma parameters. With favorable drift, power thresholds are close to the standard scaling for L-H transitions, while with unfavorable drift they are ˜ 1.5-3 times higher, increasing with Ip. Global energy confinement in both drift configurations is comparable to H-mode scalings, while density profiles and impurity confinement are close to those in L-mode. Transport analysis of the edge region shows a decrease in edge χeff, by typically a factor of 3, between L- and I-mode. The decrease correlates with a drop in mid-frequency fluctuations (f ˜ 50-150 kHz) observed on both density and magnetics diagnostics. Edge fluctuations at higher frequencies often increase above L-mode levels, peaking at f ˜ 250 kHz. This weakly coherent mode is clearest and has narrowest width (Δf/f ˜ 0.45) at low q95 and high Tped, up to 1 keV. The Er well in I-mode is intermediate between L- and H-mode and is dominated by the diamagnetic contribution in the impurity radial force balance, without the Vpol shear typical of H-modes.

  16. Performance Assessment of the C-Mod Multi-Spectral Line Polarization MSE (MSE-MSLP) Diagnostic

    Science.gov (United States)

    Scott, Steven; Mumgaard, Robert; Khoury, Matthew

    2015-11-01

    The accuracy of the Alcator C-Mod Motional Stark Effect (MSE) diagnostic is limited primarily by partially polarized background light that varies rapidly both in time (1 ms) and space - factor 10 variations are observed between adjacent spatial channels. ITER is likely to operate in a similar regime. Visible Bremsstrahlung, divertor molecular D2 emission, and glowing invessel structures generate unpolarized light that becomes partially polarized upon reflection. Because all three sources are broadband, the background light can be measured in real-time at wavelengths close to the MSE spectrum, thereby allowing the background to be interpolated in wavelength rather than in time. A 10-spatial-channel, 4-wavelength MSE-MSLP system has been developed using polarization polychromators that measure simultaneously the MSE pi- and sigma- lines as well as two nearby wavelengths that were chosen to avoid both the MSE spectrum and all known impurity lines on each sightline. Initial performance evaluation indicates that the background channel measurements faithfully track the background light in the pi- and sigma- lines. The improvement in accuracy of pitch-angle measurements and increased diagnostic flexibility over a wide range of plasma conditions will be reported. This work is supported by USDoE awards DE-FC02-99ER54512 and DE-AC02-09CH11466.

  17. Edge Transport Modeling using the 3D EMC3-Eirene code on Tokamaks and Stellarators

    Science.gov (United States)

    Lore, J. D.; Ahn, J. W.; Briesemeister, A.; Ferraro, N.; Labombard, B.; McLean, A.; Reinke, M.; Shafer, M.; Terry, J.

    2015-11-01

    The fluid plasma edge transport code EMC3-Eirene has been applied to aid data interpretation and understanding the results of experiments with 3D effects on several tokamaks. These include applied and intrinsic 3D magnetic fields, 3D plasma facing components, and toroidally and poloidally localized heat and particle sources. On Alcator C-Mod, a series of experiments explored the impact of toroidally and poloidally localized impurity gas injection on core confinement and asymmetries in the divertor fluxes, with the differences between the asymmetry in L-mode and H-mode qualitatively reproduced in the simulations due to changes in the impurity ionization in the private flux region. Modeling of NSTX experiments on the effect of 3D fields on detachment matched the trend of a higher density at which the detachment occurs when 3D fields are applied. On DIII-D, different magnetic field models were used in the simulation and compared against the 2D Thomson scattering diagnostic. In simulating each device different aspects of the code model are tested pointing to areas where the model must be further developed. The application to stellarator experiments will also be discussed. Work supported by U.S. DOE: DE-AC05-00OR22725, DE AC02-09CH11466, DE-FC02-99ER54512, and DE-FC02-04ER54698.

  18. Novel energy resolving x-ray pinhole camera on Alcator C-Moda)

    Science.gov (United States)

    Pablant, N. A.; Delgado-Aparicio, L.; Bitter, M.; Brandstetter, S.; Eikenberry, E.; Ellis, R.; Hill, K. W.; Hofer, P.; Schneebeli, M.

    2012-10-01

    A new energy resolving x-ray pinhole camera has been recently installed on Alcator C-Mod. This diagnostic is capable of 1D or 2D imaging with a spatial resolution of ≈1 cm, an energy resolution of ≈1 keV in the range of 3.5-15 keV and a maximum time resolution of 5 ms. A novel use of a Pilatus 2 hybrid-pixel x-ray detector [P. Kraft et al., J. Synchrotron Rad. 16, 368 (2009), 10.1107/S0909049509009911] is employed in which the lower energy threshold of individual pixels is adjusted, allowing regions of a single detector to be sensitive to different x-ray energy ranges. Development of this new detector calibration technique was done as a collaboration between PPPL and Dectris Ltd. The calibration procedure is described, and the energy resolution of the detector is characterized. Initial data from this installation on Alcator C-Mod is presented. This diagnostic provides line-integrated measurements of impurity emission which can be used to determine impurity concentrations as well as the electron energy distribution.

  19. Measurement and simulation of ICRF wave intensity with a recalibrated phase contrast imaging diagnostic on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Tsujii, N., E-mail: tsujii@k.u-tokyo.ac.jp [The University of Tokyo, Kashiwa (Japan); Porkolab, M.; Bonoli, P. T.; Edlund, E. M.; Ennever, P. C.; Lin, Y.; Wright, J. C.; Wukitch, S. J. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts (United States); Jaeger, E. F. [XCEL Engineering, Oak Ridge, Tennessee (United States); Green, D. L. [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States); Harvey, R. W. [CompX, Del Mar, California (United States)

    2015-12-10

    Waves in the ion cyclotron range of frequencies (ICRF) are one of the major tools to heat fusion plasmas. Full-wave simulations are essential to predict the wave propagation and absorption quantitatively, and it is important that these codes be validated against actual experimental measurements. In this work, the absolute intensity of the ICRF waves previously measured with a phase contrast imaging diagnostic was recalibrated and compared once more with full-wave predictions. In the earlier work, significant discrepancies were found between the measured and the simulated mode converted wave intensity [N. Tsujii et al., Phys. Plasmas 19, 082508]. With the new calibration of the detector array, the measured mode converted wave intensity is now in much better agreement with the full-wave predictions. The agreement is especially good for comparisons performed close to the antenna.

  20. Measurements of electron temperature profiles on Alcator C-Mod using a novel energy-resolving x-ray camera

    Science.gov (United States)

    Maddox, J.; Delgado, L.; Pablant, N.; Hill, K. W.; Bitter, M.; Efthimion, P.; Rice, J.

    2015-11-01

    The most common electron temperature diagnostics, Thomson Scattering (TS) and Electron Cyclotron Emission (ECE), both require large diagnostic footprints and expensive optics. Another electron temperature diagnostic is the Pulse-Height-Analysis (PHA) system, which derives the electron temperature from the x-ray bremsstrahlung continuum. However, the main disadvantage of the PHA method is poor temporal resolution of the Si(Li) diode detectors. This paper presents a novel x-ray pinhole camera, which uses a pixilated Pilatus detector that allows single photon counting at a rate 2MHz per pixel and the setting of energy thresholds. The detector configuration is optimized by Shannon-sampling theory, such that spatial profiles of the x-ray continuum intensity can be obtained simultaneously for different energies, in the range from 4 to 16 keV. The exponential-like dependence of the x-ray intensity with photon energies is compared with a model describing the Be filter, attenuation in air, and detector efficiency, as well as different sets of energy thresholds. Electron temperature measurements are compared with TS and ECE measurements. This work was supported by the US DOE Contract No.DE-AC02-09CH11466 and the DoE Summer Undergraduate Laboratory Internship (SULI) program.

  1. Measurement and simulation of ICRF wave intensity with a recalibrated phase contrast imaging diagnostic on Alcator C-Mod

    International Nuclear Information System (INIS)

    Waves in the ion cyclotron range of frequencies (ICRF) are one of the major tools to heat fusion plasmas. Full-wave simulations are essential to predict the wave propagation and absorption quantitatively, and it is important that these codes be validated against actual experimental measurements. In this work, the absolute intensity of the ICRF waves previously measured with a phase contrast imaging diagnostic was recalibrated and compared once more with full-wave predictions. In the earlier work, significant discrepancies were found between the measured and the simulated mode converted wave intensity [N. Tsujii et al., Phys. Plasmas 19, 082508]. With the new calibration of the detector array, the measured mode converted wave intensity is now in much better agreement with the full-wave predictions. The agreement is especially good for comparisons performed close to the antenna

  2. Heat-flux footprints for I-mode and EDA H-mode plasmas on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Terry, J.L., E-mail: terry@psfc.mit.edu [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); LaBombard, B.; Brunner, D.; Hughes, J.W.; Reinke, M.L.; Whyte, D.G. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States)

    2013-07-15

    IR thermography is used to measure the heat flux footprints on C-Mod’s outer target in I-mode and EDA H-mode plasmas. The footprint profiles are fit to a function with a simple physical interpretation. The fit parameter that is sensitive to the power decay length into the SOL, λ{sub SOL}, is ∼1–3× larger in I-modes than in H-modes at similar plasma current, which is the dominant dependence for the H-mode λ{sub SOL}. In contrast, the fit parameter sensitive to transport into the private-flux-zone along the divertor leg is somewhat smaller in I-mode than in H-mode, but otherwise displays no obvious dependence on I{sub p}, B{sub t}, or stored energy. A third measure of the footprint width, the “integral width”, is not significantly different between H- and I-modes. Also discussed are significant differences in the global power flows of the H-modes with “favorable”∇B drift direction and those of the I-modes with “unfavorable”∇B drift direction.

  3. Heat-flux footprints for I-mode and EDA H-mode plasmas on Alcator C-Mod

    International Nuclear Information System (INIS)

    IR thermography is used to measure the heat flux footprints on C-Mod’s outer target in I-mode and EDA H-mode plasmas. The footprint profiles are fit to a function with a simple physical interpretation. The fit parameter that is sensitive to the power decay length into the SOL, λSOL, is ∼1–3× larger in I-modes than in H-modes at similar plasma current, which is the dominant dependence for the H-mode λSOL. In contrast, the fit parameter sensitive to transport into the private-flux-zone along the divertor leg is somewhat smaller in I-mode than in H-mode, but otherwise displays no obvious dependence on Ip, Bt, or stored energy. A third measure of the footprint width, the “integral width”, is not significantly different between H- and I-modes. Also discussed are significant differences in the global power flows of the H-modes with “favorable”∇B drift direction and those of the I-modes with “unfavorable”∇B drift direction

  4. Characterization and performance of a field aligned ion cyclotron range of frequency antenna in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Wukitch, S. J.; Garrett, M. L.; Ochoukov, R.; Terry, J. L.; Hubbard, A.; Labombard, B.; Lau, C.; Lin, Y.; Lipschultz, B.; Miller, D.; Reinke, M. L.; Whyte, D. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Collaboration: Alcator C-Mod Team

    2013-05-15

    Ion cyclotron range of frequency (ICRF) heating is expected to provide auxiliary heating for ITER and future fusion reactors where high Z metallic plasma facing components (PFCs) are being considered. Impurity contamination linked to ICRF antenna operation remains a major challenge particularly for devices with high Z metallic PFCs. Here, we report on an experimental investigation to test whether a field aligned (FA) antenna can reduce impurity contamination and impurity sources. We compare the modification of the scrape of layer (SOL) plasma potential of the FA antenna to a conventional, toroidally aligned (TA) antenna, in order to explore the underlying physics governing impurity contamination linked to ICRF heating. The FA antenna is a 4-strap ICRF antenna where the current straps and antenna enclosure sides are perpendicular to the total magnetic field while the Faraday screen rods are parallel to the total magnetic field. In principle, alignment with respect to the total magnetic field minimizes integrated E|| (electric field along a magnetic field line) via symmetry. A finite element method RF antenna model coupled to a cold plasma model verifies that the integrated E|| should be reduced for all antenna phases. Monopole phasing in particular is expected to have the lowest integrated E||. Consistent with expectations, we observed that the impurity contamination and impurity source at the FA antenna are reduced compared to the TA antenna. In both L and H-mode discharges, the radiated power is 20%–30% lower for a FA-antenna heated discharge than a discharge heated with the TA-antennas. However, inconsistent with expectations, we observe RF induced plasma potentials (via gas-puff imaging and emissive probes to be nearly identical for FA and TA antennas when operated in dipole phasing). Moreover, the highest levels of RF-induced plasma potentials are observed using monopole phasing with the FA antenna. Thus, while impurity contamination and sources are indeed reduced with the FA antenna configuration, the mechanism determining the SOL plasma potential in the presence of ICRF and its impact on impurity contamination and sources remains to be understood.

  5. Characterization and performance of a field aligned ion cyclotron range of frequency antenna in Alcator C-Mod

    International Nuclear Information System (INIS)

    Ion cyclotron range of frequency (ICRF) heating is expected to provide auxiliary heating for ITER and future fusion reactors where high Z metallic plasma facing components (PFCs) are being considered. Impurity contamination linked to ICRF antenna operation remains a major challenge particularly for devices with high Z metallic PFCs. Here, we report on an experimental investigation to test whether a field aligned (FA) antenna can reduce impurity contamination and impurity sources. We compare the modification of the scrape of layer (SOL) plasma potential of the FA antenna to a conventional, toroidally aligned (TA) antenna, in order to explore the underlying physics governing impurity contamination linked to ICRF heating. The FA antenna is a 4-strap ICRF antenna where the current straps and antenna enclosure sides are perpendicular to the total magnetic field while the Faraday screen rods are parallel to the total magnetic field. In principle, alignment with respect to the total magnetic field minimizes integrated E|| (electric field along a magnetic field line) via symmetry. A finite element method RF antenna model coupled to a cold plasma model verifies that the integrated E|| should be reduced for all antenna phases. Monopole phasing in particular is expected to have the lowest integrated E||. Consistent with expectations, we observed that the impurity contamination and impurity source at the FA antenna are reduced compared to the TA antenna. In both L and H-mode discharges, the radiated power is 20%–30% lower for a FA-antenna heated discharge than a discharge heated with the TA-antennas. However, inconsistent with expectations, we observe RF induced plasma potentials (via gas-puff imaging and emissive probes to be nearly identical for FA and TA antennas when operated in dipole phasing). Moreover, the highest levels of RF-induced plasma potentials are observed using monopole phasing with the FA antenna. Thus, while impurity contamination and sources are indeed reduced with the FA antenna configuration, the mechanism determining the SOL plasma potential in the presence of ICRF and its impact on impurity contamination and sources remains to be understood

  6. Scrape-off layer reflectometer for Alcator C-Moda)

    Science.gov (United States)

    Lau, Cornwall; Hanson, Greg; Wilgen, John; Lin, Yijun; Wukitch, Steve

    2010-10-01

    A swept-frequency X-mode reflectometer is being built for Alcator C-Mod to measure the scrape-off layer density profiles at the top, middle, and bottom locations in front of both the new lower hybrid launcher and the new ion cyclotron range of frequencies antenna. The system is planned to operate between 100 and 146 GHz at sweep rates from 10 μs to 1 ms, and will cover a density range of approximately 1016-1020 m-3 at B0=5-5.4 T. To minimize the effects of density fluctuations, both differential phase and full phase reflectometry will be employed. Design, test data, and calibration results of this electronics system will be discussed. To reduce attenuation losses, tallguide (TE01) will be used for most of the transmission line system. Simulations of high mode conversion in tallguide components, such as e-plane hyperbolic secant radius of curvature bends, tapers, and horn antennas will be shown. Experimental measurements of the total attenuation losses of these components in the lower hybrid waveguide run will also be presented.

  7. Design of a correlation electron cyclotron emission diagnostic for Alcator C-Moda)

    Science.gov (United States)

    Sung, C.; White, A. E.; Irby, J. H.; Leccacorvi, R.; Vieira, R.; Oi, C. Y.; Peebles, W. A.; Nguyen, X.

    2012-10-01

    A correlation electron cyclotron emission (CECE) diagnostic has been installed in Alcator C-Mod. In order to measure electron temperature fluctuations, this diagnostic uses a spectral decorrelation technique. Constraints obtained with nonlinear gyrokinetic simulations guided the design of the optical system and receiver. The CECE diagnostic is designed to measure temperature fluctuations which have kθ ≤ 4.8 cm-1 (kθρs < 0.5) using a well-focused beam pattern. Because the CECE diagnostic is a dedicated turbulence diagnostic, the optical system is also flexible, which allows for various collimating lenses and antenna to be used. The system overview and the demonstration of its operability as designed are presented in this paper.

  8. First results of the SOL reflectometer on Alcator C-Moda)

    Science.gov (United States)

    Lau, C.; Hanson, G.; Lin, Y.; Wilgen, J.; Wukitch, S.; Labombard, B.; Wallace, G.

    2012-10-01

    A swept-frequency X-mode reflectometer has been built on Alcator C-Mod to measure the scrape-off layer (SOL) density profiles adjacent to the lower hybrid launcher. The reflectometer system operates between 100 and 146 GHz at sweep rates from 10 μs to 1 ms and covers a density range of ˜1016-1020 m-3 at B0 = 5-5.4 T. This paper discusses the analysis of reflectometer density profiles and presents first experimental results of SOL density profile modifications due to the application of lower hybrid range-of-frequencies power to L-mode discharges. Comparison between density profiles measured by the X-mode reflectometer and scanning Langmuir probes is also shown.

  9. Design of a microwave calorimeter for the microwave tokamak experiment

    Energy Technology Data Exchange (ETDEWEB)

    Marinak, M. (California Univ., Berkeley, CA (USA))

    1988-10-07

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs.

  10. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

  11. Microtearing modes and anomalous transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Drake, J.F.; Gladd, N.T.; Liu, C.S.; Chang, C.L.

    1980-04-14

    Microtearing (high-m) modes driven by the electron temperature gradient are found to be unstable for present tokamak parameters. A self-consistent calculation of the nonlinear saturation of this instability yields magnetic fluctuations vertical-barBvertical-bar/B approx. = rho/sub e//L/sub T/. The associated crossfield electron thermal conductivity is shown to be inversely proportional to density, consistent with Alcator scaling, and comparable in magnitude with that inferred from experiments.

  12. Structure and motion of edge turbulence in the National Spherical Torus Experiment and Alcator C-Moda)

    Science.gov (United States)

    Zweben, S. J.; Maqueda, R. J.; Terry, J. L.; Munsat, T.; Myra, J. R.; D'Ippolito, D.; Russell, D. A.; Krommes, J. A.; LeBlanc, B.; Stoltzfus-Dueck, T.; Stotler, D. P.; Williams, K. M.; Bush, C. E.; Maingi, R.; Grulke, O.; Sabbagh, S. A.; White, A. E.

    2006-05-01

    In this paper we compare the structure and motion of edge turbulence observed in L-mode vs. H-mode plasmas in the National Spherical Torus Experiment (NSTX) [M. Ono, M. G. Bell, R. E. Bell et al., Plasma Phys. Controlled Fusion 45, A335 (2003)]. The radial and poloidal correlation lengths are not significantly different between the L-mode and the H-mode in the cases examined. The poloidal velocity fluctuations are lower and the radial profiles of the poloidal turbulence velocity are somewhat flatter in the H-mode compared with the L-mode plasmas. These results are compared with similar measurements Alcator C-Mod [E. Marmar, B. Bai, R. L. Boivin et al., Nucl. Fusion 43, 1610 (2003)], and with theoretical models.

  13. Phase contrast imaging measurements of reversed shear Alfvén eigenmodes during sawteeth in Alcator C-Moda)

    Science.gov (United States)

    Edlund, E. M.; Porkolab, M.; Kramer, G. J.; Lin, L.; Lin, Y.; Wukitch, S. J.

    2009-05-01

    Reversed shear Alfvén eigenmodes (RSAEs) have been observed with the phase contrast imaging diagnostic and Mirnov coils during the sawtooth cycle in Alcator C-mod [M. Greenwald et al., Nucl. Fusion 45, S109 (2005)] plasmas with minority ion-cyclotron resonance heating. Both down-chirping RSAEs and up-chirping RSAEs have been observed during the sawtooth cycle. Experimental measurements of the spatial structure of the RSAEs are compared to theoretical models based on the code NOVA [C. Z. Cheng and M. S. Chance, J. Comput. Phys. 71, 124 (1987)] and used to derive constraints on the q profile. It is shown that the observed RSAEs can be understood by assuming a reversed shear q profile (up chirping) or a q profile with a local maximum (down chirping) with q ≈1.

  14. Modification of ordinary-mode reflectometry system to detect lower-hybrid waves in Alcator C-Moda)

    Science.gov (United States)

    Baek, S. G.; Shiraiwa, S.; Parker, R. R.; Dominguez, A.; Kramer, G. J.; Marmar, E. S.

    2012-10-01

    Backscattering experiments to detect lower-hybrid (LH) waves have been performed in Alcator C-Mod, using the two modified channels (60 GHz and 75 GHz) of an ordinary-mode reflectometry system with newly developed spectral recorders that can continuously monitor spectral power at a target frequency. The change in the baseline of the spectral recorder during the LH wave injection is highly correlated to the strength of the X-mode non-thermal electron cyclotron emission. In high density plasmas where an anomalous drop in the lower hybrid current drive efficiency is observed, the observed backscattered signals are expected to be generated near the last closed flux surface, demonstrating the presence of LH waves within the plasma. This experimental technique can be useful in identifying spatially localized LH electric fields in the periphery of high-density plasmas.

  15. Boronization of Russian tokamaks from carborane precursors

    International Nuclear Information System (INIS)

    A new and cheap boronization technique using the nontoxic and nonexplosive solid substance carborane has been developed and successfully applied to the Russian tokamaks T-11M, T-3M, T-10 and TUMAN-3. The glow discharge in a mixture of He and carborane vapor produced the amorphous B/C coating with the B/C ratio varied from 2.0-3.7. The deposition rate was about 150 nm/h. The primary effect of boronization was a significant reduction of the impurity influx and the plasma impurity contamination, a sharp decrease of the plasma radiated power, and a decrease of the effective charge. Boronization strongly suppressed the impurity influx caused by additional plasma heating. ECR- and ICR-heating as well as ECR current drive were more effective in boronized vessels. Boronization resulted in a significant extension of the Ne- and q-region of stable tokamak operation. The density limit rose strongly. In Ohmic H-mode energy confinement time increased significantly (by a factor of 2) after boronization. It rose linearly with plasma current Ip and was 10 times higher than Neo-Alcator time at maximum current. ((orig.))

  16. LPS impairs oxygen utilization in epithelia by triggering degradation of the mitochondrial enzyme Alcat1.

    Science.gov (United States)

    Zou, Chunbin; Synan, Matthew J; Li, Jin; Xiong, Sheng; Manni, Michelle L; Liu, Yuan; Chen, Bill B; Zhao, Yutong; Shiva, Sruti; Tyurina, Yulia Y; Jiang, Jianfei; Lee, Janet S; Das, Sudipta; Ray, Anuradha; Ray, Prabir; Kagan, Valerian E; Mallampalli, Rama K

    2016-01-01

    Cardiolipin (also known as PDL6) is an indispensable lipid required for mitochondrial respiration that is generated through de novo synthesis and remodeling. Here, the cardiolipin remodeling enzyme, acyl-CoA:lysocardiolipin-acyltransferase-1 (Alcat1; SwissProt ID, Q6UWP7) is destabilized in epithelia by lipopolysaccharide (LPS) impairing mitochondrial function. Exposure to LPS selectively decreased levels of carbon 20 (C20)-containing cardiolipin molecular species, whereas the content of C18 or C16 species was not significantly altered, consistent with decreased levels of Alcat1. Alcat1 is a labile protein that is lysosomally degraded by the ubiquitin E3 ligase Skp-Cullin-F-box containing the Fbxo28 subunit (SCF-Fbxo28) that targets Alcat1 for monoubiquitylation at residue K183. Interestingly, K183 is also an acetylation-acceptor site, and acetylation conferred stability to the enzyme. Histone deacetylase 2 (HDAC2) interacted with Alcat1, and expression of a plasmid encoding HDAC2 or treatment of cells with LPS deacetylated and destabilized Alcat1, whereas treatment of cells with a pan-HDAC inhibitor increased Alcat1 levels. Alcat1 degradation was partially abrogated in LPS-treated cells that had been silenced for HDAC2 or treated with MLN4924, an inhibitor of Cullin-RING E3 ubiquitin ligases. Thus, LPS increases HDAC2-mediated Alcat1 deacetylation and facilitates SCF-Fbxo28-mediated disposal of Alcat1, thus impairing mitochondrial integrity. PMID:26604221

  17. Observation of ion cyclotron range of frequencies mode conversion plasma flow drive on Alcator C-Moda)

    Science.gov (United States)

    Lin, Y.; Rice, J. E.; Wukitch, S. J.; Greenwald, M. J.; Hubbard, A. E.; Ince-Cushman, A.; Lin, L.; Marmar, E. S.; Porkolab, M.; Reinke, M. L.; Tsujii, N.; Wright, J. C.; Alcator C-Mod Team

    2009-05-01

    At modest H3e levels (n3He/ne˜8%-12%), in relatively low density D(H3e) plasmas, n¯e≤1.3×1020 m-3, heated with 50 MHz rf power at Bt0˜5.1 T, strong (up to 90 km/s) toroidal rotation (Vϕ) in the cocurrent direction has been observed by high-resolution x-ray spectroscopy on Alcator C-Mod. The change in central Vϕ scales with the applied rf power (≤30 km s-1 MW-1), and is generally at least a factor of 2 higher than the empirically determined intrinsic plasma rotation scaling. The rotation in the inner plasma (r /a≤0.3) responds to the rf power more quickly than that of the outer region (r /a≥0.7), and the rotation profile is broadly peaked for r /a≤0.5. Localized poloidal rotation (0.3≤r/a≤0.6) in the ion diamagnetic drift direction (˜2 km/s at 3 MW) is also observed, and similarly increases with rf power. Changing the toroidal phase of the antenna does not affect the rotation direction, and it only weakly affects the rotation magnitude. The mode converted ion cyclotron wave (MC ICW) has been detected by a phase contrast imaging system and the MC process is confirmed by two-dimensional full wave TORIC simulations. The simulations also show that the MC ICW is strongly damped on H3e ions in the vicinity of the MC layer, approximately on the same flux surfaces where the rf driven flow is observed. The flow shear in our experiment is marginally sufficient for plasma confinement enhancement based on the comparison of the E ×B shearing rate and gyrokinetic linear stability analysis.

  18. Transport-driven scrape-off layer flows and the x-point dependence of the L-H power threshold in Alcator C-Moda)

    Science.gov (United States)

    LaBombard, B.; Rice, J. E.; Hubbard, A. E.; Hughes, J. W.; Greenwald, M.; Granetz, R. S.; Irby, J. H.; Lin, Y.; Lipschultz, B.; Marmar, E. S.; Marr, K.; Mossessian, D.; Parker, R.; Rowan, W.; Smick, N.; Snipes, J. A.; Terry, J. L.; Wolfe, S. M.; Wukitch, S. J.

    2005-05-01

    Factor of ˜2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B ×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B ×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL flows are also found similar, suggesting a connection.

  19. High-performance finite-difference time-domain simulations of C-Mod and ITER RF antennas

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, Thomas G., E-mail: tgjenkins@txcorp.com; Smithe, David N., E-mail: smithe@txcorp.com [Tech-X Corporation, 5621 Arapahoe Avenue Suite A, Boulder, CO 80303 (United States)

    2015-12-10

    Finite-difference time-domain methods have, in recent years, developed powerful capabilities for modeling realistic ICRF behavior in fusion plasmas [1, 2, 3, 4]. When coupled with the power of modern high-performance computing platforms, such techniques allow the behavior of antenna near and far fields, and the flow of RF power, to be studied in realistic experimental scenarios at previously inaccessible levels of resolution. In this talk, we present results and 3D animations from high-performance FDTD simulations on the Titan Cray XK7 supercomputer, modeling both Alcator C-Mod’s field-aligned ICRF antenna and the ITER antenna module. Much of this work focuses on scans over edge density, and tailored edge density profiles, to study dispersion and the physics of slow wave excitation in the immediate vicinity of the antenna hardware and SOL. An understanding of the role of the lower-hybrid resonance in low-density scenarios is emerging, and possible implications of this for the NSTX launcher and power balance are also discussed. In addition, we discuss ongoing work centered on using these simulations to estimate sputtering and impurity production, as driven by the self-consistent sheath potentials at antenna surfaces.

  20. Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lepson, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jernigan, J. Garrett [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-02-05

    We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.

  1. Status of tokamak research

    International Nuclear Information System (INIS)

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  2. Tokamak Systems Code

    International Nuclear Information System (INIS)

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  3. Lower hybrid experiments at the 1 MW level on Alcator C: heating and current drive

    International Nuclear Information System (INIS)

    Lower hybrid current drive and heating experiments have been carried out on the Alcator C tokamak at power levels up to 1.15 MW in the density range 1.0 x 1013 less than or equal to anti n/sub e/(cm-3) less than or equal to 1.0 x 1014. By launching waves with 670, 900 or 1120 phasing of adjacent waveguides, maximum flat-top current drive efficiencies of eta = R(m) x n(1014cm-3)I(MA)/P(MW) = 0.12 at B = 10 tesla, and eta approx. = 0.08 at B = 8 tesla were obtained with molybdenum limiters. With graphite, or silicon-carbide coated graphite limiters the efficiencies were 30 to 40% lower. Current ramping experiments have also been carried out at densities up to anti n approx. = 6 x 1013cm-3. By phasing the adjacent waveguides at 1800, heating experiments were performed in the density range 8 x 1013 less than or equal to anti n(cm-3) less than or equal to 2 x 1014 in both hydrogen and deuterium plasmas. This range of densities corresponds to the electron Landau heating mode. Using molybdenum limiters, typical heating rates of the order of eta/sub H/ = delta Σ n/sub j/T/sub j//P/sub rf/ approx. = 10 eV/kW 1013cm-3 were obtained, whereas with the SiC coated graphite limiters heating rates up to eta/sub H/ approx. = 22 were achieved. Measurements of soft and hard x-rays indicate the presence of substantial electron tails in both the current drive and the electron heating regimes. A combined transport, ray-tracing and Fokker-Planck code is used to analyze and model both the heating and the current drive results

  4. Study of parametric instabilities during the Alcator C lower hybrid wave heating experiments

    International Nuclear Information System (INIS)

    Parametric excitation of ion-cyclotron quasi-modes (ω/sub R/ approx. = nω/sub ci/) and ion-sound quasi-modes (ω/sub R/ approx. = k/sub parallel to/v/sub ti/) during lower hybrid wave heating of tokamak plasmas have been studied in detail. Such instabilities may significantly modify the incident wavenumber spectrum near the plasma edge. Convective losses for these instabilities are high if well-defined resonance cones exist, but they are significantly reduced if the resonance cones spread and fill the plasma volume (or some region of it). These instabilities preferentially excite lower hybrid waves with larger values of n/sub parallel to/ than themselves possess, and the new waves tend to be absorbed near the outer layers of the plasma. Parametric instabilities during lower hybrid heating of Alcator C plasmas have been investigated using rf probes (to study tilde phi and tilde n/sub i/) and CO2 scattering technique (to study tilde n/sub e/). At lower densities (anti n/sub e/ less than or equal to 0.5 x 1014cm-3) where waves observed in the plasma interior using CO2 scattering appear to be localized, parametric decay is very weak. Both ion-sound and ion-cyclotron parametric decay processes have been observed at higher densities (anti n greater than or equal to 1.5 x 1014cm-3) where waves appear to be unlocalized. Finally, at still higher densities (anti n /sub e/ greater than or equal to 2 x 104cm-3) pump depletion has been observed. Above these densities heating and current drive efficiencies are expected to degrade significantly

  5. Numerical modelling of lower hybrid RF heating and current drive experiments in the Alcator C tokamak

    International Nuclear Information System (INIS)

    A simulation model is described for lower hybrid (LH) current drive, rampup, heating, and sawtooth stabilization. The model incorporates a one-dimensional radial transport code, parallel velocity Fokker-Planck calculation, and a toroidal ray tracing code. For steady LH current drive it is found that the RF current generation is accurately predicted by a fast electron confinement time of the form τL=τ0(±)γ3, with τ0(±)=3ms in the density range of 3x1019m-3 e 19m-3 (where ± distinguishes electrons moving parallel (antiparallel) to the current drive direction). Also in this range, the theoretically predicted wave absorption and experimentally measured electron temperatures and stored energy were found to be consistent with an electron thermal diffusivity whose magnitude is independent of ne. To reproduce the experimentally measured values of LH rampup efficiency at n-bare=3x1019m-3, it was necessary to take τ0(±)=3ms. For LH heating at densities of n-bare approx.= 1.4x1020m-3, the power lost due to collisional damping of the LH ray trajectories at the plasma periphery was found to be significant, because of higher edge densities. Studies of LHRF sawtooth stabilization experiments with RF current drive indicated the possibility of creating stable profiles of the safety factor, q, via the generation of positive RF current near the q=1 surface, thus producing a current 'pedestal'. (author). 41 refs, 13 figs, 3 tabs

  6. Turbulence at the transition to the high density H-mode in Wendelstein 7-AS plasmas

    DEFF Research Database (Denmark)

    Basse, N.P.; Zoletnik, S.; Baumel, S.;

    2003-01-01

    Recently a new improved confinement regime was found in the Wendelstein 7-AS (W7-AS) stellarator (Renner H. et al 1989 Plasma Phys. Control. Fusion 31 1579). The discovery of this high density high confinement mode (HDH-mode) was facilitated by the installation of divertor modules. In this paper......-mode (EDA H-mode) found in the Alcator C-Mod tokamak and the HDH-mode in W7-AS is carried out....

  7. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  8. Tokamak concept innovations

    International Nuclear Information System (INIS)

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  9. Development of a High Resolution X-Ray Imaging Crystal Spectrometer for Measurement of Ion-Temperature and Rotation-Velocity Profiles in Fusion Energy Research Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hill, K W; Broennimann, Ch; Eikenberry, E F; Ince-Cushman, A; Lee, S G; Rice, J E; Scott, S

    2008-01-29

    A new imaging high resolution x-ray crystal spectrometer (XCS) has been developed to measure continuous profiles of ion temperature and rotation velocity in fusion plasmas. Following proof-of-principle tests on the Alcator C-Mod tokamak and the NSTX spherical tokamak, and successful testing of a new silicon, pixilated detector with 1 MHz count rate capability per pixel, an imaging XCS is being designed to measure full profiles of Ti and vφ on C-Mod. The imaging XCS design has also been adopted for ITER. Ion-temperature uncertainty and minimum measurable rotation velocity are calculated for the C-Mod spectrometer. The affects of x-ray and uclear-radiation background on the measurement uncertainties are calculated to predict performance on ITER.

  10. Development of a High Resolution X-Ray Imaging Crystal Spectrometer for Measurement of Ion-Temperature and Rotation-Velocity Profiles in Fusion Energy Research Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hill, K W; Broennimann, Ch; Eikenberry, E F; Ince-Cushman, A; Lee, S G; Rice, J E; Scott, S

    2008-02-27

    A new imaging high resolution x-ray crystal spectrometer (XCS) has been developed to measure continuous profiles of ion temperature and rotation velocity in fusion plasmas. Following proof-of-principle tests on the Alcator C-Mod tokamak and the NSTX spherical tokamak, and successful testing of a new silicon, pixilated detector with 1MHz count rate capability per pixel, an imaging XCS is being designed to measure full profiles of Ti and vφ on C-Mod. The imaging XCS design has also been adopted for ITER. Ion-temperature uncertainty and minimum measurable rotation velocity are calculated for the C-Mod spectrometer. The affects of x-ray and nuclear-radiation background on the measurement uncertainties are calculated to predict performance on ITER.

  11. The effects of main-ion dilution on turbulence in low q95 C-Mod ohmic plasmas, and comparisons with nonlinear GYRO

    Science.gov (United States)

    Ennever, P.; Porkolab, M.; Candy, J.; Staebler, G.; Reinke, M. L.; Rice, J. E.; Rost, J. C.; Ernst, D.; Hughes, J.; Baek, S. G.

    2016-08-01

    Recent experiments on C-mod seeding nitrogen into ohmic plasmas with q95 = 3.4 found that the seeding greatly reduced long-wavelength (ITG-scale) turbulence. The long-wavelength turbulence that was reduced by the nitrogen seeding was localized to the region of r /a ≈0.85 , where the turbulence is well above marginal stability (as evidenced by Qi/QGB≫1 ). The nonlinear gyrokinetic code GYRO was used to simulate the expected turbulence in these plasmas, and the simulated turbulent density fluctuations and turbulent energy fluxes quantitatively agreed with the experimental measurements both before and after the nitrogen seeding. Unexpectedly, the intrinsic rotation of the plasma was also found to be affected by the nitrogen seeding, in a manner apparently unrelated to a change in the electron-ion collisionality that was proposed by other experiments.

  12. Full wave simulations of fast wave mode conversion and lower hybrid wave propagation in tokamaks

    DEFF Research Database (Denmark)

    Wright, J.C.; Bonoli, P.T.; Brambilla, M.;

    2004-01-01

    Fast wave (FW) studies of mode conversion (MC) processes at the ion-ion hybrid layer in toroidal plasmas must capture the disparate scales of the FW and mode converted ion Bernstein and ion cyclotron waves. Correct modeling of the MC layer requires resolving wavelengths on the order of k(perpendi......Fast wave (FW) studies of mode conversion (MC) processes at the ion-ion hybrid layer in toroidal plasmas must capture the disparate scales of the FW and mode converted ion Bernstein and ion cyclotron waves. Correct modeling of the MC layer requires resolving wavelengths on the order of k......(perpendicular to)rho(i)similar to1 which leads to a scaling of the maximum poloidal mode number, M-max, proportional to 1/rho(*) (rho(*)equivalent torho(i)/L). The computational resources needed scale with the number of radial (N-r), poloidal (N-theta), and toroidal (N-phi) elements as N-r * N-phi * N-theta(3...... time are capable of achieving the resolution and speed necessary to address mode conversion phenomena in full two-dimensional (2-D) toroidal geometry. These codes have been used in conjunction with theory and experimental data from the Alcator C-Mod [I. H. Hutchinson , Phys. Plasmas 1, 1511 (1994...

  13. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  14. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  15. Research using small tokamaks

    International Nuclear Information System (INIS)

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  16. Time-Dependent Distribution Functions in C-Mod Calculated with the CQL3D-Hybrid-FOW, AORSA Full-Wave, and DC Lorentz Codes

    Science.gov (United States)

    Harvey, R. W. (Bob); Petrov, Yu. V.; Jaeger, E. F.; Berry, L. A.; Bonoli, P. T.; Bader, A.

    2015-11-01

    A time-dependent simulation of C-Mod pulsed ICRF power is made calculating minority hydrogen ion distribution functions with the CQL3D-Hybrid-FOW finite-orbit-width Fokker-Planck code. ICRF fields are calculated with the AORSA full wave code, and RF diffusion coefficients are obtained from these fields using the DC Lorentz gyro-orbit code. Prior results with a zero-banana-width simulation using the CQL3D/AORSA/DC time-cycles showed a pronounced enhancement of the H distribution in the perpendicular velocity direction compared to results obtained from Stix's quasilinear theory, in general agreement with experiment. The present study compares the new FOW results, including relevant gyro-radius effects, to determine the importance of these effects on the the NPA synthetic diagnostic time-dependence. The new NPA results give increased agreement with experiment, particularly in the ramp-down time after the ICRF pulse. Funded, through subcontract with Massachusetts Institute of Technology, by USDOE sponsored SciDAC Center for Simulation of Wave-Plasma Interactions.

  17. Improved Controls for Fusion RF Systems. Final technical report

    International Nuclear Information System (INIS)

    We have addressed the specific requirements for the integrated systems controlling an array of klystrons used for Lower Hybrid Current Drive (LHCD). The immediate goal for our design was to modernize the transmitter protection system (TPS) for LHCD on the Alcator C-Mod tokamak at the MIT Plasma Science and Fusion Center (MIT-PSFC). Working with the Alcator C-Mod team, we have upgraded the design of these controls to retrofit for improvements in performance and safety, as well as to facilitate the upcoming expansion from 12 to 16 klystrons. The longer range goals to generalize the designs in such a way that they will be of benefit to other programs within the international fusion effort was met by designing a system which was flexible enough to address all the MIT system requirements, and modular enough to adapt to a large variety of other requirements with minimal reconfiguration

  18. Improved Controls for Fusion RF Systems. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Casey, Jeffrey A. [Rockfield Research Inc., Las Vegas, NV (United States)

    2011-11-08

    We have addressed the specific requirements for the integrated systems controlling an array of klystrons used for Lower Hybrid Current Drive (LHCD). The immediate goal for our design was to modernize the transmitter protection system (TPS) for LHCD on the Alcator C-Mod tokamak at the MIT Plasma Science and Fusion Center (MIT-PSFC). Working with the Alcator C-Mod team, we have upgraded the design of these controls to retrofit for improvements in performance and safety, as well as to facilitate the upcoming expansion from 12 to 16 klystrons. The longer range goals to generalize the designs in such a way that they will be of benefit to other programs within the international fusion effort was met by designing a system which was flexible enough to address all the MIT system requirements, and modular enough to adapt to a large variety of other requirements with minimal reconfiguration.

  19. Spectroscopic diagnostics of high temperature plasmas

    International Nuclear Information System (INIS)

    A three-year research program for the development of novel XUV spectroscopic diagnostics for magnetically confined fusion plasmas is proposed. The new diagnostic system will use layered synthetic microstructures (LSM) coated, flat and curved surfaces as dispersive elements in spectrometers and narrow band XUV filter arrays. In the framework of the proposed program we will develop impurity monitors for poloidal and toroidal resolved measurements on PBX-M and Alcator C-Mod, imaging XUV spectrometers for electron density and temperature fluctuation measurements in the hot plasma core in TEXT or other similar tokamaks and plasma imaging devices in soft x-ray light for impurity behavior studies during RF heating on Phaedrus T and carbon pellet ablation in Alcator C-Mod. Recent results related to use of multilayer in XUV plasma spectroscopy are presented. We also discuss the latest results reviewed to qo and local poloidal field measurements using Zeeman polarimetry

  20. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  1. Studies of ELM heat load, SOL flow and carbon erosion from existing tokamak experiments, and projections for ITER

    International Nuclear Information System (INIS)

    Three important physics issues for the ITER divertor design and operation are summarized based on the experimental and numerical work from multi-machine database (JET, JT-60U, ASDEX Upgrade, DIII-D, Alcator C-Mod and TEXTOR). (i) The energy load associated with Type-I ELMs is of great concern for the lifetime of the ITER divertor target. In order t o understand the physics base of the scaling models[1], the ELM heat and particle transport from the edge pedestal to the divertor is investigated. Convective transport during ELMs plays an important role in heat transport to the divertor. (ii) Determination of the SOL flow pattern and the driving mechanism has progressed experimentally and numerically. Influences of the drift effects on the SOL and divertor plasma transport were discussed. (iii) Carbon erosion and redeposition are of great importance in particular for tritium retention via codeposition. Characteristics of chemical yield at two different deposited carbon surfaces, i.e. erosion- and redeposition-dominated areas, have been studied. Progress in the understanding of the chemical erosion is reviewed. (author)

  2. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  3. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  4. Simulations of the L-H transition on experimental advanced superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Weiland, Jan [Department Applied Physics, Chalmers University of Technology and Euratom-VR Association, S41296 Gothenburg (Sweden)

    2014-12-15

    We have simulated the L-H transition on the EAST tokamak [Baonian Wan, EAST and HT-7 Teams, and International Collaborators, “Recent experiments in the EAST and HT-7 superconducting tokamaks,” Nucl. Fusion 49, 104011 (2009)] using a predictive transport code where ion and electron temperatures, electron density, and poloidal and toroidal momenta are simulated self consistently. This is, as far as we know, the first theory based simulation of an L-H transition including the whole radius and not making any assumptions about where the barrier should be formed. Another remarkable feature is that we get H-mode gradients in agreement with the α – α{sub d} diagram of Rogers et al. [Phys. Rev. Lett. 81, 4396 (1998)]. Then, the feedback loop emerging from the simulations means that the L-H power threshold increases with the temperature at the separatrix. This is a main feature of the C-mod experiments [Hubbard et al., Phys. Plasmas 14, 056109 (2007)]. This is also why the power threshold depends on the direction of the grad B drift in the scrape off layer and also why the power threshold increases with the magnetic field. A further significant general H-mode feature is that the density is much flatter in H-mode than in L-mode.

  5. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  6. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  7. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1998-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  8. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1996-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  9. Research using small tokamaks

    International Nuclear Information System (INIS)

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  10. Research using small tokamaks

    International Nuclear Information System (INIS)

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  11. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  12. Next tokamak facility

    International Nuclear Information System (INIS)

    Design studies on a superconducting, long-pulse, current-driven, ignited tokamak, called the Toroidal Fusion Core Demonstration (TFCD), are being conducted by the Fusion Engineering Design Center (FEDC) and Princeton Plasma Physics Laboratory (PPPL) with additional broad community involvement. Options include the use of all-superconducting toroidal field (TF) coils, a superconducting-copper hybrid arrangement of TF coils, or all-copper TF coils. Only the first two options have been considered to date. The general feasibility of these approaches has been established with the goal of high performance (ignition, approx. 390 MW; wall loading approx. 2.2 MW/m2) at minimum capital cost. The preconceptual effort will be completed in early FY 1984 and a selection made from the indicated options. The TFCD is judged to represent a reasonable necessary step between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  13. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the 3He from the D(D,n)3He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general)

  14. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  15. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  16. [High beta tokamak research

    International Nuclear Information System (INIS)

    Our activities on High Beta Tokamak Research during the past 20 months of the present grant period can be divided into six areas: reconstruction and modeling of high beta equilibria in HBT; measurement and analysis of MHD instabilities observed in HBT; measurements of impurity transport; diagnostic development on HBT; numerical parameterization of the second stability regime; and conceptual design and assembly of HBT-EP. Each of these is described in some detail in the sections of this progress report

  17. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    NARCIS (Netherlands)

    Box, F. M. A.; Howard, J.; VandeKolk, E.; Meijer, F. G.

    1997-01-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. S

  18. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  19. Magnetic confinement experiment -- 1: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1994-12-31

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization.

  20. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  1. Tokamak burn control

    International Nuclear Information System (INIS)

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  2. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System.

  3. D-D fusion neutron-spectra measurements and ion-temperature determination at Alcator C

    International Nuclear Information System (INIS)

    A neutron spectrometer system has been designed, assembled, and used to measure the D-D neutron spectrum at Alcator C. The design of the shielding and collimation was critical to the successful measurement of the spectrum and involved an integral approach in which the neutronics of the Alcator C was exploited to obtain a successful system. The system consists of a 3He ionization chamber mounted in a multi-component shield system. Essentially the outermost part of the shield and collimator has been designed to moderate the MeV-range neutrons to thermal energies. The inner part of the shield is designed to capture the thermalized neutrons with a minimum of gamma production. As a result, measurements during plasma discharges indicate that the ratio of the number of counts in the 2.45-MeV peak to the total number of neutron counts in the ion chamber is 1.67

  4. ITER tokamak device

    Science.gov (United States)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  5. Edge turbulence in tokamaks

    Science.gov (United States)

    Nedospasov, A. V.

    1992-12-01

    Edge turbulence is of decisive importance for the distribution of particle and energy fluxes to the walls of tokamaks. Despite the availability of extensive experimental data on the turbulence properties, its nature still remains a subject for discussion. This paper contains a review of the most recent theoretical and experimental studies in the field, including mainly the studies to which Wootton (A.J. Wooton, J. Nucl. Mater. 176 & 177 (1990) 77) referred to most in his review at PSI-9 and those published later. The available theoretical models of edge turbulence with volume dissipation due to collisions fail to fully interpret the entire combination of experimental facts. In the scrape-off layer of a tokamak the dissipation prevails due to the flow of current through potential shifts near the surface of limiters of divertor plates. The different origins of turbulence at the edge and in the core plasma due to such dissipation are discussed in this paper. Recent data on the electron temperature fluctuations enabled one to evaluate the electric probe measurements of turbulent flows of particles and heat critically. The latest data on the suppression of turbulence in the case of L-H transitions are given. In doing so, the possibility of exciting current instabilities in biasing experiments (rather than only to the suppression of existing turbulence) is given some attention. Possible objectives of further studies are also discussed.

  6. Dust Measurements in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  7. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  8. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Box, F.M.A.; Kolk, E. van de [Associatie Euratom-FOM, Nieuwegein (Netherlands). FOM-Instituut voor Plasmafysica; Howard, J. [Plasma Research Laboratory, Research School of Physical Science and Engineering, Australian National University, Canberra 0200 (Australia); Meijer, F.G. [Physics Faculty, University of Amsterdam, Amsterdam (Netherlands)

    1997-03-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. Several spectrometers, equipped with a charge-coupled device array, are being used with spectral ranges in the visible, the vacuum UV and the extreme UV. (orig.)

  9. Research using small tokamaks

    International Nuclear Information System (INIS)

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  10. Spheromak injection into a tokamak

    OpenAIRE

    Brown, M R; Bellan, P. M.

    1990-01-01

    Recent results from the Caltech spheromak injection experiment [to appear in Phys. Rev. Lett.] are reported. First, current drive by spheromak injection into the ENCORE tokamak as a result of the process of magnetic helicity injection is observed. An initial 30% increase in plasma current is observed followed by a drop by a factor of 3 because of sudden plasma cooling. Second, spheromak injection results in an increase of tokamak central density by a factor of 6. The high-current/high-density...

  11. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cTe/eB(δni/ni)rms which is also derived by a simple theory, the cross-field diffusion time, tp=a2/D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  12. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  13. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  14. Free-boundary toroidal Alfvén eigenmodes

    Science.gov (United States)

    Chen, Eugene Y.; Berk, H. L.; Breizman, B.; Zheng, L. J.

    2011-05-01

    A numerical study is presented for the n = 1 free-boundary toroidal Alfvén eigenmodes (TAE) in tokamaks, which shows that there is considerable sensitivity of n = 1 modes to the position of the conducting wall. An additional branch of the TAE is shown to emerge from the upper continuum as the ratio of conducting wall radius to plasma radius increases. Such phenomena arise in plasma equilibria with both circular and shaped cross sections, where the shaped profile studied here is similar to that found in Alcator C-Mod.

  15. Improved L-mode discharges using ion cyclotron resonance frequency heating on Tore Supra

    International Nuclear Information System (INIS)

    Recent results of ion cyclotron minority heating have been obtained with an improved L-mode confinement, close to ELMy H-mode prediction, at relatively high density. The confinement exceeds the standard L-mode by a factor up to 1.7. The improvement of the confinement is observed in both electron and ion channels with reduction of heat diffusivities. This improved confinement regime presents some features similar to the results previously observed in many tokamaks: ALCATOR C-MOD [1] with ICRH, and radiation improved confinement (RI) mode with neutral beam injection in TEXTOR [2], TFTR [3], DIII-D [4

  16. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  17. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D3He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  18. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  19. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  20. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  1. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  2. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  3. D-D neutron energy-spectra measurements in Alcator C

    International Nuclear Information System (INIS)

    Measurements of energy spectra of neutrons produced during high density (anti n/sub e/ > 2 x 1014 cm-3) deuterium discharges have been performed using a proton-recoil (NE 213) spectrometer. A two foot section of light pipe (coupling the scintillator and photomultiplier) was used to extend the scintillator into a diagnostic viewing port to maximize the neutron detection efficiency while not imposing excessive magnetic shielding requirements. A derivative unfolding technique was used to deduce the energy spectra. The results showed a well defined peak at 2.5 MeV which was consistent with earlier neutron flux measurements on Alcator C that indicated the neutrons were of thermonuclear origin

  4. Numerical modeling of a fast-neutron collimator for the Alcator A fusion device

    International Nuclear Information System (INIS)

    A numerical procedure is developed to analyze neutron collimators used for spatial neutron measurements of plasma neutrons. The procedure is based upon Monte-Carlo methods and uses a standard Monte-Carlo code. The specific developments described herein involve a new approach to represent complex spatial details in a method that is conservative of computer time, retains accuracy and required only modest changes in already-developed Monte-Carlo procedures. The procedure was used to model the Alcator A collimator. The collimator consists of 448 cells and has a measured spatial point source response of 0.7 cm. The numerical procedure successfully predicts this response

  5. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (ne and Te) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  6. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  7. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  8. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  9. Pioneering Structural Solutions for Compact High Field Experiments Developed for the Alcator and the Ignitor Programs

    Science.gov (United States)

    Salvetti, M.; Coppi, B.

    2015-11-01

    Recently there has been an increased awareness of the fact that the line of research based on compact high field machines is the most promising to approach ignition conditions in DT burning plasmas and has acquired new perspectives for its applications. Then the technological solutions that have made these machines possible have become subject to new attention and, in some cases, to rediscovery. The Alcator Program and, followed by Ignitor Program, has led to invent the coupled air-core former poloidal field system that has made compact machine possible and has been adopted on all advanced toroidal machines that came after Alcator. A recently rediscovered solution aimed at reducing the mechanical stresses in the inner legs of the toroidal magnet coils is the ``Upper and Lower Bracing Rings'' system that has had a key role in the design of the Ignitor machine and its evolution. Another solution to minimize the machine dimensions while maintaining high toroidal fields, in order to achieve high plasma current densities, is that of ``bucking and wedging'' of the toroidal magnet by coupling it mechanically to the central solenoid. Sponsored in part by the U.S. DoE.

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

    Science.gov (United States)

    Whyte, Dennis; ADX Team

    2015-11-01

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

  11. Time-dependent distribution functions and resulting synthetic NPA spectra in C-Mod calculated with the CQL3D-Hybrid-FOW, AORSA full-wave, and DC Lorentz codes

    Science.gov (United States)

    Harvey, R. W.; Petrov, Yu.; Jaeger, E. F.; Berry, L. A.; Bonoli, P. T.; Bader, A.

    2015-12-01

    A time-dependent simulation of C-Mod pulsed TCRF power is made obtaining minority hydrogen ion distributions with the CQL3D-Hybrid-FOW finite-orbit-width Fokker-Planck code. Cyclotron-resonant TCRF fields are calculated with the AORSA full wave code. The RF diffusion coefficients used in CQL3D are obtained with the DC Lorentz gyro-orbit code for perturbed particle trajectories in the combined equilibrium and TCRF electromagnetic fields. Prior results with a zero-banana-width simulation using the CQL3D/AORSA/DC time-cycles showed a pronounced enhancement of the H distribution in the perpendicular velocity direction compared to results obtained from Stix's quasilinear theory, and this substantially increased the rampup rate of the observed vertically-viewed neutral particle analyzer (NPA) flux, in general agreement with experiment. However, ramp down of the NPA flux after the pulse, remained long compared to the experiment. The present study compares the new FOW results, including relevant gyro-radius effects, to determine the importance of these new effects on the the NPA time-dependence.

  12. Assembly of Aditya upgrade tokamak

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. At present the existing ADITYA tokamak has been dismantled up to bottom plinth on which the whole assembly of toroidal field (TF) coils and vacuum vessel rested. The major components of ADITYA machine includes 20 TF coils and its structural components, 9 Ohmic coils and its clamps, 4 BV coils and its clamps as well as their busbar connections, vacuum vessel and its supports and buckling cylinder, which are all being dismantled. The re-assembly of the ADITYA Upgrade tokamak started with installation and positioning of new buckling cylinder and central solenoid (TR1) coil. After that the inner sections of TF coils are placed following which in-situ winding, installation, positioning and support mounting of two pairs of new inner divertor coils have been carried out. After securing the TF coils with top I-beams the new torus shaped vacuum vessel with circular cross-section in 2 halves have been installed. The assembly of TF structural components such as top and bottom guiding wedges, driving wedges, top and bottom compression ring, inner and outer fish plates and top inverted triangle has been carried out in an appropriate sequence. The assembly of outer sections of TF coils along with the proper placements of top auxiliary TR and vertical field coils with proper alignment and positioning with the optical metrology instrument mainly completes the reassembly. Detailed re-assembly steps and challenges faced during re-assembly will be discussed in this paper. (author)

  13. Transport of Dust Particles in Tokamak Devices

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  14. Bootstrap Current in Spherical Tokamaks

    Institute of Scientific and Technical Information of China (English)

    王中天; 王龙

    2003-01-01

    Variational principle for the neoclassical theory has been developed by including amomentum restoring term in the electron-electron collisional operator, which gives an additionalfree parameter maximizing the heat production rate. All transport coefficients are obtained in-cluding the bootstrap current. The essential feature of the study is that the aspect ratio affects thefunction of the electron-electron collision operator through a geometrical factor. When the aspectratio approaches to unity, the fraction of circulating particles goes to zero and the contribution toparticle flux from the electron-electron collision vanishes. The resulting diffusion coefficient is inrough agreement with Hazeltine. When the aspect ratio approaches to infinity, the results are inagreement with Rosenbluth. The formalism gives the two extreme cases a connection. The theoryis particularly important for the calculation of bootstrap current in spherical tokamaks and thepresent tokamaks, in which the square root of the inverse aspect ratio, in general, is not small.

  15. Options for an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon ..beta../sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed.

  16. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

    The low aspect ration tokamak is proposed for the driver of a transmutation reactor.The main parameters of the reactor core,neutronic analysis of the blanket are given>the neutron wall loading can be lowered from the magnitude order of 1 MW/m2 to 0.5MW/m2 which is much easier to reach in the near future,and the transmutation efficiency (fission/absorption ratio)is raised further.The blanket power density is about 200MW/m3 which is not difficult to deal with.The key components such as diverter and center conductor post are also designed and compared with conventional TOkamak,Finally,by comparison with the other drivers such as FBR,PWR and accelerator,it can be anticipated that the low aspect ratio transmutation reactor would be one way of fusion energy applications in the near future.

  17. Equilibrium Reconstruction in EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QIAN Jinping; WAN Baonian; L. L. LAO; SHEN Biao; S. A. SABBAGH; SUN Youwen; LIU Dongmei; XIAO Singjia; REN Qilong; GONG Xianzu; LI Jiangang

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of toka-mak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier ex-pansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.

  18. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

  19. Magnetic confinement experiment. I: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 2.5x greater than ELM`ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices.

  20. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  1. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  2. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  3. The disruptive instability in Tokamak plasmas

    NARCIS (Netherlands)

    Salzedas, F.J.B.

    2001-01-01

    Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative te

  4. The role of limiter in Egyptor Tokamak

    CERN Document Server

    Ei-Sisi, A B

    2002-01-01

    In Egyptor Tokamak, the limiter is used for separation of the plasma from the vessel. In this work an overview of limiter types, and construction of limiter in Egyptor Tokamak is discussed. Also simulation results of the radial electron density distribution in case of limiter are presented. The results of the simulation are in agreement with the experimental and analytical results.

  5. Breakdown in the pretext tokamak

    International Nuclear Information System (INIS)

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges

  6. Atomic physics in tokamak plasmas

    International Nuclear Information System (INIS)

    Tokamak discharges produce hydrogen-isotope plasmas in a quasi-steady state, with radial electron temperature, Tsub(e)(r), and density nsub(e)(r), distribution usually centrally peaked, with typical values Tsub(e)(0) approx.= 1 - 3 keV, nsub(e)(r) approx.= 1014 cm-3. Besides hydrogen, the plasma contains small quantities of carbon, oxygen, various construction or wall-conditioning materials such as Fe, Cr, Ni, Ti, Zr, Mo, and perhaps elements added for special diagnostic purposes, e.g., Si, Sc, Al, or noble gases. These elements are spatially fairly homogeneously distributed, with the different ionization states occurring near radial locations where Tsub(e)(r) approx.= Esub(i), the ionization potential. Thus, spectroscopic measurements of various plasma properties, such as ion temperatures, plasma motions or oscillations, radial transport rates, etc. are automatically endowed with spatial resolution. Furthermore the emitted spectra, even of heavier elements such as Fe or Ni, are fairly simple because only the ground levels are appreciably populated under the prevailing plasma conditions. Identification of near-ground transitions, including particularly magnetic dipole and intercombination transitions of ions with ionization potentials in the several keV range, and determination of their collisional and radiative transition probabilities will be required for development of appropriate diagnostics of tokamak-type plasma approaching the prospective fusion reactor conditions. (orig.)

  7. Papers presented at the eleventh topical conference on high-temperature plasma diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This report contains the following eleven papers presented at the conference: Neutral Beam Diagnostics for Alcator C-Mod; A Study for the Installation of the TEXT HIBP on DIII-D; Time-domain Triple-probe Measurement of Edge Plasma Turbulence on TEXT-U; A Langmuir/Mach Probe Array for Edge Plasma Turbulence and Flow; Determination of Field Line Location and Safety Factor in TEXT-U; Hybrid ECE Imaging Array System for TEXT-U; First Results from the Phase Contrast Imaging System on TEXT-U; A Fast Tokamak Plasma Flux and Electron Density Reconstruction Technique; Time-series Analysis of Nonstationary Plasma Fluctuations Using Wavelet Transforms; Quantitative Modeling of 3-D Camera Views for Tokamak Divertors; and Variable-frequency Complex Demodulation Technique for Extracting Amplitude and Phase Information. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  8. Control of a burning tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Burmeister, R.E.; Mandrekas, J.; Stacey, W.M.

    1993-03-01

    This report is a review of the literature relevant to the control of the thermonuclear burn in a tokamak plasma. Some basic tokamak phenomena are reviewed, and then control by modulation of auxiliary heating and fueling is discussed. Other possible control methods such as magnetic ripple, plasma compression, and impurity injection as well as more recent proposed methods such as divertor biasing and L- to H-mode transition are also reviewed. The applications of modern control theory to the tokamak burn control problem are presented. The control results are summarized and areas of further research are identified.

  9. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    D Raju; R Jha; P K Kaw; S K Mattoo; Y C Saxena; Aditya Team

    2000-11-01

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.

  10. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  11. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  12. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  13. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  14. D-D tokamak reactor studies

    International Nuclear Information System (INIS)

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated

  15. Simulating Plasma Turbulence in Tokamaks

    CERN Document Server

    Kepner, J V; Decyk, V; Kepner, Jeremy; Parker, Scott; Decyk, Viktor

    1997-01-01

    A challenging and fundamental research problem is the better understanding and control of the turbulent transport of heat in present-day tokamak fusion experiments. Recent developments in numerical methods along with enormous gains in computing power have made large-scale simulations an important tool for improving our understanding of this phenomena. Simulating this highly non-linear behavior requires solving for the perturbations of the phase space distribution function in five dimensions. We use a particle-in-cell approach to solve the equations. The code has been parallelized for a variety of architectures (C90, CM-5, T3D) using a 1-D domain decomposition along the toroidal axis, for which the number of particles in each cell remains approximately constant. The quasi-uniform distribution of particles, which minimizes load imbalance, coupled with the relatively small movement of particles across cells, which minimizes communications, makes this problem ideally suited to massively parallel architectures. We...

  16. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

  17. The JT-60 tokamak machine

    International Nuclear Information System (INIS)

    JT-60 is a large tokamak experimental device under construction at JAERI with main device parameters of R=3.0m, a=0.95m, Bsub(t)=45kG, and Isub(p)=2.7Ma. Its basic aim is to produce and confine hydrogen plasmas of temperatures in a multi-keV range and of confinement times comparable to a second, and to study its plasma-physics properties as well as engineering problems associated with them. The JT-60 tokamak machine is mainly composed of a vacuum vessel, toroidal field (TF) coils, poloidal field (PF) coils, and support structures. The vacuum vessel is a high toroidal chamber with an egg-shaped crossection, consisting of sectorial rigid rings and parallel bellows made from Inconel 625. It is baked out at a maximum temperature up to 5000C. Several kinds of first walls made from molybdenum are bolt-jointed to the vacuum vessel for its protection. The vacuum vessel is almost completely finished with design and is deeply into manufacturing. The TF system consists of 18 unit coils located around a torus axis at regular intervals. The unit coil composed of two pancakes are wedge-shaped at the section close to a torus axis and encased in a high-manganese non-magnetic steel case. Fabrication of the TF coils will be finished in May 1981. The PF coils are composed of ohmic heating coils, vertical field coils, horizontal field coils, and quadrupole field coils located inside the TF coil bore and outside the vacuum vessel, and magnetic limiter coils placed in the vacuum vessel. Its mechanical and thermal design is almost completed are composed of the upper and lower support structures, support comuns of the vacuum vessel, and central column made from high-manganese non-magnetic steel. The structural analysis was completed including a seismic analysis and the fabrication is now in progress. The first plasma is expected to be produced in October 1984. (orig.)

  18. Characteristics of Plasma Turbulence in the Mega Amp Spherical Tokamak

    CERN Document Server

    Ghim, Young-chul

    2013-01-01

    Turbulence is a major factor limiting the achievement of better tokamak performance as it enhances the transport of particles, momentum and heat which hinders the foremost objective of tokamaks. Hence, understanding and possibly being able to control turbulence in tokamaks is of paramount importance, not to mention our intellectual curiosity of it.

  19. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  20. Electron thermal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (108 K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called 'tokamak' this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high 'fusion' temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This 'anomalous transport' of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL)

  1. Simulation of burning tokamak plasmas

    International Nuclear Information System (INIS)

    To simulate dynamical behaviour of tokamak fusion reactors, a zero-dimensional time-dependent particle and power balance code has been developed. The zero-dimensional plasma model is based on particle and power balance equations that have been integrated over the plasma volume using prescribed profiles for plasma parameters. Therefore, the zero-dimensional model describes the global dynamics of a fusion reactor. The zero-dimensional model has been applied to study reactor start-up, and plasma responses to changes in the plasma confinement, fuelling rate, and impurity concentration, as well as to study burn control via fuelling modulation. Predictions from the zero-dimensional code have been compared with experimental data and with transport calculations of a higher dimensionality. In all cases, a good agreement was found. The advantage of the zero-dimensional code, as compared to higher-dimensional transport codes, is the possibility to quickly scan the interdependencies between reactor parameters. (88 refs., 58 figs., 6 tabs.)

  2. Microtearing modes in tokamak discharges

    Science.gov (United States)

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

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

    Science.gov (United States)

    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

  4. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

    The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

  5. Mass spectrometry instrumentation in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

    The mass spectrophotometry in the residual gases analysis in high vacuum systems, in particular in the Novillo Tokamak (TN), where pressures are required to be of the order 10-7 Torr, is carried out through an instrumental support with infrastructure configured in parallel to the experimental planning in this device. In the Novillo as well as other Tokamaks, it is necessary to condition the vacuum chamber for improving the main discharge parameters. At the present time, in this Tokamak the conditioning quality is presented determined by means of a mass spectrophotometer. A general instrumental description is presented associated with the Novillo conditioning, as well as the spectras obtained before and after operation. (Author)

  6. The Spherical Tokamak MEDUSA for Mexico

    Science.gov (United States)

    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 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.

  7. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  8. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  9. Electron cyclotron emission diagnostics on KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, S. H. [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Daejeon 305-353 (Korea, Republic of); Lee, K. D.; Kwon, M. [National Fusion Research Institute, 113 Gwahangno, Daejeon 305-333 (Korea, Republic of); Kogi, Y. [Fukuoka Institute of Technology, Higashiku, Fukuoka 811-0295 (Japan); Kawahata, K.; Nagayama, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Mase, A. [KASTEC, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  10. A method for tokamak neutronics calculations

    International Nuclear Information System (INIS)

    This paper presents a new method for neutron transport calculation in tokamak fusion reactors. The computational procedure is based on the solution of the even-parity transport equation in a toroidal geometry. The angular neutron distribution is treated by even-parity spherical harmonic expansion, while the spatial dependence is approximated by using R-function finite elements that are defined for regions of arbitrary geometric shape. In order to test the method, calculation of a simplified tokamak model is carried out. The results are compared with the results from the literature and for the same order of accuracy a reduction of the number of spatial unknowns is shown. (author)

  11. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  12. Multichannel submillimeter interferometer for tokamak density measurements

    International Nuclear Information System (INIS)

    A two-channel, submillimeter (SMM) laser, electron-density interferometer has been operated successfully on the ISX tokamak. The interferometer is the first phase of a diagnostic system to measure the tokamak plasma current density using the Faraday rotation of the polarization vector of SMM laser beams. Deuterated formic acid lasers (lambda = 0.381 mm) have produced cw power of 10 mW. The interferometer has performed successfully for line-averaged electron densities as high as 8 x 1013 cm-3

  13. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

    The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs

  14. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration. PMID:21033954

  15. Electronic system of TBR tokamak device

    International Nuclear Information System (INIS)

    The electronics developed as a part of the TBR project, which involves the construction of a small tokamak at the Physics Institute of the University of Sao Paulo, is described. On the basis of tokamak parameter values, the electronics for the toroidal field, ohmic/heating and vertical field systems is presented, including capacitors bank, switches, triggering circuits and power supplies. A controlled power oscilator used in discharge cleaning and pre-ionization is also described. The performance of the system as a function of the desired plasma parameters is discussed. (Author)

  16. Spontaneous generation of rotation in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Parra Diaz, Felix [Oxford University

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  17. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  18. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M

    2003-10-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance.

  19. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Robert John [UCLA/retired

    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 reaching 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.

  20. Toroidal Alfven wave stability in ignited tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

  1. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  2. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  3. Radioactivity evaluation for the KSTAR tokamak

    International Nuclear Information System (INIS)

    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 1016 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 106 Bq kg-1 in the carbon graphite of a plasma-facing wall. (authors)

  4. Analysis of sawtooth relaxation oscillations in tokamaks

    International Nuclear Information System (INIS)

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated

  5. Tokamak Transport Studies Using Perturbation Analysis

    NARCIS (Netherlands)

    Cardozo, N. J. L.; Dehaas, J. C. M.; Hogeweij, G. M. D.; Orourke, J.; Sips, A.C.C.; Tubbing, B. J. D.

    1990-01-01

    Studies of the transport properties of tokamak plasmas using perturbation analysis are discussed. The focus is on experiments with not too large perturbations, such as sawtooth induced heat and density pulse propagation, power modulation and oscillatory gas-puff experiments. The approximations made

  6. Plasma physics and controlled nuclear fusion research 1994. V. 1. Proceedings of the fifteenth international conference

    International Nuclear Information System (INIS)

    This volume contains (i) the traditional Artsimovich Memorial Lecture; (ii) nine presentations giving an overview of toroidal confinement systems (TFTR, JT-60U, JET, DIII-D, TORE SUPRA, Alcator C-Mod, JFT-2M and T-10 tokamaks and the Wendelstein 7-AS stellarator), (iii) twenty-three presentations on core plasma physics (mostly on charged-particle transport and improved confinement regimes), (iv) eight presentations on plasma heating and current drive, (v) twelve presentations on divertors and edge physics, (vi) thirteen on concept optimization (shaping of magnetic field configuration, control of plasma profiles and of disruptions, a.o.), and (vii) six on helical systems (stellarators, including torsatron/heliotron). Refs, figs and tabs

  7. Intra-shot MSE Calibration Technique For LHCD Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Jinseok; Scott, Steve; Shiraiwa, Syun' ichi; Greenwald, Martin; Parker, Ronald; Wallace, Gregory

    2009-11-23

    The spurious drift in pitch angle of order several degrees measured by the Motional Stark Effect (MSE) diagnostic in the Alcator C-Mod tokamak1 over the course of an experimental run day has precluded direct utilization of independent absolute calibrations. Recently, the underlying cause of the drift has been identified as thermal stress-induced birefringence in a set of in-vessel lenses. The shot-to-shot drift can be avoided by using MSE to measure only the change in pitch angle between a reference phase and a phase of physical interest within a single plasma discharge. This intra-shot calibration technique has been applied to the Lower Hybrid Current Drive (LHCD) experiments and the measured current profiles qualitatively demonstrate several predictions of LHCD theory such as an inverse dependence of current drive efficiency on the parallel refractive index and the presence of off-axis current drive.

  8. Imaging with Spherically Bent Crystals or Reflectors

    Energy Technology Data Exchange (ETDEWEB)

    Bitter, M; Hill, K W; Scott, S; Ince-Cushman, A; Reinke, M; Podpaly, Y; Rice, J E; Beiersdorfer, P

    2010-06-01

    This paper consists of two parts: Part I describes the working principle of a recently developed x-ray imaging crystal spectrometer, where the astigmatism of spherically bent crystals is being used with advantage to record spatially resolved spectra of highly charged ions for Doppler measurements of the ion-temperature and toroidal plasmarotation- velocity profiles in tokamak plasmas. This type of spectrometer was thoroughly tested on NSTX and Alcator C-Mod, and its concept was recently adopted for the design of the ITER crystal spectrometers. Part II describes imaging schemes, where the astigmatism has been eliminated by the use of matched pairs of spherically bent crystals or reflectors. These imaging schemes are applicable over a wide range of the electromagnetic radiation, which includes microwaves, visible light, EUV radiation, and x-rays. Potential applications with EUV radiation and x-rays are the diagnosis of laserproduced plasmas, imaging of biological samples with synchrotron radiation, and lithography.

  9. Non-resonant destabilization of (1/1) internal kink mode by suprathermal electron pressure

    Energy Technology Data Exchange (ETDEWEB)

    Delgado-Aparicio, L.; Gates, D. A.; Gorelenkov, N.; Scott, S.; Bertelli, N.; Wilson, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Sugiyama, L. [MIT - Laboratory of Nuclear Science, Cambridge, Massachusetts 02139 (United States); Shiraiwa, S.; Irby, J.; Granetz, R.; Parker, R.; Baek, S. G.; Faust, I.; Wallace, G.; Mumgaard, R.; Gao, C.; Greenwald, M.; Hubbard, A.; Hughes, J.; Marmar, E. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); and others

    2015-05-15

    New experimental observations are reported on the structure and dynamics of short-lived periodic (1, 1) “fishbone”-like oscillations that appear during radio frequency heating and current-drive experiments in tokamak plasmas. For the first time, measurements can directly relate changes in the high energy electrons to the mode onset, saturation, and damping. In the relatively high collisionality of Alcator C-Mod with lower hybrid current drive, the instability appears to be destabilized by the non-resonant suprathermal electron pressure—rather than by wave-particle resonance, rotates toroidally with the plasma and grows independently of the (1, 1) sawtooth crash driven by the thermal plasma pressure.

  10. Characterization of the Tokamak Novillo in cleaning regime; Caracterizacion del Tokamak Novillo en regimen de limpieza

    Energy Technology Data Exchange (ETDEWEB)

    Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E

    1992-02-15

    In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip{sub t} like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I{sub (p)}t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

  11. First experiments on the TO-2 tokamak with a divertor

    International Nuclear Information System (INIS)

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  12. Banana orbits in elliptic tokamaks with hole currents

    Science.gov (United States)

    Martin, P.; Castro, E.; Puerta, J.

    2015-03-01

    Ware Pinch is a consequence of breaking of up-down symmetry due to the inductive electric field. This symmetry breaking happens, though up-down symmetry for magnetic surface is assumed. In previous work Ware Pinch and banana orbits were studied for tokamak magnetic surface with ellipticity and triangularity, but up-down symmetry. Hole currents appear in large tokamaks and their influence in Ware Pinch and banana orbits are now considered here for tokamaks magnetic surfaces with ellipticity and triangularity.

  13. Systems studies of high-field tokamak ignition experiments

    International Nuclear Information System (INIS)

    A study of the interaction between the physics of ignition and the engineering constraints in the design of compact, high-field tokamak ignition demonstration devices is presented. The studies investigate the effects the various electron and ion thermal diffusivities, which result from the many tokamak scaling laws, have on the design parameters of an ignition device and show the feasibility of building and igniting a compact tokamak (R<1m). The relevant machine technology is discussed

  14. Steady State Advanced Tokamak (SSAT): The mission and the machine

    International Nuclear Information System (INIS)

    Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

  15. Numerical studies of edge localized instabilities in tokamaks

    International Nuclear Information System (INIS)

    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

  16. KTM Tokamak operation scenarios software infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, V.; Baystrukov, K.; Golobkov, YU.; Ovchinnikov, A.; Meaentsev, A.; Merkulov, S.; Lee, A. [National Research Tomsk Polytechnic University, Tomsk (Russian Federation); Tazhibayeva, I.; Shapovalov, G. [National Nuclear Center (NNC), Kurchatov (Kazakhstan)

    2014-10-15

    One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.

  17. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1996-06-01

    A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).

  18. Boundary Plasma Turbulence Simulations for Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Xu, X; Umansky, M; Dudson, B; Snyder, P

    2008-05-15

    The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.

  19. The Spherical Tokamak MEDUSA for Costa Rica

    Science.gov (United States)

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

    2012-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R>=3.6, under design[2]) and also the ongoing activities in low temperature plasmas. Courses in plasma physics at undergraduate and post-graduate joint programme levels are regularly conducted. The scientific programme is intend to clarify several issues in relevant physics for conventional and mainly STs, including transport, heating and current drive via Alfv'en wave, and natural divertor STs with ergodic magnetic limiter[3,4]. [1] G.D.Garstka, PhD thesis, University of Wisconsin at Madison, 1997 [2] L.Barillas et al., Proc. 19^th Int. Conf. Nucl. Eng., Japan, 2011 [3] C.Ribeiro et al., IEEJ Trans. Electrical and Electronic Eng., 2012(accepted) [4] C.Ribeiro et al., Proc. 39^th EPS Conf. Contr. Fusion and Plasma Phys., Sweden, 2012

  20. Module description of TOKAMAK equilibrium code MEUDAS

    International Nuclear Information System (INIS)

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  1. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  2. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-01-01

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  3. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  4. EU Integrated Tokamak Modelling (ITM) Task Force

    Institute of Scientific and Technical Information of China (English)

    A Becoulet

    2007-01-01

    @@ At the end of 2003, the European Fusion Development Agreement (EFDA) structure set-up a long-term European task force (TF) in charge of "co-ordinating the development of a coherent set of validated simulation tools for the purpose of benchmarking on existing tokamak experiments, with the ultimate aim of providing a comprehensive simulation package for ITER plasmas" [http://www.efda-taskforce-itm.org/].

  5. Smaller coil systems for tokamak reactors

    International Nuclear Information System (INIS)

    Ripple reduction by ferro-magnetic iron shielding is used to reduce the size of the toroidal field coils down to 7.8 by 10.4 m bore for a commercial tokamak reactor design with plasma parameters similar to STARFIRE. For maximum effectiveness, it is found that the blocks of ferromagnetic iron shielding should have triangular cross section and should be placed as close to the plasma as possible

  6. Resistive interchange instability in reversed shear tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, Masaru; Nakamura, Yuji; Wakatani, Masahiro [Graduate School of Energy Science, Kyoto University, Uji, Kyoto (Japan)

    1999-04-01

    Resistive interchange modes become unstable due to the magnetic shear reversal in tokamaks. In the present paper, the parameter dependences, such as q (safety factor) profile and the magnetic surface shape are clarified for improving the stability, using the local stability criterion. It is shown that a significant reduction of the beta limit is obtained for the JT-60U reversed shear configuration with internal transport barrier, since the local pressure gradient increases. (author)

  7. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

    This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

  8. Self-Organized Stationary States of Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Jardin, S. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. [General Atomics, San Diego, CA (United States); Krebs, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Max-Plank-Institut fur Plasmaphysik, Garching, Germany

    2015-11-01

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  9. Self-Organized Stationary States of Tokamaks.

    Science.gov (United States)

    Jardin, S C; Ferraro, N; Krebs, I

    2015-11-20

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  10. Neoclassical transport in high [beta] tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high [beta] large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high [beta] large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low [beta] values by a factor ([var epsilon]/q[sup 2][beta])[sup [1/2

  11. SST and ADITYA tokamak research in India

    International Nuclear Information System (INIS)

    Steady state operation of tokamaks plays an important role in high temperature magnetically confined plasma research. Steady state Superconducting Tokamak (SST) programme in India deals with the development of various technologies in this direction. SST-1 machine has been engineered and is being fabricated at the Institute for Plasma Research. The objectives of the machine are to study physics of plasma processes under steady state condition and develop the technologies related to steady state operation. Various sub-systems are being prototyped and developed. SST-1 is a large aspect ratio machine with a major radius of 1.1 m and a plasma minor radius of 0.2 m with elongation of 1.7 to 1.9 and triangularity of 0.5 to 0.7. It has been designed for 1000 sec operation at 3 T toroidal magnetic eld. Neutral beam Injection and Radio frequency heating systems are being developed to heat the plasma. Lower hybrid Current Drive system would sustain 200 kA of plasma current during 1000 sec operation. ADITYA tokamak has been upgraded with new diagnostics and RF heating systems. Thomson Scattering and ECE diagnostics have been operated. 200 kW Ion Cyclotron Resonance Heating (ICRH) and 200 kW Electron Cyclotron Resonance Heating (ECRH) systems have been successfully commissioned. RF assisted initial breakdown experiments have been initiated with these systems. (author)

  12. ECH on the MTX [Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    The Microwave Tokamak Experiment (MTX) at LLNL is investigating the heating of high density Tokamak plasmas using an intense pulse FEL. Our first experiments, now beginning, will study the absorption and plasma heating of single FEL pulses (20 ns pulse length and peak power up to 2 GW) at a frequency of 140 GHz. A later phase of experiments also at 140 GHz will study FEL heating at 5 kHz rate for a pulse train up to 50 pulses (35 ns pulse length and peak power up to 4 GW). Future operations are planned at 250 GHz with an average power of 2 MW for a pulse train of 0.5 s. The microwave output of the FEL is transported quasi-optically to the tokamak through a window-less, evacuated pipe of 20 in. diameter, using a six mirror system. Computational modelling of the non-linear absorption for the MTX geometry predicts single-pass absorption of 40% at a density and temperature of 1.8 /times/ 1020m/sup /minus/3/ and 1 keV, respectively. To measure plasma microwave absorption and backscatter, diagnostics are available to measure forward and reflected power (parallel wire grid beam-splitter and mirror directional couplers) and power transmitted through the plasma (segmented calorimeter and waveguide detector). Other fast diagnostics include ECE, Thompson scattering, soft x-rays, and fast magnetic probes. 8 refs., 2 figs

  13. ADX - Advanced Divertor and RF Tokamak Experiment

    Science.gov (United States)

    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.

  14. SOL Width Scaling in the MAST Tokamak

    Science.gov (United States)

    Ahn, Joon-Wook; Counsell, Glenn; Connor, Jack; Kirk, Andrew

    2002-11-01

    Target heat loads are determined in large part by the upstream SOL heat flux width, Δ_h. Considerable effort has been made in the past to develop analytical and empirical scalings for Δh to allow reliable estimates to be made for the next-step device. The development of scalings for a large spherical tokamak (ST) such as MAST is particularly important both for development of the ST concept and for improving the robustness of scalings derived for conventional tokamaks. A first such scaling has been developed in MAST DND plasmas. The scaling was developed by flux-mapping data from the target Langmuir probe arrays to the mid-plane and fitting to key upstream parameters such as P_SOL, bar ne and q_95. In order to minimise the effects of co-linearity, dedicated campaigns were undertaken to explore the widest possible range of each parameter while keeping the remainder as fixed as possible. Initial results indicate a weak inverse dependence on P_SOL and approximately linear dependence on bar n_e. Scalings derived from consideration of theoretical edge transport models and integration with data from conventional devices is under way. The established scaling laws could be used for the extrapolations to the future machine such as Spherical Tokamak Power Plant (STPP). This work is jointly funded by Euratom and UK Department of Trade and Industry. J-W. Ahn would like to recognise the support of a grant from the British Foreign & Commonwealth Office.

  15. Relativistic runaway electrons in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Jaspers, R.E.

    1995-02-03

    Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP).

  16. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (-6 to 4.5 x 10-6 Ω-m, thus taking the Zef value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow

  17. Soft-X-Ray Tomography Diagnostic at the Rtp Tokamak

    NARCIS (Netherlands)

    Da Cruz, D. F.; Donne, A. J. H.

    1994-01-01

    An 80-channel soft x-ray tomography system has been constructed for diagnosing the RTP (Rijnhuizen Tokamak Project) tokamak plasma. Five pinhole cameras, each with arrays of 16 detectors are distributed more or less homogeneously around a poloidal plasma cross section. The cameras are positioned clo

  18. Commercial feasibility of fusion power based on the tokamak concept

    International Nuclear Information System (INIS)

    The impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants is determined. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  19. Fokker-Planck/Transport model for neutral beam driven tokamaks

    International Nuclear Information System (INIS)

    The application of nonlinear Fokker-Planck models to the study of beam-driven plasmas is briefly reviewed. This evolution of models has led to a Fokker-Planck/Transport (FPT) model for neutral-beam-driven Tokamaks, which is described in detail. The FPT code has been applied to the PLT, PDX, and TFTR Tokamaks, and some representative results are presented

  20. Recent progress on the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule

  1. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    Haverkort, J.W.

    2013-01-01

    One of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma rotation, primarily

  2. Experimental studies of tokamak plasma in IPP Prague

    International Nuclear Information System (INIS)

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR during recent years. At present, investigation is primarily aimed at the anomalous transport and plasma-wall interaction in the tokamak under conditions of combined OH/LHCD regimes. Moreover, some New diagnostic methods were also developed and certain improvements in the CASTOR performance were achieved. (author). 41 refs

  3. Role of the tokamak ISTTOK on the EURATOM fusion programme

    International Nuclear Information System (INIS)

    This paper describes the role of the tokamak ISTTOK on the development of the portuguese fusion research team, in the frame of the EURATOM Fusion Programme. Main tasks on education and training, control and data acquisition, diagnostics and tokamak physics are summarized. Work carried out on ISTTOK in collaboration with foreign teams is also reported. (author)

  4. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

    Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

  5. A simulation study of a controlled tokamak plasma

    Science.gov (United States)

    Fujii, N.; Niwa, Y.

    1980-03-01

    A tokamak circuit theory, including results of numerical simulation studies, is applied to a control system synthesized for a Joule heated tokamak plasma. The treatment is similar to that of Ogata and Ninomiya (1979) except that in this case a quadrupole field coil current is considered coexisting with image induced on a vacuum chamber.

  6. Design and construction of electronic components for a ''Novillo'' Tokamak

    International Nuclear Information System (INIS)

    The goal of this effort was to design, construct and make functional the electronic components for a ''Novillo'' Tokamak currently being experimentally investigated at the National Institute of Nuclear Research in Mexico. The problem was to develop programmable electronic switches capable of discharging high voltage kilowatt energies stored in capacitator banks onto the coils of the Tokamak. (author)

  7. Tokamak plasma self-organization-synergetics of magnetic trap plasmas

    NARCIS (Netherlands)

    Razumova, K. A.; Andreev, V. F.; Eliseev, L. G.; Kislov, A. Y.; La Haye, R. J.; Lysenko, S. E.; Melnikov, A. V.; Notkin, G. E.; Pavlov, Y. D.; Kantor, M. Y.

    2011-01-01

    Analysis of a wide range of experimental results in plasma magnetic confinement investigations shows that in most cases, plasmas are self-organized. In the tokamak case, it is realized in the self-consistent pressure profile, which permits the tokamak plasma to be macroscopically MHD stable. Existin

  8. Recent progress on the Compact Ignition Tokamak (CIT)

    Energy Technology Data Exchange (ETDEWEB)

    Ignat, D.W.

    1987-01-01

    This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule.

  9. Experimental data base of Tokamak KTM physical diagnostics

    International Nuclear Information System (INIS)

    The process of software creation of experimental data storage of Tokamak KTM physical diagnostics based on analysis of storage methods of operating Tokamaks data is considered. Task of specific kinds of information storage is solved; experimental data base that is thr part of system providing information analysis performance in the post-start period is developed.(author)

  10. Desirable engineering features of the next-generation tokamak device

    International Nuclear Information System (INIS)

    Recent scoping studies examined a series of superconducting, long-pulse Driven Current Tokamak (DCT) devices. One class of options is an ignited, D-T burning device designated DCT-8. It was concluded that the DCT-8 is a most attracttive engineering option to adequately bridge the gap between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  11. Atomic physics for fusion plasma spectroscopy; a soft x-ray study of molybdenum ions

    International Nuclear Information System (INIS)

    Understanding the radiative patterns of the ions of heavy atoms (Z approx-gt 18) is crucial to fusion experiments. The present thesis applies ab initio, relativistic calculations of atomic data to modeling the emission of molybdenum (Z = 42) ions in magnetically confined fusion plasmas. The models are compared to observations made in the Alcator C-Mod tokamak (Plasma Fusion Center, Massachusetts Institute of Technology), and the Frascati Tokamak Upgrade. Experimental confirmation of these models allows confidence in calculations of the total molybdenum concentration and quantitative estimates of the total power lost from the plasmas due to molybdenum line radiation. Charge states in the plasma core (Mo33+ to Mo29+) emit strong x-ray and XUV spectra which allow benchmarking of models for the spatial distribution of highly stripped molybdenum ions; the models only achieve agreement with observations when the rates of indirect ionization and recombination processes are included in the calculation of the charge state distribution of the central molybdenum ions. The total concentration of molybdenum in the core of the plasma is found, and the total power radiated from the plasma core is computed. Observations of line emission from more highly charged molybdenum ions (Mo36+ to Mo34+) are presented. open-quotes Bulkclose quotes molybdenum charge states (Mo25+ to Mo23+) emit complicated XUV spectra from a position in the plasma near C-Mod's half radius; spatial profiles of these ions' emission are analyzed. Models for the line-emission spectra of adjacent ions (Mo28+ to Mo26+) are offered, and the accuracy and limits of ab initio energy level calculations are discussed. open-quotes Edgeclose quotes charge states (Mo22+ to Mo15) extend to the last closed magnetic flux surface of the C-Mod plasma. The strongest features from these charge states are emitted in a narrow band from ∼70 Angstrom

  12. Tokamak Plasmas : Measurement of temperature fluctuations and anomalous transport in the SINP tokamak

    Indian Academy of Sciences (India)

    R Kumar; S K Saha

    2000-11-01

    Temperature fluctuations have been measured in the edge region of the SINP tokamak. We find that these fluctuations have a comparatively high level (30–40%) and a broad spectrum. The temperature fluctuations show a quite high coherence with density and potential fluctuations and contribute considerably to the anomalous particle flux.

  13. The Experiments of the small Spherical Tokamak Gutta

    International Nuclear Information System (INIS)

    GUTTA is a small spherical tokamak (R = 16cm, a = 8cm, Ip = 150kA) operating at the St. Petersburg State University since 2004 in the scope of the IAEA CRP ''Joint Research using Small Tokamaks''. Main scientific activities on GUTTA include development of new and improvement of existing mathematical models of plasma control, relevant for application on large tokamaks and ITER and verification of them on GUTTA; studies on the ECRH/EBW assisted breakdown and non-solenoid plasma formation in low aspect ratio tokamak; development of diagnostics; training and education of students.In this paper design properties of Gutta will be presented. Regimes of operation of the tokamak and plasma shape parameters are described and first results of the plasma formation and start-up studied will be discussed

  14. MHD stability of advanced tokamak scenarios

    International Nuclear Information System (INIS)

    Tokamak plasmas with a non-monotonic q-profile (current profile) and negative shear in the plasma centre have been associated with improved confinement and large pressure gradients in the region of negative shear. In JET, this regime, has been obtained with pellet injection (the PEP mode) and in DIII-D by ramping the plasma elongation. In JET, the phase of improved confinement is transient and usually ends in a collapse due to an MHD instability which leads to a redistribution of the current and a monotonic q-profile. The infernal mode, which is driven by a large pressure gradient in the region of low shear near the minimum in the q-profile, is the most likely candidate for the observed instability. To extend the transient phase to steady state, control of the shape of the current density profile is essential. The modelling of these advanced tokamak scenarios with a non-monotonic q-profile using non-inductive current drive of lower hybrid waves, fast waves, and neutral beams is discussed elsewhere. The aim is to find suitable initial states and to maintain MHD stability when the plasma β is built up. For this purpose, the robustness of the MHD stability of these configurations is studied with respect to changes in the position and in the depth of the minimum in q, and in the shape of the q and pressure profile. The classes of equilibria chosen for the analysis are based on the modelling of the current-drive schemes for advanced tokamak scenarios in JET. The toroidal ideal and resistive MHD stability code CASTOR is used for the stability calculations. (author) 7 refs., 4 figs

  15. Industry roles in the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    There are several distinguishing features of the Tokamak Physics Experiment (TPX) to be found in the TPX program and in the organizations for constructing and operating the machine. Programmatically, TPX addresses several issues critical to the viability of magnetic fusion power plants. Organizationally, it is a multi-institutional partnership to construct and operate the machine and carry out its program mission. An important part of the construction partnership is the integrated industrial responsibility for design, R ampersand D, and construction. The TPX physics design takes advantage of recent research on advanced tokamak operating modes achieved for time scales of the order of seconds that are consistent with continuous operation. This synergism of high performance (higher power density) modes with plasma current driven mostly by internal pressure (boot-strap effect) points toward tokamak power plants that will be cost-competitive and operate continuously. A large fraction of the project is subcontracted to industry. By policy, these contracts are at a high level in the project breakdown of work, giving contractors much of the overall responsibility for a given major system. That responsibility often includes design and R ampersand D in addition to the fabrication of the system in question. Each contract is managed through one of three national laboratories: PPPL, LLNL, and ORNL. Separate contracts for system integration and construction management round out the industry involvement in the project. This integrated, major responsibility attracts high-level corporate attention within each company, which are major corporations with long-standing interest in fusion. Through the contracts already established on the TPX project, a new standard for industry involvement in fusion has been set, and these industries will be well prepared for future fusion projects

  16. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

    Full Text Available Modeling of 3D electromagnetic phenomena in TOKAMAK with typically distributed main and additional coils is not an easy business. Evaluated must be not only distribution of the magnetic field, but also forces acting in particular coils. Use of differential methods (such as FDM or FEM for this purpose may be complicated because of geometrical incommensurability of particular subregions in the investigated area or problems with the boundary conditions. That is why integral formulation of the problem may sometimes be an advantages. The theoretical analysis is illustrated on an example processed by both methods, whose results are compared and discussed.

  17. Scaling studies of beam-heated tokamaks

    International Nuclear Information System (INIS)

    Parametric scaling of neutral beam-heated tokamaks is examined to determine the trade-off between beam energy and power. It is shown that over a wide range of plasma parameters and assumed transport properties, the center mean plasma temperature is a function of P/sub A/E/sub B//sup delta/, where E/sub B/ and P/sub A/ are the beam energy and power per unit area, respectively, and delta is a calculable constant of order unity

  18. Spherical tokamak research for fusion reactor

    International Nuclear Information System (INIS)

    Between ITER and the commercial fusion reactor, there are many technological problems to be solved such as cost, neutron and steady-state operation. In the conceptual design of VECTOR and Slim CS reactors it was shown that the key is 'low aspect ratio'. The spherical tokamak (ST) has been expected as the base for fusion reactors. In US, ST is considered as a non-superconducting reactor for use in the neutron irradiation facility. Conceptual design of the superconducting ST reactor is conducted in Japan and Korea independently. In the present article, the prospect of the ST reactor design is discussed. (author)

  19. Iron forbidden lines in tokamak discharges

    International Nuclear Information System (INIS)

    Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions

  20. Profile control for an alternative spherical tokamak

    International Nuclear Information System (INIS)

    Magnetically driven plasma guns that are inserted around a flux conserver at definite angular intervals are considered. The creation and the control of plasma channels are examined. By means of the hybrid model developed, both a system analysis of the Alternative Spherical Tokamak (AST) and relevant computational experiments have been carried out. In addition, by using the results obtained from the numerical scheme, the complex non-inductive current drive mechanisms of bootstrap and helicity injection in the AST system are discussed in detail. (author). 2 refs, 2 figs

  1. DIII-D Advanced Tokamak Research Overview

    International Nuclear Information System (INIS)

    This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously βNH of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues

  2. Application of MDSplus on EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QU Lianzheng; LUO Jiarong; LI lingling; ZHANG Mingxing; WANG Yong

    2007-01-01

    EAST is a fully superconducting Tokamak in China used for controlled fusion research. MDSplus, a special software package for fusion research, has been used successfully as a central repository for analysed data and PCS (Plasma Control System) data since the debugging experiment in the spring of 2006 . In this paper, the reasons for choosing MDSplus as the analysis database and the way to use it are presented in detail, along with the solution to the problem that part of the MDSplus library does not work in the multithread mode. The experiment showed that the data system based on MDSplus operated stably and it could provide a better performance especially for remote users.

  3. Bolometer measurement on HT-6B tokamak

    International Nuclear Information System (INIS)

    This paper discribes the structure, methods of calibration and measurement system of a metal foil resistor bolometer which is developed for measuring the radiation power of high temperature plasmas. The radiation loss and neutral flux loss in HT-6B tokamak have been measured by using the bolometer. The following results were obtained: (1) A large, nearly constant fraction (∼50%) of the input power was lost to the wall by radiation and energetic neutrals during the quasisteady phase of a normal discharges; (2) The power loss linearly increased with the discharge current Ip; (3) During disruption, most of the plasma energy was lost by radiation and neutrals

  4. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  5. Tokamak advanced pump limiter experiments and analysis

    International Nuclear Information System (INIS)

    Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment

  6. Magnetic diagnostics for the lithium tokamak experiment.

    Science.gov (United States)

    Berzak, L; Kaita, R; Kozub, T; Majeski, R; Zakharov, L

    2008-10-01

    The lithium tokamak experiment (LTX) is a spherical tokamak with R(0)=0.4 m, a=0.26 m, B(TF) approximately 3.4 kG, I(P) approximately 400 kA, and pulse length approximately 0.25 s. The focus of LTX is to investigate the novel low-recycling lithium wall operating regime for magnetically confined plasmas. This regime is reached by placing an in-vessel shell conformal to the plasma last closed flux surface. The shell is heated and then coated with liquid lithium. An extensive array of magnetic diagnostics is available to characterize the experiment, including 80 Mirnov coils (single and double axis, internal and external to the shell), 34 flux loops, 3 Rogowskii coils, and a diamagnetic loop. Diagnostics are specifically located to account for the presence of a secondary conducting surface and engineered to withstand both high temperatures and incidental contact with liquid lithium. The diagnostic set is therefore fabricated from robust materials with heat and lithium resistance and is designed for electrical isolation from the shell and to provide the data required for highly constrained equilibrium reconstructions. PMID:19044600

  7. Compact ignition tokamak physics and engineering basis

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is a high-field, compact tokamak design whose objective is the study of physics issues associated with burning plasmas. The toroidal and poloidal field coils employ a copper-steel laminate, manufactured by explosive-bonding techniques, to support the forces generated by the design fields: 10 T toroidal field at the plasma center; 21 T in the OH solenoid. A combination of internal and external PF coils provides control of the equilibrium and the ability to sweep the magnetic separatrix across the divertor plates during a pulse. At temperatures and βα levels characteristic of ITER designs, the fusion power in CIT approaches 800 MW and can be the limiting factor in the pulse length. Ignition requires that the confinement time exceed present L-mode scalings by about a factor of two, which is anticipated to occur as a result of the operational flexibility incorporated into the design. Conventional operating limits given by 20 e and qψ ≤ 3.2 have been chosen and, in the case of MHD limits, have been justified by ideal stability analysis. The power required for CIT ignition ranges from 10 MW to 40 MW or more, depending on confinement assumptions, and either ICRF or ECRF heating, or both, will be used. (author). 17 refs, 6 figs, 1 tab

  8. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

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

  9. Conceptual tokamak design at high neutron fluence

    International Nuclear Information System (INIS)

    For the future fusion reactor, it is important to design an experimental device that can be performed testing in-vessel components including tritium breeding modules relevant to the future fusion reactor with high neutron fluence. To realize this requirement, a conceptual tokamak design has been performed in accordance with plasma performance and shape at quasi-steady-state operation. One of the promising scenarios for this purpose is proposed to produce the plasma at the outward shifted radial position with a small minor radius for reasonable plasma parameters. From the analytical results, an appropriate space can be found for neutron shielding so that additional neutron shielding can be installed to protect the tokamak components from any neutron damages under the neutron fluence of 1 MWa m-2. Based on the structural analyses, a two-stage blanket module concept is proposed, i.e. one shielding block with the first wall assembly during high Q operation and two shielding blocks or additional tritium breeding modules during quasi-steady state operation

  10. Electromagnetic simulations of tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Michael; Mishchenko, Alexey [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, Wendelsteinstrasse 1, 17491 Greifswald (Germany)

    2014-07-01

    A practical fusion reactor will require a plasma β of around 5%. In this range Alfvenic effects become important. Since a practical reactor will also produce energetic alpha particles, the interaction between Alfvenic instabilities and fast ions is of particular interest. We have developed a fluid electron, kinetic ion hybrid model that can be used to study this problem. Compared to fully gyrokinetic electromagnetic codes, hybrid codes offer faster running times and greater flexibility, at the cost of reduced completeness. The model has been successfully verified against the worldwide ITPA Toroidal Alfven Eigenmode (TAE) benchmark, and the ideal MHD code CKA for the internal kink mode in a tokamak. Use of the model can now be turned toward cases of practical relevance. Current work focuses on simulating fishbones in a tokamak geometry, which may be of relevance to ITER, and producing the first non-perturbative self-consistent simulations of TAE in a stellarator, which may be of relevance both to Wendelstein 7-X and any future stellarator reactor. Preliminary results of these studies are presented.

  11. Ion cyclotron system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

  12. Ion cyclotron system design for KSTAR tokamak

    International Nuclear Information System (INIS)

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

  13. System studies of compact ignition tokamaks

    International Nuclear Information System (INIS)

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B2a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations

  14. Thomson scattering on the PRETEXT Tokamak

    International Nuclear Information System (INIS)

    Ruby laser Thomson scattering was performed on the PRETEXT tokamak. A 10 Joule Q-switched laser and a 1 meter 10 channel polychromator were used to diagnose the electron temperature and density profiles in the PRETEXT plasma. These parameters were measured as a function of time and radial position on a shot to shot basis. The density measurement was calibrated by Rayleigh and Raman scattering and by comparison with data from a 4 mm microwave interferometer. Electron densities ranging from 1 x 1012 cm-3 to 2 x 1013 cm-3 and temperatures ranging from 3 eV to 400 eV were observed. Detailed measurements were made throughout the 40 ms discharge with particular emphasis on the current rise phase. The Thomson scattering data was used as input to a one dimensional magnetic diffusion code. This code modelled the evolution of the current density and safety factor profiles. The results of this analysis were compared with existing theories of tokamak current penetration. The growth of resitive MHD tearing modes was proposed as a likely explanation for the anomalously rapid current penetration observed in PRETEXT

  15. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author)

  16. System studies of compact ignition tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Galambos, J.D.; Blackfield, D.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Selcow, E.

    1987-08-01

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B/sup 2/a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations.

  17. Alfven wave studies on a tokamak

    International Nuclear Information System (INIS)

    The continuum modes of the shear Alfven resonance are studied on the Tokapole II device, a small tokamak operated in a four node poloidal divertor configuration. A variety of antenna designs and the efficiency with which they deliver energy to the resonant layer are discussed. The spatial structure of the driven waves is studied by means of magnetic probes inserted into the current channel. In an attempt to optimize the coupling of energy in to the resonant layer, the angle of antenna currents with respect to the equilibrium field, antenna size, and plasma-to-antenna distance are varied. The usefulness of Faraday shields, particle shields, and local limiters are investigated. Antennas should be well shielded, either a dense Faraday shield or particle shield being satisfactory. The antenna should be large and very near to the plasma. The wave magnetic fields measured show a spatial resonance, the position of which varies with the value of the equilibrium field and mass density. They are polarized perpendicular to the equilibrium field. A wave propagates radially in to the resonant surface where it is converted to the shear Alfven wave. The signal has a short risetime and does not propagate far toroidally. These points are all consistent with a strongly damped shear Alfven wave. Comparisons of this work to theoretical predictions and results from other tokamaks are made

  18. Anomalous transport in the tokamak edge

    International Nuclear Information System (INIS)

    The tokamak edge has been studied with arrays of Langmuir and magnetic probes on the DITE and COMPASS-C devices. Measurements of plasma parameters such as density, temperature and radial magnetic field were taken in order to elucidate the character, effect on transport and origin of edge fluctuations. The tokamak edge is a strongly-turbulent environment, with large electrostatic fluctuation levels and broad spectra. The observations, including direct correlation measurements, are consistent with a picture in which the observed magnetic field fluctuations are driven by the perturbations in electrostatic parameters. The propagation characteristics of the turbulence, investigated using digital spectral techniques, appear to be dominated by the variation of the radial electric field, both in limiter and divertor plasmas. A shear layer is formed, associated in each case with the last closed flux surface. In the shear layer, the electrostatic wavenumber spectra are significantly broader. The predictions of a drift wave model (DDGDT) and of a family of models evolving from the rippling mode (RGDT group), are compared with experimental results. RGDT, augmented by impurity radiation effects, is shown to be the most reasonable candidate to explain the nature of the edge turbulence, only failing in its estimate of the wavenumber range. (Author)

  19. Equilibrium reconstruction in the START tokamak

    Science.gov (United States)

    Appel, L. C.; Bevir, M. K.; Walsh, M. J.

    2001-02-01

    The computation of magnetic equilibria in the START spherical tokamak is more difficult than those in more conventional large aspect ratio tokamaks. This difficulty arises partly as a result of the use of induction compression to generate high current plasma, as this precludes the positioning of magnetic diagnostics close to the outboard side of the plasma. In addition, the effect of a conducting wall with a high, but finite, conductivity must be included. A method is presented for obtaining plasma equilibrium reconstructions based on the EFIT code. New constraints are used to relate isoflux surface locations deduced from radial profile measurements of electron temperature. A model of flux diffusion through the vessel wall is developed. It is shown that neglecting flux diffusion in the vessel wall can lead to a significant underestimate in the calculation of the plasma βt. Using a relatively sparse set of magnetic signals, βt can be obtained to within a fractional error of +/-10%. Using constraints to relate isoflux surface locations, the principle involved in determining the internal q profile is demonstrated.

  20. Soft x-ray tomography on tokamaks using flux coordinates

    International Nuclear Information System (INIS)

    Methods of inverting line integrated data using coordinates, which are adapted to problems arising from Hamiltonian flows, are presented. They are exemplified for measurements of soft x-rays on tokamaks with widely arbitrary poloidal cross section. Boundary conditions can be met and cause fewer 'ghosts' for most of the present day tokamaks. The soft x-ray measurements are then used to improve the flux function Ψ as obtained from codes using magnetic measurements as input. We investigate oscillatory phenomena such as sawtooth crash precursors on tokamaks by decomposing the profiles into space-like eigenfunctions and their time dependencies. (author)

  1. Ions Measurement at the Edge of HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    Ling Bili; Wang Enyao; Gao wei; Wan Baonian; Li Jiangang

    2005-01-01

    A reliable method of measuring ions and ion temperature in tokamak plasma is necessary, for which an omegatron-like instrument has been developed on the HT-7 tokamak. The basic layout of the omegatron-like instrument is shown in this article. The measurement of working gas ion has been performed in the last experimental campaign on HT-7 tokamak. The relations among ion current, the electron repeller voltage and trap voltage have been investigated. This omegatron-like instrument has also provided the edge-plasma ion temperature.

  2. A systems assessment of the five Starlite tokamak power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C.G.

    1996-07-01

    The ARIES team has assessed the power-plant attractiveness of the following five tokamak physics regimes: (1) steady state, first stability regime; (2) pulsed, first stability regime; (3) steady state, second stability regime; (4) steady state, reversed shear; and (5) steady state, low aspect ratio. Cost-based systems analysis of these five tokamak physics regimes suggests that an electric power plant based upon a reversed-shear tokamak is significantly more economical than one based on any of the other four physics regimes. Details of this comparative systems analysis are described herein.

  3. Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

    Science.gov (United States)

    Zheng, Linjin; Horton, W.; Miura, H.; Shi, T. H.; Wang, H. Q.

    2016-08-01

    Owing to the large ion orbits, excessive electrons can accumulate at tokamak edge. We find that the nonneutralized electrons at tokamak edge can contribute an electric compressive stress in the direction parallel to magnetic field by their mutual repulsive force. By extending the Chew-Goldburger-Low theory (Chew et al., 1956 [13]), it is shown that this newly recognized compressive stress can significantly change the plasma average magnetic well, so that a stabilization of magnetohydrodynamic modes in the pedestal can result. This linear stability regime helps to explain why in certain parameter regimes the tokamak high confinement can be rather quiet as observed experimentally.

  4. Tokamak Plasmas : Internal magnetic field measurement in tokamak plasmas using a Zeeman polarimeter

    Indian Academy of Sciences (India)

    M Jagadeeshwari; J Govindarajan

    2000-11-01

    In a tokamak plasma, the poloidal magnetic field profile closely depends on the current density profile. We can deduce the internal magnetic field from the analysis of circular polarization of the spectral lines emitted by the plasma. The theory of the measurement and a detailed design of the Zeeman polarimeter constructed to measure the poloidal field profile in the ADITYA tokamak are presented. The Fabry-Perot which we have employed in our design, with photodiode arrays followed by lock-in detection of the polarization signal, allows the measurement of the fractional circular polarization. In this system He-II line with wavelength 4686 Å is adopted as the monitoring spectral line. The line emission used in the present measurement is not well localized in the plasma, necessiating the use of a spatial inversion procedure to obtain the local values of the field.

  5. Tokamak Plasmas : Observation of floating potential asymmetry in the edge plasma of the SINP tokamak

    Indian Academy of Sciences (India)

    Krishnendu Bhattacharyya; N R Ray

    2000-11-01

    Edge plasma properties in a tokamak is an interesting subject of study from the view point of confinement and stability of tokamak plasma. The edge plasma of SINP-tokamak has been investigated using specially designed Langmuir probes. We have observed a poloidal asymmetry of floating potentials, particularly the top-bottom floating potential differences are quite noticeable, which in turn produces a vertical electric field (v). This v remains throughout the discharge but changes its direction at certain point of time which seems to depend on applied vertical magnetic field v).

  6. Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR

    Energy Technology Data Exchange (ETDEWEB)

    Sa, Wanderley Pires de

    1996-12-31

    The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.

  7. Development of tokamak reactor system analysis code NEW-TORSAC

    Science.gov (United States)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  8. Engineering development aspects of the HL-2A tokamak

    International Nuclear Information System (INIS)

    The HL-2A tokamak (design values: major radius 1.65 m, minor radius 0.4 m, plasma current 0.48 MA and toroidal field 2.8 T) is the first tokamak with an operating divertor in China. It is characterized by a large closed divertor chamber. This unique feature will make significant contributions to enhance our understanding of complex divertor plasma physics and to help validating divertor physics modelings. The engineering design, development, testing and commissioning of the HL-2A tokamak are described in this paper. Preliminary results show that the HL-2A tokamak has been successfully operated in the divertor configuration. The major parameters: plasma current Ip=168 kA, toroidal field BT=1.4 T, plasma line average density ne=1.7 x 1019 m-3, limiting vacuum pv=4.6 x 10-6 Pa, were achieved at the end of 2003. (authors)

  9. HYFIRE: a tokamak- high-temperature electrolysis system

    International Nuclear Information System (INIS)

    Brookhaven National Laboratory is involved in a conceptual design study of a commercial nuclear power system which utilizes high-temperature electrolysis to produce synthetic fuels. The system is called HYFIRE. It includes a tokamak fusion power reactor supplying electrical and thermal energy to an array of electrolytes. The electrolytes produce hydrogen which can be used either directly as a fuel or in the production of hydrocarbons. The purpose of the study is to provide a mechanism for DOE to further assess the commercial potential of fusion using a tokamak reactor to produce synthetic fuel. The HYFIRE design is based on the tokamak commercial power reactor, STARFIRE. STARFIRE uses the deuterium/tritium/lithium fuel cycle. The HYFIRE study assumes the plasma shape and characteristics of STARFIRE study but uses a different blanket design. This study is particularly interested in the possibility of using the STARFIRE tokamak in the production of synthetic fuels

  10. Toroidicity Dependence of Tokamak Edge Safety Factor and Shear

    Institute of Scientific and Technical Information of China (English)

    SHIBingren

    2002-01-01

    In large tokamak device and reactor designs, the relationship between the toroidal current and the edge safety factor is very important because this will determine the eventual device or reactor size according to MHD stability requirements. In many preliminary

  11. Compact Ignition Tokamak Program: status of FEDC studies

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A.

    1985-01-01

    Viewgraphs on the Compact Ignition Tokamak Program comprise the report. The technical areas discussed are the mechanical configuration status, magnet analysis, stress analysis, cooling between burns, TF coil joint, and facility/device layout options. (WRF)

  12. Plasma X-ray emission in the 20-500 keV range during lower hybrid current drive on Alcator

    International Nuclear Information System (INIS)

    An array of eight 1'' x 3'' NaI scintillators has been used to collect plasma hard x-ray spectra (E/sub γ/>20 keV) emitted perpendicular to the magnetic axis during lower hybrid current drive on Alcator. The spectra exhibit a tail extending out to at least 300 keV and the profiles are generally peaked. These results show that the slope of the x-ray spectra increases with increasing plasma radius. Equivalently, the emission profiles tend to broaden with increasing photon energy. Also, the x-ray spectra slope increases at each radial location as the relative phasing of adjacent waveguides in the grill antenna is decreased. Preliminary results also suggest that the x-ray spectra tend to flatten and that the emission profiles tend to peak up with decreasing plasma density or increasing magnetic field. In addition, the initial results of an array for measuring the high energy x-ray emission from Alcator as a function of the emission angle relative to the magnetic axis are presented

  13. NEOCLASSICAL TRANSPORT IN A TOKAMAK WITH ELECTRIC SHEAR

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Neoclassical transport theory for a tokamak in the presence of a large radial electric field with shear is developed using Hamiltonian formalism. Diffusion coefficients are derived in both the plateau and banana regimes where the squeezing factor in coefficients can greatly affect diffusion at the plasma edge. Rotation speeds are calculated in the scrape-off region. They are in good agreement with the measurements on the TdeV tokamak.

  14. An emerging understanding of H-mode discharges in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, R.J.

    1992-12-01

    A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the [upsilon][sub E][sup [yields

  15. Technique for plasma filament stabilization in a tokamak

    International Nuclear Information System (INIS)

    The invention is related to the field of automatic control of thermonuclear device processes and can be used in control systems of plasma filament stabilization by large radius in tokamak type thermolnuclear devices. The economic effect of the suggested technique is caused by improvement of stabilization of optimum (from the viewpoint of the decrease of plasma energy losses) plasma filament position in the tokamak-reactor which results in the decrease of power of additional plasma heating systems

  16. Development of tokamak reactor systems analysis code 'TORSAC'

    International Nuclear Information System (INIS)

    This report describes Tokamak Reactor Systems Analysis Code ''TORSAC'' which has been developed in order to assess the impact of the design choises on reactor systems and to improve tokamak designs in wide parameter range. This computer code has following functions. (1) Systematic sensitivity analysis for a set of given design parameters, (2) Cost calculation of a new reactor concept designed automatically as a result of systematic sensitivity analysis. (author)

  17. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The

  18. Cooldown of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs

  19. Beam-induced tensor pressure tokamak equilibria

    International Nuclear Information System (INIS)

    D-shaped tensor pressure tokamak equilibria induced by neutral-beam injection are computed. The beam pressure components are evaluated from the moments of a distribution function that is a solution of the Fokker-Planck equation in which the pitch-angle scattering operator is ignored. The level-psub(perpendicular) contours undergo a significant shift away from the outer edge of the device with respect to the flux surfaces for perpendicular beam injection into broad-pressure-profile equilibria. The psub(parallel) contours undergo a somewhat smaller inward shift with respect to the flux surfaces for both parallel and perpendicular injection into broad-pressure-profile equilibria. For peaked-pressure-profile equilibria, the level pressure contours nearly co-incide with the flux surfaces. (author)

  20. Cooldown of the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Keeton, D.C.

    1987-08-01

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.

  1. Radiation power measurement on the ADITYA tokamak

    Science.gov (United States)

    Tahiliani, Kumudni; Jha, Ratneshwar; Gopalkrishana, M. V.; Doshi, Kalpesh; Rathod, Vipal; Hansalia, Chandresh; ADITYA Team

    2009-08-01

    The radiation power loss and its variation with plasma density and current are studied in the ADITYA tokamak. The radiation power loss varies from 20% to 40% of the input power for different discharges. The radiation fraction decreases with increasing plasma current but it increases with increasing line-averaged central density. The radiated power behavior has also been studied in discharges with short pulses of molecular beam injection (MBI) and gas puff (GP). The increase in radiation loss is limited to the edge chords in the case of GP, but it extends to the core region for MBI fueling. The MBI seems to indicate reduction in the edge recycling. It is observed that during the density limit disruption, the radiated power loss is more in the current quench phase as compared with the thermal quench phase and comes mainly from the plasma edge.

  2. Passive runaway electron suppression in tokamak disruptions

    International Nuclear Information System (INIS)

    Runaway electrons created in disruptions pose a serious problem for tokamaks with large current. It would be desirable to have a runaway electron suppression method which is passive, i.e., a method that does not rely on an uncertain disruption prediction system. One option is to let the large electric field inherent in the disruption drive helical currents in the wall. This would create ergodic regions in the plasma and increase the runaway losses. Whether these regions appear at a suitable time and place to affect the formation of the runaway beam depends on disruption parameters, such as electron temperature and density. We find that it is difficult to ergodize the central plasma before a beam of runaway current has formed. However, the ergodic outer region will make the Ohmic current profile contract, which can lead to instabilities that yield large runaway electron losses

  3. Ignition probabilities for Compact Ignition Tokamak designs

    International Nuclear Information System (INIS)

    A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs

  4. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  5. Sliding Mode Control of a Tokamak Transformer

    Energy Technology Data Exchange (ETDEWEB)

    Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

    2012-06-08

    A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

  6. Low Z impurity transport in tokamaks

    International Nuclear Information System (INIS)

    Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed

  7. Instrumentation and controls of an ignited tokamak

    International Nuclear Information System (INIS)

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented

  8. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  9. Safety factor profile control in a tokamak

    CERN Document Server

    Bribiesca Argomedo, Federico; Prieur, Christophe

    2014-01-01

    Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the  spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...

  10. The Procedure for Assembling the EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    Wu Songtao

    2005-01-01

    Due to the complicated constitution and high precision requirements of the EAST superconducting tokamak, a meticulous assembling procedure and measurement scheme must be established. The big size and mass of the EAST machine's components and complicated configuration with tight installation tolerances call for a highly careful assembling procedure. The assembling procedure consists of three main sub-procedures for the assembling of the base, of the tori of the VV, the vacuum vessel TS and the TF, and of the peripheral parts respectively. Before the assembly, a reference framework has been set up by means of an industrial measurement system with reference fiducial targets fixed on the wall of the test hall. In this paper, the assembling procedure is described in detail, the survey control system of the assembly is discussed, and progress in the assembly work is also reported.

  11. Natural Fueling of a Tokamak Fusion Reactor

    CERN Document Server

    Wan, Weigang; Chen, Yang; Perkins, Francis W

    2009-01-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

  12. Interactions of tokamak plasma with solid walls

    International Nuclear Information System (INIS)

    The interactions of tokamak fusion plasmas with solid walls of the devices were investigated on special model systems. The elastic recoil detection method was used for the determination of absolute hydrogen concentration. For the calibration of the method the scattering cross sections were measured in large ranges of scattering angle and energy. The erosion and deformation of wall surfaces were investigated by reemission of accelerated He ions. Theoretical models were developed to describe the surface undulation discovered earlier, caused by large dose He irradiation. The surface sputtering and segregation were investigated by nuclear methods and the mechanism of sputtering was simulated by computer. The surface deformation and gas reemission of Al surfaces were analyzed by Ar implementation and heat treatment. (D.Gy.) 6 figs

  13. Real time analysis of tokamak discharge parameters

    International Nuclear Information System (INIS)

    The techniques used in implementing two applications of real time analysis of data from the DIII-D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 msec intervals

  14. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

  15. Pumped limiter results on TFR Tokamak plasma

    International Nuclear Information System (INIS)

    Pump limiter experiments are carried out in the TFR Tokamak. The pump limiter is located in the outer part of the torus, its double- throat head is made of graphite tiles and it is pumped by a 2000 ls-1 titanium sublimation pump. The first attempts showed that the exhaust efficiency of this pump limiter was low (ε = 1.5% of the total plasma particle efflux). To improve these results, a new limiter head with a single longer throat has been built; particles were better trapped and the pumping provided an important decrease of the recycling coefficient. Geometric features mainly explain the increase by a factor 3.5 of the exhaust efficiency (ε = 5.5%). Ion temperature of the order of a few eV has been deduced from Doppler broadening measurements at the neutralizer plate of the pump limiter

  16. Next tokamak design. Swimming pool type

    International Nuclear Information System (INIS)

    In order to relieve the difficulties of repair and maintenance and to make the reactor size compact, a concept of swimming pool type reactor which is installed in a waterpool has been proposed. A design study of the concept as the Next Tokamak has been carried out with the following major parameters. The reactor has a double null poloidal divertor and blanket with tritium breeding ratio of >1.0, fusion power 420 MW, major radius 5.3 m, plasma radius 1.1 m, Bt on axis 5.2 T, plasma current 3.9 MA. The design study covers the reactor overall systems including reactor structure, reactor cooling system, repair and maintenance, reactor building, etc. As the result of this study the following conclusions were reached. The advantages over a conventional tokamak reactor are as follows: (1) The size of TF coil can be considerably reduced while retaining sufficient space for repair and maintenance because a solid shield is eliminated. (2) Since the distances between plasma and PF coils become small, the required capacity of electric power supply is reduced. (3) Technologies for the repair and maintenance are simplified and disassembling and reassembling of vacuum vessel can be done with realistic and credible remote handling technique. (4) The problem caused by radiation streaming can be considerably eased. (5) Radioactive waste disposal is reduced considerably because a solid shield is eliminated. (6) Because a vacuum vessel may be easily replaced in this concept, it will have a convenient flexibility for an experimental reactor. (7) Advantages of this concept can be also applied to a power reactor. Recently we started a new design of SPTR with slightly modified plasma parameters aiming for smaller-size reactor. In this paper the new design will be discussed briefly. (author)

  17. Ion cyclotron emission in tokamak plasmas

    International Nuclear Information System (INIS)

    Detection of α(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, α particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. α particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central α density in a reactor. (author)

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

    Science.gov (United States)

    Hubbard, A. E.; Osborne, T.; Ryter, F.; Austin, M.; Barrera Orte, L.; Churchill, R. M.; Cziegler, I.; Fenstermacher, M.; Fischer, R.; Gerhardt, S.; Groebner, R.; Gohil, P.; Happel, T.; Hughes, J. W.; Loarte, A.; Maingi, R.; Manz, P.; Marinoni, A.; Marmar, E. S.; McDermott, R. M.; McKee, G.; Rhodes, T. L.; Rice, J. E.; Schmitz, L.; Theiler, C.; Viezzer, E.; Walk, J. R.; White, A.; Whyte, D.; Wolfe, S.; Wolfrum, E.; Yan, Z.; Alcator C-Mod, the; Upgrade, ASDEX; DIII-D Teams

    2016-08-01

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

  19. Disruption mitigation using high pressure gas jets. Final report

    International Nuclear Information System (INIS)

    The goal of this research is to establish credible disruption mitigation scenarios based on the technique of massive gas injection. Disruption mitigation seeks to minimize or eliminate damage to internal components that can occur due to the rapid dissipation of thermal and magnetic energy during a tokamak disruption. In particular, the focus of present research is extrapolating mitigation techniques to burning plasma experiments such as ITER, where disruption-caused damage poses a serious threat to the lifetime of internal vessel components. A majority of effort has focused on national and international collaborative research with large tokamaks: DIII-D, Alcator C-Mod, JET, and ASDEX Upgrade. The research was oriented towards empirical trials of gas-jet mitigation on several tokamaks, with the goal of developing and applying cohesive models to the data across devices. Disruption mitigation using gas jet injection has proven to be a viable candidate for avoiding or minimizing damage to internal components in burning plasma experiments like ITER. The physics understanding is progress towards a technological design for the required gas injection system in ITER.

  20. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The

  1. Abstracts of the International seminar 'Experimental possibilities of KTM tokamak and research programme'

    International Nuclear Information System (INIS)

    The International seminar 'Experimental possibilities of KTM tokamak and research programme' was held in 10-12 October 2005 in Astana city (Kazakhstan). The seminar was dedicated to problems of KTM tokamak commissioning. The Collection of abstracts comprises 45 papers

  2. Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

    International Nuclear Information System (INIS)

    SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

  3. Basic Physics of Tokamak Transport Final Technical Report.

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Amiya K.

    2014-05-12

    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 difficult 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

  4. Importance of effects due to fusion α-particles for tokamak reactor design

    International Nuclear Information System (INIS)

    Issues related to the presence of fusion α-particles which are of importance for the design of a tokamak reactor are listed and shortly discussed. It is concluded that these issues, although to a large extent directly connected with the general problems of tokamak physics, require more attention to provide the information needed for designing a tokamak reactor. (orig.)

  5. Development of 3D ferromagnetic model of tokamak core with strong toroidal asymmetry

    DEFF Research Database (Denmark)

    Markovič, Tomáš; Gryaznevich, Mikhail; Ďuran, Ivan;

    2015-01-01

    Fully 3D model of strongly asymmetric tokamak core, based on boundary integral method approach (i.e. characterization of ferromagnet by its surface) is presented. The model is benchmarked on measurements on tokamak GOLEM, as well as compared to 2D axisymmetric core equivalent for this tokamak...

  6. Fast bolometric measurements on the TCV tokamak

    Science.gov (United States)

    Furno, I.; Weisen, H.; Mlynar, J.; Pitts, R. A.; Llobet, X.; Marmillod, Ph.; Pochon, G. P.

    1999-12-01

    The design and first results are presented from a bolometric diagnostic with high temporal resolution recently installed on the TCV tokamak. The system consists of two pinhole cameras viewing the plasma from above and below at the same toroidal location. Each camera is equipped with an AXUV-16ELO linear array of 16 p-n junction photodiodes, characterized by a flat spectral sensitivity from ultraviolet to x-ray energies, a high temporal response (<0.5 μs), and insensitivity to low-energy neutral particles emitted by the plasma. This high temporal resolution allows the study of transient phenomena such as fast magnetohydrodynamic (MHD) activity hitherto inaccessible with standard bolometry. In the case of purely electromagnetic radiation, good agreement has been found when comparing results from the new diagnostic with those from a standard metal foil bolometer system. This comparison has also revealed that the contribution of neutrals to the foil bolometer measurements can be extremely important under certain operating conditions, precluding the application of tomographic techniques for reconstruction of the radiation distribution.

  7. Demonstration tokamak-power-plant study (DEMO)

    International Nuclear Information System (INIS)

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li2O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li2O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B4C, lead, and FE 1422 structural material

  8. Aspects of Tokamak toroidal magnet protection

    International Nuclear Information System (INIS)

    Simple but conservative geometric models are used to estimate the potential for damage to a Tokamak reactor inner wall and blanket due to a toroidal magnet field collapse. The ofly potential hazard found to exist is due to the MHD pressure rise in a lithium blanket. A survey is made of proposed protection methods for superconducting torgidal magnets. It is found that the two general classificatigls of protectign methods are thermal and electrical. Computer programs were developed which aldow the toroidal magnet set to be modeled as a set of circular filaments. A simple thermal model of the conductor was used which allows heat transfer to the magnet structure and which includes the effect of temperature dependent properties. To be effective in large magnets an electrical protection system should remove at least 50% of the stored energy in the protection circuit assuming that all of the superconductor in the circuit quenches when the circuit is activated. A protection system design procedure based on this criterion was developed

  9. Plasma transport in a Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Nominal predicted plasma conditions in a Compact Ignition Tokamak (CIT) are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models that have given almost equally good fits to experimental data. Using a transport model that best fits the data, thermonuclear ignition occurs in a CIT design with a major radius of 1.32 m, plasma half-width of 0.43 mn, elongation of 2.0, and toroidal field and plasma current ramped in 6 s from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the /sup 3/He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates have a large effect on ignition and on the maximum beta that can be achieved

  10. Physics aspects of the compact ignition tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve nτE ≅ 2 x 1020 s/m3 required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which provides a high level of ohmic heating, improves the energy confinement, and allows a relatively high beta (≅ 6%). The present CIT design also has a high degree of elongation (κ ≅ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and pellet injection are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Auxiliary heating is expected to be necessary to achieve ignition, and 10-20 MW of ICRF is to be provided. (orig.)

  11. Experimental results from the TFTR tokamak

    International Nuclear Information System (INIS)

    Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam-heated discharges. The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2 T). Initial neutral-beam-heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 1020 m-3 in high current discharges. At the opposite extreme, by operating at low plasma current (I/sub p/ ∼ 0.8 MA) and low density (anti n/sub e/ ∼ 1 x 1019 m-3), high ion temperatures (9 +- 2 keV) and rotation speeds (7 x 105 m/s) have been achieved during injection. In addition, plasma compression experiments have demonstrated acceleration of beam ions from 82 keV to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam-heated discharges. The results of these confinement studies are presented

  12. Industry roles in the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) is the first major fusion project opportunity in many years for US industry. Both the TPX management and the Department of Energy's Office of Fusion Energy are committed to creating industry roles that are integrated throughout the project and that appropriately use the capabilities they offer. To address industry roles in TPX it is first appropriate to describe the collaborative national approach taken for this program. The Director of the Princeton Plasma Physics Laboratory (PPPL) was asked by DOE to set up this national team structure, and the current senior management positions and delegated responsibilities reflect that approach. While reporting lines and delegated roles are clear in the organization chart for TPX, one way to view, it, different from that of the individuals responsible upward through this management structure for various elements of the project, is through institutional responsibilities to the senior management team. In this view the management team relies on several national laboratories, each using industry contracts for major sub-systems and components, to execute the project. These responsibilities for design and for contracting are listed, showing that all major contracts will come through three national laboratories, forming teams for their responsible activities

  13. Aspects of Tokamak toroidal magnet protection

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.W.; Kazimi, M.S.

    1979-07-01

    Simple but conservative geometric models are used to estimate the potential for damage to a Tokamak reactor inner wall and blanket due to a toroidal magnet field collapse. The only potential hazard found to exist is due to the MHD pressure rise in a lithium blanket. A survey is made of proposed protection methods for superconducting toroidal magnets. It is found that the two general classifications of protection methods are thermal and electrical. Computer programs were developed which allow the toroidal magnet set to be modeled as a set of circular filaments. A simple thermal model of the conductor was used which allows heat transfer to the magnet structure and which includes the effect of temperature dependent properties. To be effective in large magnets an electrical protection system should remove at least 50% of the stored energy in the protection circuit assuming that all of the superconductor in the circuit quenches when the circuit is activated. A protection system design procedure based on this criterion was developed.

  14. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m2 and a particle heat flux of 1 MW/m2. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  15. Empirical particle transport model for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Petravic, M.; Kuo-Petravic, G.

    1986-08-01

    A simple empirical particle transport model has been constructed with the purpose of gaining insight into the L- to H-mode transition in tokamaks. The aim was to construct the simplest possible model which would reproduce the measured density profiles in the L-regime, and also produce a qualitatively correct transition to the H-regime without having to assume a completely different transport mode for the bulk of the plasma. Rather than using completely ad hoc constructions for the particle diffusion coefficient, we assume D = 1/5 chi/sub total/, where chi/sub total/ approx. = chi/sub e/ is the thermal diffusivity, and then use the kappa/sub e/ = n/sub e/chi/sub e/ values derived from experiments. The observed temperature profiles are then automatically reproduced, but nontrivially, the correct density profiles are also obtained, for realistic fueling rates and profiles. Our conclusion is that it is sufficient to reduce the transport coefficients within a few centimeters of the surface to produce the H-mode behavior. An additional simple assumption, concerning the particle mean-free path, leads to a convective transport term which reverses sign a few centimeters inside the surface, as required by the H-mode density profiles.

  16. Physics aspects of the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Post, D.; Bateman, G.; Houlberg, W.; Bromberg, L.; Cohn, D.; Colestock, P.; Hughes, M.; Ignat, D.; Izzo, R.; Jardin, S.

    1986-11-01

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha-particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve the condition of ntau/sub E/ approx. 2 x 10/sup 20/ sec m/sup -3/ required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k approx. 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided.

  17. Neoclassical transport of impurtities in tokamak plasmas

    International Nuclear Information System (INIS)

    Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I/2/n/sub H/e2 greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included

  18. Control of the vertical instability in tokamaks

    International Nuclear Information System (INIS)

    The problem of control of the vertical instability is formulated for a massless filamentary plasma. The massless approximation is justified by an examination of the role of inertia in the control problem. The system is solved using Laplace transform techniques. The linear system is studied to determine the stability boundaries. It is found that the system can be stabilized up to a critical decay index, which is predominantly a function of the geometry of the passive stabilizing shell. A second, smaller critical index, which is a function of the geometry of the control coils, determines the limit of stability in the absence of derivative gain in the control circuit. The system is also studied numerically in order to incorporate the non-linear effects of power supply dynamics. The power supply bandwidth requirement is determined by the open-loop growth rate of the instability. The system is studied for a number of control coil options which are available on the DIII-D tokamak. It is found that many of the coils will not provide adequate stabilization and that the use of inboard coils is advantageous in stabilizing the system up to the critical index. Experiments carried out on DIII-D confirm the appropriateness of the model. Using the results of the model study, we have stabilized DIII-D plasmas with decay indices up to 98% of the critical index. Measurement of the plasma vertical position is also discussed. (author) 27 figs., 6 refs

  19. Neoclassical transport in high {beta} tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high {beta} large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high {beta} large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low {beta} values by a factor ({var_epsilon}/q{sup 2}{beta}){sup {1/2}} II. This factor is the ratio of plasma volume in the boundary layer to the volume in the core. The fraction of trapped particles on a given flux surface (f{sub t}) is also reduced by this factor so that {approximately} {sub ({var_epsilon}}/q{sup 2}{beta}){sup {1/2}}. Special attention is given to the current equation, since this is thought to be relevant at low 3 and therefore may also be relevant at high {beta}. The bootstrap current term is found to exceed the actual current by a factor of the square root of the aspect ratio.

  20. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    Science.gov (United States)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  1. TSC (Tokamak Simulation Code) disruption scenarios and CIT (Compact Ignition Tokamak) vacuum vessel force evolution

    Energy Technology Data Exchange (ETDEWEB)

    Sayer, R.O.; Peng, Y.K.M.; Strickler, D.J.; Jardin, S.C.

    1990-01-01

    The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be F{sub R}={minus}12.0 MN/rad and F{sub Z}={minus}3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme F{sub R} by 15-50{percent} and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab.

  2. TSC [Tokamak Simulation Code] disruption scenarios and CIT [Compact Ignition Tokamak] vacuum vessel force evolution

    International Nuclear Information System (INIS)

    The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be FR=-12.0 MN/rad and FZ=-3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme FR by 15-50% and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab

  3. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  4. Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R

    2006-07-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission

  5. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

    Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)

    2013-10-15

    Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].

  6. Hybrid Method for Tokamak MHD Equilibrium Configuration Reconstruction

    Institute of Scientific and Technical Information of China (English)

    HE Hong-Da; DONG Jia-Qi; ZHANG Jin-Hua; JIANG Hai-Bin

    2007-01-01

    A hybrid method for tokamak MHD equilibrium configuration reconstruction is proposed and employed in the modified EFIT code. This method uses the free boundary tokamak equilibrium configuration reconstruction algorithm with one boundary point fixed. The results show that the position of the fixed point has explicit effects on the reconstructed divertor configurations. In particular, the separatrix of the reconstructed divertor configuration precisely passes the required position when the hybrid method is used in the reconstruction. The profiles of plasma parameters such as pressure and safety factor for reconstructed HL-2A tokamak configurations with the hybrid and the free boundary methods are compared. The possibility for applications of the method to swing the separatrix strike point on the divertor target plate is discussed.

  7. The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

    Directory of Open Access Journals (Sweden)

    Jing Li

    2014-01-01

    Full Text Available 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 Z2-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.

  8. Impact of the Pedestal on Global Performance and Confinement Scalings in I-mode

    Science.gov (United States)

    Walk, John; Hughes, Jerry; Hubbard, Amanda; Whyte, Dennis; White, Anne; Alcator C-Mod Team

    2015-11-01

    The I-mode is a novel high-confinement regime pioneered on Alcator C-Mod, notable for its strong temperature pedestal without the accompanying density pedestal found in conventional H-modes. This separation in transport channels gives the desired improved energy confinement while maintaining low particle confinement, avoiding excessive impurity accumulation. Moreover, I-mode operation is naturally free of deleterious Edge-Localized Modes (ELMs). Recent experiments on Alcator C-Mod have characterized the pedestal structure in I-mode. The impact of the pedestal response (particularly to fueling and heating power) and core profile stiffness on global performance and confinement have demonstrated confinement metrics competitive with H-mode operation on Alcator C-Mod, and consistent with concepts for I-mode access & operation on ITER. Following the practice of the ITER89 and ITER98 scaling laws for L- and H-mode energy confinement, an initial, illustrative attempt at an I-mode confinement scaling has also been developed. The initial characterization from C-Mod data is consistent with the observed pedestal properties in I-mode, particularly the weak degradation of energy confinement with heating power, and comparatively strong positive response to fueling and increased magnetic field. Supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, a DOE Office of Science User Facility.

  9. Fractal structure of films deposited in a tokamak

    Science.gov (United States)

    Budaev, V. P.; Khimchenko, L. N.

    2007-04-01

    The surface of amorphous films deposited in the T-10 tokamak was studied in a scanning tunnel microscope. The surface relief on a scale from 10 nm to 100 μm showed a stochastic surface topography and revealed a hierarchy of grains. The observed variety of irregular structures of the films was studied within the framework of the concept of scale invariance using the methods of fractal geometry and statistical physics. The experimental probability density distribution functions of the surface height variations are close in shape to the Cauchy distribution. The stochastic surface topography of the films is characterized by a Hurst parameter of H = 0.68-0.85, which is evidence of a nontrivial self-similarity of the film structure. The fractal character and porous structure of deposited irregular films must be considered as an important issue related to the accumulation of tritium in the ITER project. The process of film growth on the surface of tokamak components exposed to plasma has been treated within the framework of the general concept of inhomogeneous surface growth. A strong turbulence of the edge plasma in tokamaks can give rise to fluctuations in the incident flux of particles, which leads to the growth of fractal films with grain dimensions ranging from nano-to micrometer scale. The shape of the surface of some films found in the T-10 tokamak has been interpreted using a model of diffusion-limited aggregation (DLA). The growth of films according to the discrete DLA model was simulated using statistics of fluctuations observed in a turbulent edge plasma of the T-10 tokamak. The modified DLA model reproduces well the main features of the surface of some films deposited in tokamaks.

  10. Simulation of EAST vertical displacement events by tokamak simulation code

    Science.gov (United States)

    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.

  11. Injection of intense ion beam into a tokamak

    International Nuclear Information System (INIS)

    We describe an experiment to investigate the direct injection of an intense ion beam into a tokamak by means of the polarization drift. Confinement of 100 keV ions in the UCI tokamak (r = 15 cm, R = 60 cm, B/sub T/ = 6 kG) requires operation with a plasma current of 56 kA corresponding to q (limiter) = 2. Trapped ions are to be detected by a charge-exchange analyzer. The present status of the experiment will be discussed

  12. Soft X-ray tomography on HT-6B tokamak

    International Nuclear Information System (INIS)

    The tomography method for deriving soft X-ray local emissivities on HT-6B tokamak, using one horizontal array of 23 soft X-ray detectors, is described. This method has been applied to study of sawtooth oscillation and large m = 1 oscillation on HT-6B tokamak. It has been found that the large m = 1 oscillation on soft X-ray signal is caused by the rotation of plasma column containing perturbation. The reconstructed images in the phase before a sawtooth crash have shown the hot plasma core move gradually toward one side, this is considered that as the increasing m=1 kink mode. (author). 5 refs, 7 figs

  13. Fusion-product transport in axisymmetric tokamaks: losses and thermalization

    International Nuclear Information System (INIS)

    High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-β, non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated

  14. Data processing system for spectroscopy at Novillo Tokamak

    International Nuclear Information System (INIS)

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  15. Stability of Tokamaks with respect to slip motions

    International Nuclear Information System (INIS)

    Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.)

  16. A CONCEPT FOR NEXT STEP ADVANCED TOKAMAK FUSION DEVICE

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    A concept is introduced for initiating the design study of a special class of tokamak,which has a magnetic confinement configuration intermediate between contemporary advanced tokamak and the recently established spherical torus (ST,also well known by the name "spherical tokamak").The leading design parameter in the present proposal is a dimensionless geometrical parameter, the machine aspect ratio A=R0/a0=2.0,where the parameters a0 and R0 denote,respectively,the plasma (equatorial) minor radius and the plasma major radius.The aim of this choice is to technologically and experimentally go beyond the aspect ratio frontier (R0/a0≈2.5) of present day tokamaks and enter a broad unexplored domain existing on the (a0,R0) parameter space in current international tokamak database,between the data region already moderately well covered by the advanced conventional tokamaks and the data region planned to be covered by STs.Plasma minor radius a0 has been chosen to be the second basic design parameter, and consequently,the plasma major radius R0 is regarded as a dependent design parameter.In the present concept,a nominal plasma minor radius a0=1.2m is adopted to be the principal design value,and smaller values of a0 can be used for auxiliary design purposes,to establish extensive database linkage with existing tokamaks.Plasma minor radius can also be adjusted by mechanical and/or electromagnetic means to smaller values during experiments,for making suitable data linkages to existing machines with higher aspect ratios and smaller plasma minor radii.The basic design parameters proposed enable the adaptation of several confinement techniques recently developed by STs,and thereby a specially arranged central-bore region inside the envisioned tokamak torus,with retrieved space in the direction of plasma minor radius,will be available for technological adjustments and maneuverings to facilitate implementation of engineering instrumentation and real time high

  17. Research using small tokamaks. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The technical reports in these proceedings were presented at the IAEA Technical Committee Meeting on research Using Small Tokamaks, held in Ahmedabad, India, 6-7 December 1995. The purpose of this annual meeting is to provide a forum for the exchange of information on various small and medium sized plasma experiments, not only for tokamaks. The potential benefits of these research programmes are to: test theories, such as effects of the plasma rotation; check empirical scalings, such as density limits; develop fusion technology hardware; develop plasma diagnostics; such as tomography; and to train scientists, engineers, technicians, and students, particularly in developing IAEA Member States

  18. Adaptive grid finite element model of the tokamak scrapeoff layer

    Energy Technology Data Exchange (ETDEWEB)

    Kuprat, A.P.; Glasser, A.H. [Los Alamos National Lab., NM (United States)

    1995-07-01

    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.

  19. What is the fate of runaway positrons in tokamaks?

    International Nuclear Information System (INIS)

    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

  20. Multichannel bolometer for radiation measurements on the TCA tokamak

    International Nuclear Information System (INIS)

    A multichannel radiation bolometer has been developed for the Tokamak Chauffage Alfven (TCA) tokamak. It has 16 equally spaced chords that view the plasma through a narrow horizontal slit. Almost an entire vertical plasma cross section can be observed. The bolometer operates on the basis of a semiconducting element which serves as a temperature-dependent resistance. A new electronic circuit has been developed which takes advantage of the semiconductor characteristics of the detector by using feedback techniques. Measurements made with this instrument are discussed

  1. Design and Analysis of the Thermal Shield of EAST Tokamak

    Science.gov (United States)

    Xie, Han; Liao, Ziying

    2008-04-01

    EAST (Experimental Advanced Superconducting Tokamak) is a tokamak with superconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.

  2. TIBER: Tokamak Ignition/Burn Experimental Research. Final design report

    International Nuclear Information System (INIS)

    The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs

  3. Atomic data for integrated tokamak modelling

    International Nuclear Information System (INIS)

    The Integrated Tokamak Modeling Task Force (ITM-TF) was set up in 2004. The main target is to coordinate the European fusion modeling effort and providing a complete European modeling structure for International Thermonuclear Experimental Reactor (ITER), with the highest degree of flexibility. For the accurate simulation of the processes in the active fusion reactor in the ITM-TF, numerous atomic, molecular, nuclear and surface related data are required. In this work we present total-, single- and multiple-ionization and charge exchange cross sections in close connection to the ITM-TF. Interpretation of these cross sections in multi-electron ion-atom collisions is a challenging task for theories. The main difficulty is caused by the many-body feature of the collision, involving the projectile, projectile electron(s), target nucleus, and target electron(s). The classical trajectory Monte Carlo (CTMC) method has been quite successful in dealing with the atomic processes in ion-atom collisions. One of the advantages of the CTMC method is that many-body interactions are exactly taken into account related CTMC simulations for a various collision systems are presented. To highlight the efficiency of the method we present electron emission cross sections in collision between dressed Alq+ ions with He target. The theory delivers separate spectra for electrons emitted from the target and the projectile. By summing these two components in the rest frame of the target we may make a comparison with available experimental data. For the collision system in question, a significant contribution from Fermi-shuttle ionization has to be expected in the spectra at energies higher than E=0.5 me (nV)2, where me is the mass of the electron, V the projectile velocity and n an integer greater than 1. We found enhanced electron yields compared to first order theory in this region of CTMC spectra, which can be directly attributed to the contribution of Fermi-shuttle type multiple scattering

  4. Magnet systems for ''Bean-Shaped'' tokamak

    International Nuclear Information System (INIS)

    Bean-shaping of tokamak plasmas offers a method of reaching stable operation at (beta) > 10%. In order to establish the indentation of the ''bean'', a set of high- current ''pushing coils'' (> 5 MA in a reactor) must be located at the midplane as close as possible to the inboard edge of the plasma. If located in the bore of the TF coils, then maintenance of the pushing coils may be impossible, and the interlocking coils may prevent reactor modularity. If located outside, the required pushing-coil current may be unacceptably large. This dilemma is overcome with a unique TF coil design in which the inboard leg is bent outward in the form of an arc. The pushing coils are housed in the midplane indentation of this arc, just outside the TF coils but adequately close to the plasma. The arched coil transfers forces to the top and bottom legs, where it can be reacted by a clamp structure if necessary. This technique would allow demountable joints to be placed near the inoard leg (for copper TF coils). Another design approach to the pushing coils is to use liquid Li or Na as the conductor and coolant. The liquid metal ''coils'' can be placed immediately adjacent to the plasma, giving optimal control of the plasma shape with minimal coil current, although modularity of the reactor may have to be surrendered. Conceptual designs are presented of PF and TF coil systems for an ignition test reactor with about 14% and for a full-scale demonstration reactor with about 20%, both using copper TF coils

  5. Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  6. Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Fraboulet, D.

    1996-09-17

    Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.

  7. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author)

  8. Modelling multi-ion plasma gun simulations of Tokamak disruptions

    International Nuclear Information System (INIS)

    The effect of impurity ions in plasma gun ablation tests of various targets is considered. Inclusion of reasonable amounts of impurity (∼10%) is adequate to explain observed energy transmission and erosion measurements. The gun tests and the computer code calculations are relevant to the parameter range expected for major disruptions on large tokamaks

  9. Economic evaluation of fissile fuel production using resistive magnet tokamaks

    International Nuclear Information System (INIS)

    The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U3O8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U3O8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

  10. Conceptual design of Remote Control System for EAST tokamak

    International Nuclear Information System (INIS)

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

  11. Evidence of Inward Toroidal Momentum Convection in the JET Tokamak

    DEFF Research Database (Denmark)

    Tala, T.; Zastrow, K.-D.; Ferreira, J.;

    2009-01-01

    Experiments have been carried out on the Joint European Torus tokamak to determine the diffusive and convective momentum transport. Torque, injected by neutral beams, was modulated to create a periodic perturbation in the toroidal rotation velocity. Novel transport analysis shows the magnitude an...

  12. Bulk Ion Heating with ICRF Waves in Tokamaks

    DEFF Research Database (Denmark)

    Mantsinen, M. J.; Bilato, R.; Bobkov, V. V.;

    2015-01-01

    Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER a...

  13. INTOR: a first-generation tokamak experimental reactor

    International Nuclear Information System (INIS)

    An intensive, year-long, international evaluation of the next major tokamak beyond the generation of large experiments currently under construction was carried out during 1979. This evaluation consisted of the definition of objectives, an assessment of the physics and technology base and R and D needs and the identification of a set of parameters that physically characterize the machine

  14. Evaluation of the average ion approximation for a tokamak plasma

    International Nuclear Information System (INIS)

    The average ion approximation, sometimes used to calculated atomic processes in plasmas, is assessed by computing deviations in various rates over a set of conditions representative of tokamak edge plasmas. Conditions are identified under which the rates are primarily a function of the average ion charge and plasma parameters, as assumed in the average ion approximation. (Author) 19 refs., tab., 5 figs

  15. Neutronics design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

  16. Gas breakers for tokamak OHMIC-heating duty

    International Nuclear Information System (INIS)

    The current interrupting capacity of air blast and SF6 breakers is reviewed for application in tokamak ohmic-heating circuits. Particular attention is paid to generator breakers for their large current interrupting capacity and suitability for ohmic-heating circuits

  17. Feedback Control for Plasma Position on HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    LIBo; SONGXianming; LILi; LIULi; WANGMinghong; FANMingjie; CHENLiaoyuan; YAOLieying; YANGQingwei

    2003-01-01

    HL-2A is a tokamak with closed divertor. It had been built at the end of 2002 and began to discharge from then on. To further study plasma discharges in HL-2A, a feedback control system (FBCS) for plasma position bad been developed in 2003.

  18. Performance and development of the DIII-D tokamak core

    International Nuclear Information System (INIS)

    The DIII-D tokamak is an upgrade of the Doublet III configuration which has operated since early 1986. This paper presents recent advances in performance using the upper divertor, fabrication development for vanadium components, operation of the helium leak checking in a high deuterium background, and restoration of the damaged Ohmic heating solenoid

  19. A synchronization system to digitalize TJ-1 Tokamak data

    International Nuclear Information System (INIS)

    At TJ-1 Tokamak signals are stored on a 60-channel magnetic memory. In this report, a system to address those channels and synchronize readout is presented. Digitalized signals are stored in structured files on PDP-11/34 magnetic disks. (author)

  20. Solenoid-free plasma start-up in spherical tokamaks

    Science.gov (United States)

    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.

  1. Investigation of the microwave emission from the PRETEXT tokamak

    International Nuclear Information System (INIS)

    A study of the microwave emission from the PRETEXT tokamak has been conducted. Two types of emission have been observed: electron cyclotron and electron plasma frequency. Three general emission regimes have been identified. These regimes are best classified by the dimensionless parameter α, where α = ω/sub pe//Ω/sub e/

  2. Gamma ray imager on the DIII-D tokamak.

    Science.gov (United States)

    Pace, D C; Cooper, C M; Taussig, D; Eidietis, N W; Hollmann, E M; Riso, V; Van Zeeland, M A; Watkins, M

    2016-04-01

    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 electrons 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. PMID:27131674

  3. Conceptual design of Remote Control System for EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

    2014-05-15

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

  4. Dynamic diagnostics of the error fields in tokamaks

    Science.gov (United States)

    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.

  5. Spatially resolved soft x-ray spectroscopy of tokamak plasmas

    International Nuclear Information System (INIS)

    We describe the space-resolved soft x-ray (1-33nm) instrumentation developed for the Tore Supra tokamak. By using a programmable hydraulic jack to move the spectrometer, several spatial profiles (up to ten) of many impurity lines are obtained during a single plasma discharge, with a time resolution which can be as short as 600 ms. (author)

  6. Tokamak Scenario Trajectory Optimization Using Fast Integrated Simulations

    Science.gov (United States)

    Urban, Jakub; Artaud, Jean-François; Vahala, Linda; Vahala, George

    2015-11-01

    We employ a fast integrated tokamak simulator, METIS, for optimizing tokamak discharge trajectories. METIS is based on scaling laws and simplified transport equations, validated on existing experiments and capable of simulating a full tokamak discharge in about 1 minute. Rapid free-boundary equilibrium post-processing using FREEBIE provides estimates of PF coil currents or forces. We employ several optimization strategies for optimizing key trajectories, such as Ip or heating power, of a model ITER hybrid discharge. Local and global algorithms with single or multiple objective functions show how to reach optimum performance, stationarity or minimum flux consumption. We constrain fundamental operation parameters, such as ramp-up rate, PF coils currents and forces or heating power. As an example, we demonstrate the benefit of current over-shoot for hybrid mode, consistent with previous results. This particular optimization took less than 2 hours on a single PC. Overall, we have established a powerful approach for rapid, non-linear tokamak scenario optimization, including operational constraints, pertinent to existing and future devices design and operation.

  7. The Effect of Recycling in the HL-1M Tokamak

    Institute of Scientific and Technical Information of China (English)

    ZHENGYongzhen

    2002-01-01

    It is often stated that even clean tokamak discharges disrupt at high density. One possibility is that such disruption result from the energy loss arising from hydrogen recycling at the edge of the plasma.this energy loss could lead to a contraction of the current channel and the production of a disruptively unstable configuration.

  8. Stability of infernal and ballooning modes in advanced tokamak scenarios

    NARCIS (Netherlands)

    Holties, H. A.; Huysmans, G. T. A.; Goedbloed, J. P.; Kerner, W.; Parail, V.V.; Soldner, F. X.

    1996-01-01

    A numerical parameter study has been performed in order to find MHD stable operating regimes for advanced tokamak experiments In this study we have concentrated on internal modes. Ballooning stability and stability with respect to infernal modes are considered. The calculations confirm that pressure

  9. General Description of Ideal Tokamak MHD Instability Ⅱ

    Institute of Scientific and Technical Information of China (English)

    石秉仁

    2002-01-01

    In this subsequent study on general description of ideal tokamak MHD instability,the part Ⅱ, by using a coordinate with rectified magnetic field lines, the eigenmode equationsdescribing the low-mode-number toroidal Alfven modes (TAE and EAE) are derived through afurther expansion of the shear Alfven equation of motion.

  10. Fokker-Planck Study of Tokamak Electron Cyclotron Resonance Heating

    Institute of Scientific and Technical Information of China (English)

    SHIBingren; LONGYongxing; DONGJiaqi; LIWenzhong; JIAOYiming; WANGAike

    2002-01-01

    In this study, we add a subroutine for describing the electron cyclotron resonant heating calculation to the Fokker-Planck code. By analyzing the wave-particle resonance condition in tokamak plasma and the fast motion of electrons along magnetic field lines, suitable quasi-linear diffusion coefficients are given.

  11. MHD analysis of edge instabilities in the JET tokamak

    NARCIS (Netherlands)

    Perez von Thun, Christian Pedro

    2004-01-01

    The aim of nuclear fusion energy research is to demonstrate the feasibility of nuclear fusion reactors as a future energy source. The tokamak is the most advanced fusion machine to date, and is most likely the first system to be converted into a reactor. An important subject of nuclear fusion resear

  12. Disruption avoidance through active magnetic feedback in tokamak plasmas

    Science.gov (United States)

    Paccagnella, Roberto; Zanca, Paolo; Yanovskiy, Vadim; Finotti, Claudio; Manduchi, Gabriele; Piron, Chiara; Carraro, Lorella; Franz, Paolo; RFX Team

    2014-10-01

    Disruptions avoidance and mitigation is a fundamental need for a fusion relevant tokamak. In this paper a new experimental approach for disruption avoidance using active magnetic feedback is presented. This scheme has been implemented and tested on the RFX-mod device operating as a circular tokamak. RFX-mod has a very complete system designed for active mode control that has been proved successful for the stabilization of the Resistive Wall Modes (RWMs). In particular the current driven 2/1 mode, unstable when the edge safety factor, qa, is around (or even less than) 2, has been shown to be fully and robustly stabilized. However, at values of qa (qa > 3), the control of the tearing 2/1 mode has been proved difficult. These results suggested the idea to prevent disruptions by suddenly lowering qa to values around 2 where the tearing 2/1 is converted to a RWM. Contrary to the universally accepted idea that the tokamaks should disrupt at low qa, we demonstrate that in presence of a well designed active control system, tokamak plasmas can be driven to low qa actively stabilized states avoiding plasma disruption with practically no loss of the plasma internal energy.

  13. High- Q plasmas in the TFTR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.; Barnes, C.W.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; Bush, C.E.; Dylla, H.F.; Efthimion, P.C.; Fredrickson, E.D.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.C.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S.J.; LaMarche, P.H.; LeBlanc, B.; Mansfield, D.K.; Marmar, E.S.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.K.; Paul, S.F.; Pitcher, S.; Ramsey, A.T.; Redi, M.H.; Sabbagh, S.A.; Scott, S.D.; Snipes, J.; Stevens, J.; Strachan, J.D.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Terry, J.L.; Timberlake, J.R.; Towner, H.H.; Ulrickson, M.; von Goeler, S.; Wieland, R.M.; Williams, M.; Wilson, J.R.; Wong, K.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J. (Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (USA))

    1991-08-01

    In the Tokamak Fusion Test Reactor (TFTR) (Plasma Phys. Controlled Fusion {bold 26}, 11 (1984)), the highest neutron source strength {ital S}{sub {ital n}} and D--D fusion power gain {ital Q}{sub DD} are realized in the neutral-beam-fueled and heated supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, {ital S}{sub {ital n}} increases approximately as {ital P}{sup 1.8}{sub {ital b}}. The highest-{ital Q} shots are characterized by high {ital T}{sub {ital e}} (up to 12 keV), {ital T}{sub {ital i}} (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad {ital T}{sub {ital e}} profiles, and lower {ital Z}{sub eff}. Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, {ital Q}{sub DD} increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness ({ital n}{sub {ital e}}(0)/{l angle}{ital n}{sub {ital e}}{r angle}) during the beam pulse. To date, the best fusion results are {ital S}{sub {ital n}}=5{times}10{sup 16} n/sec, {ital Q}{sub DD}=1.85{times}10{sup {minus}3}, and neutron yield=4.0{times}10{sup 16} n/pulse, obtained at {ital I}{sub {ital p}}=1.6--1.9 MA and beam energy {ital E}{sub {ital b}}=95--103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%--60% of {ital S}{sub {ital n}} arises from beam--target reactions, with the remainder divided between beam--beam and thermonuclear reactions, the thermonuclear fraction increasing with {ital P}{sub {ital b}}.

  14. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  15. Design and construction of Alborz tokamak vacuum vessel system

    Energy Technology Data Exchange (ETDEWEB)

    Mardani, M., E-mail: mohsenmardani@gmail.com [Amirkabir University of Technology (Tehran Polytechnic), Tehran (Iran, Islamic Republic of); Amrollahi, R.; Koohestani, S. [Amirkabir University of Technology (Tehran Polytechnic), Tehran (Iran, Islamic Republic of)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. Black-Right-Pointing-Pointer As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. Black-Right-Pointing-Pointer A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma-surface interaction and localizes the particle recycling. Black-Right-Pointing-Pointer Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma-surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

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

    International Nuclear Information System (INIS)

    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 as 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

  17. Spatially Resolved Spectra from a new X-ray Imaging Crystal Spectrometer for Measurements of Ion and Electron Temperature Profiles

    Energy Technology Data Exchange (ETDEWEB)

    Bitter, M; Stratton, B; Roquemore, A; Mastrovito, D; Lee, S; Bak, J; Moon, M; Nam, U; Smith, G; Rice, J; Beiersdorfer, P; Fraenkel, B

    2004-08-10

    A new type of high-resolution X-ray imaging crystal spectrometer is being developed to measure ion and electron temperature profiles in tokamak plasmas. The instrument is particularly valuable for diagnosing plasmas with purely Ohmic heating and rf heating, since it does not require the injection of a neutral beam - although it can also be used for the diagnosis of neutral-beam heated plasmas. The spectrometer consists of a spherically bent quartz crystal and a two-dimensional position-sensitive detector. It records spectra of helium-like argon (or krypton) from multiple sightlines through the plasma and projects a de-magnified image of a large plasma cross-section onto the detector. The spatial resolution in the plasma is solely determined by the height of the crystal, its radius of curvature, and the Bragg angle. This new X-ray imaging crystal spectrometer may also be of interest for the diagnosis of ion temperature profiles in future large tokamaks, such as KSTAR and ITER, where the application of the presently used charge-exchange spectroscopy will be difficult, if the neutral beams do not penetrate to the plasma center. The paper presents the results from proof-of-principle experiments performed with a prototype instrument at Alcator C-Mod.

  18. Development of the gas puff charge exchange recombination spectroscopy (GP-CXRS) technique for ion measurements in the plasma edge

    International Nuclear Information System (INIS)

    A novel charge-exchange recombination spectroscopy (CXRS) diagnostic method is presented, which uses a simple thermal gas puff for its donor neutral source, instead of the typical high-energy neutral beam. This diagnostic, named gas puff CXRS (GP-CXRS), is used to measure ion density, velocity, and temperature in the tokamak edge/pedestal region with excellent signal-background ratios, and has a number of advantages to conventional beam-based CXRS systems. Here we develop the physics basis for GP-CXRS, including the neutral transport, the charge-exchange process at low energies, and effects of energy-dependent rate coefficients on the measurements. The GP-CXRS hardware setup is described on two separate tokamaks, Alcator C-Mod and ASDEX Upgrade. Measured spectra and profiles are also presented. Profile comparisons of GP-CXRS and a beam based CXRS system show good agreement. Emphasis is given throughout to describing guiding principles for users interested in applying the GP-CXRS diagnostic technique

  19. Is X-ray emissivity constant on magnetic flux surfaces?

    International Nuclear Information System (INIS)

    Knowledge of the elongations and shifts of internal magnetic flux surfaces can be used to determine the q profile in elongated tokamak plasmas. X-ray tomography is thought to be a reasonable technique for independently measuring internal flux surface shapes, because it is widely believed that X-ray emissivity should be constant on a magnetic flux surface. In the Alcator C-Mod tokamak, the X-ray tomography diagnostic system consists of four arrays of 38 chords each. A comparison of reconstructed X-ray contours with magnetic flux surfaces shows a small but consistent discrepancy in the radial profile of elongation. Numerous computational tests have been performed to verify these findings, including tests of the sensitivity to calibration and viewing geometry errors, the accuracy of the tomography reconstruction algorithms, and other subtler effects. We conclude that the discrepancy between the X-ray contours and the magnetic flux surfaces is real, leading to the conclusion that X-ray emissivity is not exactly constant on a flux surface. (orig.)

  20. Development of the gas puff charge exchange recombination spectroscopy (GP-CXRS) technique for ion measurements in the plasma edge

    Energy Technology Data Exchange (ETDEWEB)

    Churchill, R. M.; Theiler, C.; Lipschultz, B. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Dux, R.; Pütterich, T.; Viezzer, E. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, D-85748 Garching (Germany); Collaboration: Alcator C-Mod Team; ASDEX Upgrade Team

    2013-09-15

    A novel charge-exchange recombination spectroscopy (CXRS) diagnostic method is presented, which uses a simple thermal gas puff for its donor neutral source, instead of the typical high-energy neutral beam. This diagnostic, named gas puff CXRS (GP-CXRS), is used to measure ion density, velocity, and temperature in the tokamak edge/pedestal region with excellent signal-background ratios, and has a number of advantages to conventional beam-based CXRS systems. Here we develop the physics basis for GP-CXRS, including the neutral transport, the charge-exchange process at low energies, and effects of energy-dependent rate coefficients on the measurements. The GP-CXRS hardware setup is described on two separate tokamaks, Alcator C-Mod and ASDEX Upgrade. Measured spectra and profiles are also presented. Profile comparisons of GP-CXRS and a beam based CXRS system show good agreement. Emphasis is given throughout to describing guiding principles for users interested in applying the GP-CXRS diagnostic technique.