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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. 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.

  3. 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.

  4. 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

  5. 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

  6. 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...

  7. 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.

  8. 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

  9. 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.

  10. The high resolution video capture system on the alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    A new system for routine digitization of video images is presently operating on the Alcator C-Mod tokamak. The PC-based system features high resolution video capture, storage, and retrieval. The captured images are stored temporarily on the PC, but are eventually written to CD. Video is captured from one of five filtered RS-170 CCD cameras at 30 frames per second (fps) with 640x480 pixel resolution. In addition, the system can digitize the output from a filtered Kodak Ektapro EM Digital Camera which captures images at 1000 fps with 239x192 resolution. Present views of this set of cameras include a wide angle and a tangential view of the plasma, two high resolution views of gas puff capillaries embedded in the plasma facing components, and a view of ablating, high speed Li pellets. The system is being used to study (1) the structure and location of visible emissions (including MARFEs) from the main plasma and divertor, (2) asymmetries in gas puff plumes due to flows in the scrape-off layer (SOL), and (3) the tilt and cigar-shaped spatial structure of the Li pellet ablation cloud. copyright 1997 American Institute of Physics

  11. 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

  12. Enhanced D-Alpha H-mode studies in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    A favorable regime of H-mode confinement, seen on the Alcator C-Mod tokamak is described. Following a brief period of ELM-free H-mode, the plasma evolves into the Enhanced D-Alpha (EDA) H-mode which is characterized by very good energy confinement, the complete absence of large, intermittent type I ELMs, finite impurity and majority species confinement, and low radiated power fraction. Accompanying the EDA H-mode, a quasi-coherent (QC) edge mode is observed, and found to be responsible for particle transport through the edge confinement barrier. The QC-mode is localized within the strong density gradient region, and has poloidal wavenumber k θ≅5cm-1 and lab-frame frequency of ≅100 kHz. Parametric studies show that the conditions which promote EDA include moderate safety factor (q95>3.5), high triangularity (δ>0.35) and high target density (ne>1.2x20m-3). EDA H-mode is readily obtained in purely ohmic and well as in ICRF auxiliary-heated discharges. (author)

  13. Overview of recent results from a Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Recent results from the compact, high field, Alcator C-Mod tokamak program are summarized. H-mode threshold studies have demonstrated that the threshold appears to be closely related to local edge plasma parameters: for fixed field and plasma current, Te(ψ95) takes on a density independent value at the transition. The Enhanced D-Alpha H-Mode (EDA) regime has been investigated. EDA is distinct from ELM free H-mode, in that there is no accumulation of impurities, and at the same time EDA does not exhibit large discrete ELMs. The energy confinement is degraded by only about 10%, compared to ELM free. Comparisons for EDA with ELMy H-Mode database scalings indicate τEDA ∼1.2 τITER97H. Strong toroidal rotation is observed in ICRF-only auxiliary heated plasmas; the rotation increases with plasma pressure, and decreases with increasing plasma current. The inferred radial electric field reaches the order of 30 kV/m near the center of the plasma. Through feedback controlled nitrogen impurity puffing, steady state detached EDA H-Modes have been achieved with Zeff E is reduced by about 10% in the detached case, compared to the confinement before the N2 puff begins. The heat load to the divertor is reduced by a factor of 4. Volume recombination rates are measured in the divertor, using 2-d tomography of Balmer series TV movies. Volume recombination can be a significant contributor to the overall reduction in ion current to the divertor plates which occurs in detachment. Particle balance measurements indicate that the divertor and main chamber plasmas are largely isolated from one another, at least with regard to particle recycling, with most of the main chamber (core plus scrape-off) fueling coming from neutrals in the main chamber volume. With the addition of Lower Hybrid Current Drive, C-Mod would be an ideal vehicle for investigation of advanced tokamak operation with fully relaxed current profiles. Detailed modeling indicates that discharges approaching the β limit (

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

    CERN Document Server

    Pankin, A Y; Kritz, A H; Park, G Y; Chang, C S; Brunner, D; Groebner, R J; Hughes, J W; LaBombard, B; Terry, J L; Ku, S

    2015-01-01

    The guiding-center kinetic neoclassical transport code, XGC0, [C.S. 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 width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current $I_{\\rm p}$. 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 collisio...

  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. 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

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

    International Nuclear Information System (INIS)

    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 Ip. 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/Ip 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/Ip scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/Ip 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

  18. 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.

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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.

  7. Development of a reciprocating probe servomotor control system with real-time feedback on plasma position for the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Kuang, A. Q.; Labombard, B.; Burke, W.

    2015-11-01

    Reciprocating probe drives are one of the diagnostic workhorses in the boundary of magnetic confinement fusion experiments. The probe is scanned into an exponentially increasing heat flux, which demands a prompt and precise turn around to maintain probe integrity. A new linear servomotor controlled reciprocating drive utilizing a commercial linear servomotor and drive controller has been developed for the Alcator C-Mod tokamak. The quick response of the controller (able to apply an impulse of 50A in about 1ms) along with real-time plasma measurements from a Mirror Langmuir Probe (MLP) allows for real-time control of the probe trajectory based on plasma conditions at the probe tip. Since the primary concern for probe operation is overheating, an analog circuit has been created that computes the surface temperature of the probe from the MLP measurements. The probe can be programmed to scan into the plasma at various times and then turns around when the computed surface temperature reaches a set threshold, maximizing the scan depth into the plasma while avoiding excessive heating. Design, integration, and first measurements with this new system will be presented. This work was supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, A DOE SC User Facility.

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. The Alcator C-MOD power system

    International Nuclear Information System (INIS)

    The-14 magnets or the Alcator C-MOD device require over 500 MJ to be delivered in a few seconds. Peak powers approaching 300 MW will be delivered from the upgraded 225-MVA alternator used to power the Alcator C device. Most magnets have dedicated power conversion systems in order to maximize flexibility and to allow for operation with a single-null divertor. Details of the power system design are presented. The Toroidal Field (TF) conversion system features 260 kA output coupled with an auto-tap changing design to increase power factor. The 50-kA Ohmic Heating (OH) system utilizes a single DC interrupter to switch 3 OH magnets and 3 OH converters. The vertical position control system uses a combination of fast, chopper-type supplies and conventional line-commutated converters to combine fast response with low cost. The radial position control system utilizes a reconfigured Alcator C-era OH supply in combination with surplus DC traction motors outfitted with flywheels. 12 refs., 9 figs

  14. Overview of Alcator C-Mod recent results

    International Nuclear Information System (INIS)

    Research on Alcator C-Mod tokamak focusses on exploiting compact, high density plasmas to understand core transport and heating, the physics of the H-mode transport barrier, and the dynamics of the scrape-off-layer and divertor. Rapid toroidal acceleration of the plasma core is observed during ohmic heated H-modes and indicates a momentum pinch or similar transport mechanism. Core thermal transport observations support a critical gradient interpretation, but with gradients that disagree with present theoretical values. High resolution measurements of the H-mode barrier have been obtained including impurity and neutral densities, and the instability apparently responsible for the favorable 'Enhanced D-alpha' regime has been identified. Divertor bypass dynamic control experiments have directly addressed the important questions surrounding main chamber recycling and the effect of divertor closure on impurities and confinement. Future plans include quasi-steady-state Advanced Tokamak plasmas using Lower Hybrid current drive. (author)

  15. 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

  16. High speed movies of turbulence in Alcator C-Mod

    International Nuclear Information System (INIS)

    A high speed (250 kHz), 300 frame charge coupled device camera has been used to image turbulence in the Alcator C-Mod Tokamak. The camera system is described and some of its important characteristics are measured, including time response and uniformity over the field-of-view. The diagnostic has been used in two applications. One uses gas-puff imaging to illuminate the turbulence in the edge/scrape-off-layer region, where D2 gas puffs localize the emission in a plane perpendicular to the magnetic field when viewed by the camera system. The dynamics of the underlying turbulence around and outside the separatrix are detected in this manner. In a second diagnostic application, the light from an injected, ablating, high speed Li pellet is observed radially from the outer midplane, and fast poloidal motion of toroidal striations are seen in the Li+ light well inside the separatrix

  17. 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.

  18. Extension of Alcator C-mod's ICRF Experimental Capability

    International Nuclear Information System (INIS)

    A new 4-strap single-ended ICRF antenna has been added to the Alcator C-Mod tokamak. PPPL designed, fabricated, and tested the antenna up to 45 kV on an rf test stand. It is capable of symmetric phasing for ICRF heating studies, and asymmetric phasing with an improved directed wave spectrum for current drive. Two new 2 MW transmitters, tunable from 40-80 MHz, allow operation in plasma at 43, 60, and 78 MHz to match a variety of toroidal fields and plasma conditions. This addition increases the total available ICRF power to 4 MW at 80 MHz plus 4 MW at 40-80 MHz. Plasma heating and current drive experiments at the extended power levels and new frequencies are planned, and initial system performance will be discussed

  19. Characteristics of high-confinement modes in Alcator C Mod

    International Nuclear Information System (INIS)

    The regime of high particle and energy confinement known as the H mode [Phys. Rev. Lett. 49, 1408 (1982)] has been extended to a unique range of operation for divertor tokamaks up to toroidal fields of nearly 8 T, line-averaged electron densities of 3x1020 m-3, and surface power densities of nearly 0.6 MW/m2 in the compact high-field tokamak Alcator C Mod [Phys. Plasmas 1, 1511 (1994)]. H modes are achieved in Alcator C Mod with Ion Cyclotron Resonant Frequency (ICRF) heating and with Ohmic heating alone without boronization of the all molybdenum tiled first wall. Large increases in charge exchange flux are observed during the H mode over the entire range of energies from 2 to 10 keV. There appears to be an upper limit to the midplane neutral pressure, of about 0.08 Pa above which no H modes have been observed. The plasmas with the best energy confinement have the lowest midplane neutral pressures, below 0.01 Pa. There is an edge electron temperature threshold such that Te≥280 eV ±40 eV for sustaining the H mode, which is equal at L endash H and H endash L transitions. The hysteresis in the threshold power between L endash H and H endash L transitions is less than 25% on average. Both core and edge particle confinement improve by a factor of 2 endash 4 from L mode to H mode. Energy confinement also improves by up to a factor of 2 over L mode. copyright 1996 American Institute of Physics

  20. 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. PMID:26931846

  1. 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

  2. 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

  3. 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.

  4. Thyristor DC circuit breakers for alcator C-MOD

    International Nuclear Information System (INIS)

    This paper reports on the Alcator C-MOD PF system which requires DC circuit breakers in 5 poloidal field magnets for plasma initiation. Vacuum circuit breakers developed for the Alcator A and Alcator C experiments offered a proven technology but marginal reliability and relatively high maintenance costs. The thick vacuum chamber walls and heavy steel superstructure of Alcator C-MOD make relatively low start-up voltages attractive to limit eddy currents during plasma initiation. Design values and simulations for 3 switches are presented. Switch sections are rated 2 kV maximum interrupting voltage with bipolar current ratings of 50, 25, and 15 kA. A pulse-forming network provides the required 400 μsec turn-off pulse to the SCRs. The design is strongly influenced by the parasitic inductance in bus and circuit components. Time and cost limits precluded the construction of prototypes, so extensive simulation of the circuit was required. Two separate simulation approaches were cross-checked for accuracy and consistency

  5. Overview of the Alcator C-Mod Program

    International Nuclear Information System (INIS)

    Research on the Alcator C-Mod tokamak has emphasized RF heating, self-generated flows, momentum transport, scrape-off layer turbulence and transport and the physics of transport barrier transitions, stability and control. The machine operates with PRF up to 6 MW corresponding to power densities on the antenna of 10 MW/m2. Analysis of rotation profile evolution, produced in the absence of external drive, allows transport of angular momentum in the plasma core to be computed and compared between various operating regimes. Momentum is clearly seen diffusing and convecting from the plasma edge on time scales similar to the energy confinement time and much faster than neo-classical transport. Scrape-off layer (SOL) turbulence and transport have been studied with fast scanning electrostatic probes, situated at several poloidal locations and with gas puff imaging. Strong poloidal asymmetries are found in profiles and fluctuations confirming the essential ballooning character of the turbulence and transport. Plasma topology has a dominant effect on the magnitude and direction of both core rotation and SOL flows. The correlation of self-generated plasma flows and topology has led to a novel explanation for the dependence of the H-mode power threshold on the ∇B drift direction. Research into internal transport barriers (ITB) has focused on control of the barrier strength, and location. The foot of the barrier could be moved to larger minor radius by lowering q or BT. The barriers, which are produced in C-Mod by off-axis RF heating, can be weakened by the application of on-axis power. Gyro-kinetic simulations suggest that the control mechanism is due to the temperature dependence of trapped electron modes (TEM) which are destabilized by the large density gradients. A set of non-axisymmetric coils was installed allowing intrinsic error fields to be measured and compensated. These also enabled the determination of the mode locking threshold and, by comparison with data from

  6. 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.

  7. 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

  8. Light impurity transport at an internal transport barrier in Alcator C-Mod

    International Nuclear Information System (INIS)

    Density profiles for a light impurity, boron, are reported for internal transport barrier (ITB) discharges in Alcator C-Mod. During the ITB, the light impurity gradient steepens because the impurity pinch increases relative to diffusion. The ITB-induced impurity profile steepening is at approximately the same major radius as that for the main-ion profile. Neoclassical transport does not describe the light impurity profiles but transport is closer to neoclassical in the ITB region. In previous work on C-Mod, profiles of seeded heavy impurities (introduced by puffing) peaked during the ITB, but a marked difference between transport of heavy and light impurities has been reported for other tokamaks. With the addition of light impurity profiles described here, the ITB on C-Mod is shown to share additional profile traits with the ITB on other tokamaks. This confirms that the macroscopic features of the C-Mod ITB are similar to those on other devices although it leaves open the details of the onset of the ITB

  9. 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.))

  10. 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.

  11. Burst statistics in Alcator C-Mod SOL turbulence

    International Nuclear Information System (INIS)

    Bursty fluctuations in the scrape-off layer (SOL) of Alcator C-Mod have been analyzed using gas puff imaging data. This reveals many of the same fluctuation properties as Langmuir probe measurements, including normal distributed fluctuations in the near SOL region while the far SOL plasma fluctuations are dominated by large amplitude bursts due to radial motion of blob-like structures. Conditional averaging reveals burst wave forms with a fast rise and slow decay and exponentially distributed burst amplitudes and waiting times. Based on this, a stochastic model of burst dynamics is constructed. The model predicts that fluctuation amplitudes should follow a Gamma distribution. This is shown to be a good description of the gas puff imaging data for a range of line-averaged densities

  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. Migration of alcator C-Mod computer infrastructure to Linux

    International Nuclear Information System (INIS)

    The Alcator C-Mod fusion experiment at MIT in Cambridge, Massachusetts has been operating for twelve years. The data handling for the experiment during most of this period was based on MDSplus running on a cluster of VAX and Alpha computers using the OpenVMS operating system. While the OpenVMS operating system provided a stable reliable platform, the support of the operating system and the software layered on the system has deteriorated in recent years. With the advent of extremely powerful low cost personal computers and the increasing popularity and robustness of the Linux operating system a decision was made to migrate the data handling systems for C-Mod to a collection of PC's running Linux. This paper will describe the new system configuration, the effort involved in the migration from OpenVMS, the results of the first run campaign under the new configuration and the impact the switch may have on the rest of the MDSplus community

  16. Perturbative thermal diffusivity from partial sawtooth crashes in Alcator C-Mod

    Science.gov (United States)

    Creely, A. J.; White, A. E.; Edlund, E. M.; Howard, N. T.; Hubbard, A. E.

    2016-03-01

    Perturbative thermal diffusivity has been measured on Alcator C-Mod via the use of the extended-time-to-peak method on heat pulses generated by partial sawtooth crashes. Perturbative thermal diffusivity governs the propagation of heat pulses through a plasma. It differs from power balance thermal diffusivity, which governs steady state thermal transport. Heat pulses generated by sawtooth crashes have been used extensively in the past to study heat pulse thermal diffusivity (Lopes Cardozo 1995 Plasma Phys. Control. Fusion 37 799), but the details of the sawtooth event typically lead to non-diffusive ‘ballistic’ transport, making them an unreliable measure of perturbative diffusivity on many tokamaks (Fredrickson et al 2000 Phys. Plasmas 7 5051). Partial sawteeth are common on numerous tokamaks, and generate a heat pulse without the ‘ballistic’ transport that often accompanies full sawteeth (Fredrickson et al 2000 Phys. Plasmas 7 5051). This is the first application of the extended-time-to-peak method of diffusivity calculation (Tubbing et al 1987 Nucl. Fusion 27 1843) to partial sawtooth crashes. This analysis was applied to over 50 C-Mod shots containing both L- and I-Mode. Results indicate correlations between perturbative diffusivity and confinement regime (L- versus I-mode), as well as correlations with local temperature, density, the associated gradients, and gradient scale lengths (a/L Te and a/L n ). In addition, diffusivities calculated from partial sawteeth are compared to perturbative diffusivities calculated with the nonlinear gyrokinetic code GYRO. We find that standard ion-scale simulations (ITG/TEM turbulence) under-predict the perturbative thermal diffusivity, but new multi-scale (ITG/TEM coupled with ETG) simulations can match the experimental perturbative diffusivity within error bars for an Alcator C-Mod L-mode plasma. Perturbative diffusivities extracted from heat pulses due to partial sawteeth provide a new constraint that can be used to

  17. Production of internal transport barriers via self-generated mean flows in Alcator C-Mod

    International Nuclear Information System (INIS)

    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.

  18. 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

  19. 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.

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

    International Nuclear Information System (INIS)

    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 auxiliary 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 neutrals 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 radial

  1. Edge turbulence in different density regimes in Alcator C-Mod experiment

    International Nuclear Information System (INIS)

    Plasma edge turbulence of Alcator C-Mod tokamak is studied with a fast camera in different density regimes. The statistical properties of the fluctuations, as well as the behaviour of the blobs, are characterized in plasma discharges at different normalized densities, studying the link between the edge turbulence and the Greenwald limit. It is shown that approaching the Greenwald density limit, the edge velocity field measured with the cross-correlation technique changes and the strong fluctuations, which for standard discharges develop mainly outside the separatrix, extend also in the radial region inside the last closed flux surface. At the same time, the blobs cover a larger radial region, suggesting a strong impact of the edge turbulence and transport on the Greenwald limit.

  2. Edge Zonal Flows and Blob Propagation in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S; Agostini, M; Davis, B; Grulke, O; Hager, R; Hughes, J; LaBombard, B; D& #x27; Ippolito, D A; Myra, J R

    2011-07-25

    Here we describe recent measurements of the 2-D motion of turbulence in the edge and scrape-off layer (SOL) of the Alcator C-Mod tokamak. This data was taken using the outer midplane gas puff imaging (GPI) camera, which views a 6 cm radial by 6 cm poloidal region near the separatrix just below the outer midplane [1]. The data were taken in Ohmic or RF heated L-mode plasmas at 400,000 frames/sec for {approx}50 msec/shot using a Phantom 710 camera in a 64 x 64 pixel format. The resulting 2-D vs. time movies [2] can resolve the structure and motion of the turbulence on a spatial scale covering 0.3-6 cm. The images were analyzed using either a 2-D cross-correlation code (Sec. 2) or a 2-D blob tracking code (Sec. 3).

  3. Impurity transport experiments in the edge plasma of Alcator C-Mod using gas injection plumes

    International Nuclear Information System (INIS)

    Understanding impurity and bulk plasma transport in the scrape-off layer (SOL) of tokamak plasmas is critical for the design of future reactors. The development of a system on Alcator C-Mod for inferring impurity transport behaviour parallel and perpendicular to local magnetic field lines from impurity emission patterns ('plumes') generated by local gas injection will be presented. Gas is injected at variable location in the SOL through the end of a reciprocating fast-scanning probe. Carbon plumes are generated by puffing ethylene gas (C2H4) through the probe over a period of ∼8-10 ms. Two intensified CCD cameras are used to record C+1 and C+2 emission patterns from near-perpendicular views. Flows parallel and perpendicular to the magnetic field have been observed in both views. In principle, the data allow a full 3D reconstruction of the impurity dispersal with ∼1 mm spatial resolution

  4. 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.

  5. 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.

  6. 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

  7. Neutral Transport Simulations of Gas Puff Imaging Experiments on Alcator C-Mod; TOPICAL

    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

  8. 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

  9. High density LHRF experiments in Alcator C-Mod and implications for reactor scale devices

    Science.gov (United States)

    Baek, S. G.; Parker, R. R.; Bonoli, P. T.; Shiraiwa, S.; Wallace, G. M.; LaBombard, B.; Faust, I. C.; Porkolab, M.; Whyte, D. G.

    2015-04-01

    Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the ‘density limit’, and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent

  10. Production of Internal Transport Barriers by Intrinsic Flow Drive in Alcator C-Mod

    International Nuclear Information System (INIS)

    Full text: New results suggest that changes observed in the intrinsic toroidal rotation influence the internal transport barrier (ITB) formation in the Alcator C-Mod tokamak. Detailed plasma rotation and ion temperature profile measurements are combined with linear and non-linear gyrokinetic simulation to examine the effects of the self-generated rotational shear on the transport changes that occur in C-Mod ITB plasmas. These arise when the resonance for 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 x 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. The newly available detailed measurement of the ion temperature demonstrates that the radial profile flattens as the ICRF resonance position moves off axis, decreasing the drive for ITG the instability as well. These results are the first evidence that intrinsic rotation can affect confinement in ITB plasmas, and suggest that this regime could be achievable in ITER and in future reactor experiments. (author)

  11. The ORNL fast wave ICRF [Ion Cyclotron Range of Frequencies] antenna for Alcator C-Mod

    International Nuclear Information System (INIS)

    A fast wave ICRF antenna is being designed for Alcator C-Mod which is prototypical in many respects of the baseline launcher design for the Compact Ignition Tokamak (CIT). The C-Mod launcher has a single current strap, with a strap and cavity geometry very similar to one quadrant of the CIT launcher, which has four straps in a 2 x 2 configuration. The antenna fits entirely within an 8 in. wide by 25 in. long port and is radially movable over a distance of 15 cm. It will operate at a frequency of 80 MHz for pulse lengths up to 1 s, at a maximum power level of 2 MW, corresponding to a power flux of >1.5 kW/cm2. The antenna is an end fed double loop configuration in which the current strap is grounded in the middle to provide mechanical support. The design includes a disruption support system which accommodates thermal expansion of the antenna box while supporting large disruption loads. It also includes a novel matching system consisting of an external resonant loop with two shunt capacitors serving as tuning/matching elements. 8 refs., 5 figs., 12 tabs

  12. Zonal flow production in the L–H transition in Alcator C-Mod

    International Nuclear Information System (INIS)

    Transitions of tokamak confinement regimes from low- to high-confinement are studied on Alcator C-Mod (Hutchinson et al 1994 Phys. Plasmas 1 1511) tokamak using gas-puff-imaging, with a focus on the interaction between the edge drift-turbulence and the local shear flow. Results show that the nonlinear turbulent kinetic energy transfer rate into the shear flow becomes comparable to the estimated value of the drift turbulence growth rate at the time the turbulent kinetic energy starts to drop, leading to a net energy transfer that is comparable to the observed turbulence losses. A corresponding growth is observed in the shear flow kinetic energy. The above behavior is demonstrated across a series of experiments. Thus both the drive of the edge zonal flow and the initial reduction of turbulence fluctuation power are shown to be consistent with a lossless kinetic energy conversion mechanism, which consequently mediates the transition into H-mode. The edge pressure gradient is then observed to build on a slower (1 ms) timescale, locking in the H-mode state. These results unambiguously establish the time sequence of the transition as: first the peaking of the normalized Reynolds power, then the collapse of the turbulence, and finally the rise of the diamagnetic electric field shear as the L–H transition occurs. (paper)

  13. 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...

  14. Survey of the TS-ECE Discrepancy and recent investigations in ICRF heated plasmas at Alcator C-Mod

    Directory of Open Access Journals (Sweden)

    Reinke M. L.

    2012-09-01

    Full Text Available This paper reports on a new investigation of the long-standing, unresolved discrepancy between Thomson Scattering (TS and Electron Cyclotron Emission (ECE measurements of electron temperature in high temperature tokamak plasmas. At the Alcator C-Mod tokamak, ion cyclotron range of frequency (ICRF heating is used to produce high temperature conditions where the TS- ECE discrepancy, as observed in the past at JET and TFTR, should appear. Plasmas with Te(0 up to 8 keV are obtained using three different heating scenarios: Ion Cyclotron Resonance Heating (ICRH, ICRF mode conversion heating and a combination of the two heating methods. This is done in order to explore the hypothesis that ICRH-generated fast ions may be related to the discrepancy. In all high temperature cases at C-Mod, we find no evidence for the type of discrepancy reported at JET and TFTR. Here we present the C-Mod results along with a summary of past work on the TS-ECE discrepancy.

  15. Characterization of enhanced Dα high-confinement modes in Alcator C-Mod

    International Nuclear Information System (INIS)

    Regimes of high-confinement mode have been studied in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Plasmas with no edge localized modes (ELM-free) have been compared in detail to a new regime, enhanced Dα (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>3.7), for moderate plasma shaping, (triangularity ∼0.35 - 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 1999 American Institute of Physics

  16. 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.

  17. Marginal stability studies of microturbulence near ITB onset on Alcator C-Mod

    Science.gov (United States)

    Baumgaertel, J. A.; Redi, M. H.; Budny, R. V.; McCune, D. C.; Dorland, W.; Fiore, C. L.

    2004-11-01

    Insight into microturbulence and transport in tokamak plasmas is being sought using linear simulations of drift waves near the onset time of an internal transport barrier (ITB) on Alcator C-Mod. The experiment being studied is an off-axis RF heated H-mode which develops an ITB with toroidal velocity reversal near the onset time [1]. We investigate marginal stability of ITG, ETG, and TEM modes for conditions in the plasma core as well as at and outside the ITB region. Experiment does not always show that drift wave modes are near marginal stability [2]. Based on input files for the parallel code GS2 [3], produced by TRXPL following TRANSP analysis, we examine the variability of mode growth rates as functions of ion temperature and density gradients. The roles of microturbulent stabilizing and driving forces and possibilities for transport barrier control are investigated. *Science Undergraduate Laboratory Intern (University of Washington) [1] C. L. Fiore, et al. Phys. Plas. 8, 2023 (2001). [2] X. Garbet, et al. EPS-2004, London, UK, paper I5-10. [3] M. Kotschenreuther, et al, Comp.Phys. Com. 88, 128 (1995).

  18. High power density H-modes in Alcator C-Mod

    International Nuclear Information System (INIS)

    Optimization studies of energy confinement in H-mode plasmas heated by ICRF waves were performed on the compact high-field tokamak Alcator C-Mod. Reduction of the radiated power from the main plasma by boronization of the molybdenum first wall made a major impact on improving the quality of H-mode, as characterized by the confinement enhancement factor over the ITER89-P L-mode scaling, HITER89-p. Longer ELM-free periods became possible, and HITER89-p = 2.5 was achieved transiently, but in this mode of operation impurity accumulation led to eventual termination of H-mode. More favorable high quality quasi-steady-state H-modes which led to steady-state levels of density (n-bare = 4 x 1020 m-3), stored energy (HITER89-p = 2.0, βN = 1.5), and radiated power (Pmainrad/Ploss ≤ 30%) were also achieved. This mode of operation is characterized by high levels of continuous Dα emission comparable to the L-mode level, with very little or no ELM activity. The H-mode behavior was affected by controlling the neutral pressure by wall conditioning and gas puffing. (author). 18 refs, 7 figs

  19. Radial impurity transport in the H mode transport barrier region in Alcator C-Mod

    International Nuclear Information System (INIS)

    Measurements of profiles of soft X ray emissivity with 1.5 mm radial resolution are combined with high resolution electron density and temperature measurements in the edge region of the Alcator C-Mod tokamak to facilitate transport analysis of medium-Z impurities during H modes. Results from detailed modelling of the radiation and transport of fluorine are compared with experimental measurements, yielding information about the transport coefficients in the H mode transport barrier region. Evidence is found for a strong inward impurity pinch just inside the separatrix. The region of strong inward pinch agrees very well with the region of strong electron density gradient, suggesting that the inward pinch could be driven by the ion density gradient, as predicted by neoclassical theory. Simulations using the neoclassical impurity convection profile agree very well with experiments. Transport modelling shows that the X ray pedestal width is largely determined by the diffusion coefficient in the transport barrier. This allows diagnosis of changes in the edge diffusion coefficient on the basis of observations of X ray pedestal width changes. Significant differences in the edge diffusion coefficient are seen between different types of H mode. Several scalings for the edge diffusion coefficient in the enhanced Dα H mode are also identified. This may help elucidate the physical processes responsible for this attractive confinement mode. (author)

  20. 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.

  1. Imaging of X-point turbulence in Alcator C-Mod

    Science.gov (United States)

    Ballinger, Sean; Terry, James; White, Anne; Zweben, Stewart

    2015-11-01

    A nearly tangential view of the lower X-point region of Alcator C-Mod has been coupled to a high-speed camera filtered for D-alpha line emission. Recording at ~400,000 frames per second, the system detects filaments propagating in the private flux region that are approximately aligned with the local magnetic field. This behavior appears similar to what has recently been observed in the MAST tokamak. Turbulence and transport into the private flux region is potentially important. It may be a mechanism to spread heat across field lines and reduce peak heat fluxes on divertor targets. It may also explain how transport-driven flows seen in the high-field side scrape-off layer are accommodated, being otherwise too large compared to the particle flux arriving at the inner divertor target plates. The dynamics of these filaments are analyzed, as is the rate at which they are generated. Correlation analysis is used to determine the speed and trajectories of the filaments. Radial speeds of ~1 km/s are found. Clear changes are observed in the X-point-region fluctuations at the L-to-H-mode transition.

  2. 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.

  3. 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

  4. 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.

  5. Comparison of Scrape-off Layer Turbulence in Alcator C-Mod with Three Dimensional Gyrofluid Computations

    International Nuclear Information System (INIS)

    This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod (S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)) and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed.

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. 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)

  9. 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.

  10. First Measurements of Edge Transport Driven by the Shoelace Antenna on Alcator C-Mod

    Science.gov (United States)

    Golfinopoulos, T.; Labombard, B.; Parker, R. R.; Burke, W. M.; Hughes, J. W.; Brunner, D. F.; Davis, E. M.; Ennever, P. C.; Granetz, R. S.; Greenwald, M. J.; Irby, J. H.; Leccacorvi, R.; Marmar, E. S.; Parkin, W. C.; Porkolab, M.; Terry, J. L.; Vieira, R. F.; Wolfe, S. M.; Wukitch, S. J.; Alcator C-Mod Team

    2015-11-01

    The Shoelace antenna is a unique device designed to couple to the Quasi-Coherent Mode (QCM, k⊥ ~ 1 . 5 cm-1, 50 MLP) on the last-closed flux surface. This has enabled the first measurements of edge transport induced by the antenna-driven fluctuation, which has been further enhanced by quadrupling the antenna source power. This work was supported by U.S. Department of Energy award DE-FC02-99ER54512, using Alcator C-Mod, a DOE SC User Facility.

  11. Measurement of fast electron transport by lower hybrid modulation experiments in Alcator C-Mod

    OpenAIRE

    Schmidt, Andrea Elizabeth; Bonoli, Paul T.; R. Parker; Porkolab, Miklos; Wallace, Gregory Marriner; Wright, John C.; Wilson, J. R.; Harvey, R. W.; A. P. Smirnov

    2009-01-01

    The Lower Hybrid Current Drive (LHCD) system on Alcator C-Mod can produce spectra with a wide range of peak parallel refractive index (n||). An experiment in which LH power is square-wave modulated 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 ny spectra. Boxcar binning of hard x-rays during LH power modulation allows f...

  12. Design of a CO2-laser Thomson scattering ion-tail diagnostic for Alcator C-Mod

    International Nuclear Information System (INIS)

    A CO2-laser Thomson scattering diagnostic has been designed for the measurement of the ICRH-produced ion tail on Alcator C-Mod. The plasma parameters and port access require that the detection of scattered radiation be made at small angles, typically 1 degree or less. The receiver system consists of five heterodyne detectors and the source laser produces an energy of 10 J per pulse with a 1--5 μs pulse length. The scattering system is currently being installed on the Alcator C-Mod experiment. Details of the diagnostic, calculations of the expected measurements, and application of the diagnostic for ITER are presented

  13. 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

  14. 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.

  15. Turbulence imaging of spatiotemporal fluctuation structures in the SOL of Alcator C-Mod

    International Nuclear Information System (INIS)

    It is well known that the transport of plasma particles and energy across the magnetic field is strongly related to turbulent fluctuations, the fluctuation-induced transport. In the last few years a contribution to cross-field transport by direct radial propagation of turbulent structures, the so-called blobs, has been discovered. Although blobs appear as intermittent events and occupy only a minor fraction of the fluctuations, they significantly contribute to the fluctuation induced transport, thereby playing a crucial role in the confinement quality of fusion devices and are important for divertor concepts and first wall recycling phenomena. Blobs are localized in the poloidal plane but form striations over long distances along the magnetic field k(par.)/k(pen.) << 1. The present paper reports on investigation of the propagation properties of structures in the edge and SOL of the Alcator C-Mod tokamak. The main diagnostic tool is turbulence imaging diagnostics. It consists of an ultra fast camera system (250-kHz frame rate) measuring the Dα emission of a localized gas puff in the poloidal plane covering the SOL and edge plasma region. This diagnostics provides real-time spatiotemporal information about the formation, propagation, and decay of turbulent density structures. It is observed that after formation close to the separatrix the propagation of turbulent structures is not purely poloidal but has a strong radial component with a typical radial propagation speed of 2-5% of the ion sound speed, thereby causing large cross-field transport events. The combination of fast Dα diode arrays (1 MHz sampling frequency) and reciprocating multi-tip Langmuir probes yields that the potential associated with turbulent density structures is of dipole shape. The associated radial drift is consistent with the observed propagation velocity of blobs. The experimental findings are compared to basic models for blob propagation and to numerical simulation results. Basic

  16. Design and Commissioning of a Novel LHCD Launcher on Alcator C-Mod

    International Nuclear Information System (INIS)

    Full text: The goal of the lower hybrid current drive (LHCD) experiment on Alcator C-Mod is to demonstrate and study the full non-inductive high performance tokamak operation using parameters close to that envisioned for ITER in terms of LHCD frequency and magnetic field. Previously, up to 1.2 MW of net LHCD power at 4.6 GHz has been successfully launched for 0.5 s with a traditional grill launcher. To explore higher power and longer pulse regimes, a new LHCD launcher (LH2) was designed and fabricated. The new launcher is based on a novel four way splitter concept which evenly splits the microwave power in the poloidal direction. Sixteen splitters are stacked in the toroidal direction, creating the 16 columns and 4 rows of active waveguides. A total of 8 passive waveguides (one on each side of each row) are installed to reduce the reflection on the edge columns. This design allows the simplification of feeding structure 'jungle gym', while keeping the flexibility to vary the launched toroidal N|| spectrum from -3.8 to 3.8. To predict the LH2 performance, an integrated modeling using TOPLHA and CST microwave studio was carried out, in which the antenna-plasma coupling problem and the vacuum side EM problem were solved self-consistently. Good plasma coupling over a wide range of edge densities and a clean N|| spectrum were confirmed. The poloidal variation of edge density was found to affect mainly the evenness of power splitting, suggesting the necessity of the forward and reflected power measurements at each row of the launcher. 16 dedicated sets of RF probes were installed on a carefully selected set of active and passive waveguides to measure the forward and reflected power in each of these waveguides. In addition, the LH2 launcher is equipped with six Langmuir probes, and three X-mode reflectometer waveguides to measure the density profile in front of the launcher. Experiments using LH2 are scheduled to start in May. Initial results will be reported. Work

  17. Web based electronic logbook and experiment run database viewer for Alcator C-Mod

    International Nuclear Information System (INIS)

    Since 1991, the scientists and engineers at the Alcator C-Mod experiment at MIT have been recording text entries about the experiments being performed in an electronic logbook. In addition, separate documents such as run plans, run summaries and experimental proposals have been created and stored in a variety of formats in computer files. This information has now been organized and made available via any modern web browser. The new web based interface permits the user to browse through all the logbook entries, run information and even view some key data traces of the experiment. Since this information is being catalogued by Internet search engines, these tools can also be used to quickly locate information. The web based logbook and run information interface provides some additional capabilities. Once logged into the web site, users can add, delete or modify logbook entries directly from their browser. The logbook window on their browser also provides dynamic updating when any new logbook entries are made. There is also live C-Mod operation status information with optional audio announcements available. The user can receive the same state change announcements such as 'entering init' or 'entering pulse' as they would if they were sitting in the C-Mod control room. This paper will describe the functionality of the web based logbook and how it was implemented

  18. Observations of core toroidal rotation reversals in Alcator C-Mod ohmic L-mode plasmas

    International Nuclear Information System (INIS)

    Direction reversals of intrinsic toroidal rotation have been observed in Alcator C-Mod ohmic L-mode plasmas following modest electron density or toroidal magnetic field ramps. The reversal process occurs in the plasma interior, inside of the q = 3/2 surface. For low density plasmas, the rotation is in the co-current direction, and can reverse to the counter-current direction following an increase in the electron density above a certain threshold. Reversals from the co- to counter-current direction are correlated with a sharp decrease in density fluctuations with kR ≥ 2 cm-1 and with frequencies above 70 kHz. The density at which the rotation reverses increases linearly with plasma current, and decreases with increasing magnetic field. There is a strong correlation between the reversal density and the density at which the global ohmic L-mode energy confinement changes from the linear to the saturated regime.

  19. Design and construction of thyristor DC circuit breakers for alcator C-MOD

    International Nuclear Information System (INIS)

    This paper reports on the detailed design of the Alcator C-MOD thyristor DC circuit breakers which required extensive simulation and careful consideration of parasitic elements as well as device recovery and turn-on limitations. The circuit uses up to 12 parallel paths to carry pulsed currents as high as 50 kA. Maximum interrupting voltages of 2 kV are supported by devices rated 4.4 kV. Each SCR is shunted by a diode in series with an air core decoupling reactor. The counterpulse capacitor is discharged though 1 or 2 SCRs and utilizes a pulse forming network to quickly reduce the main SCR current to zero, then provide an relatively low and stable reverse voltage during the commutation interval. All devices are mounted with radial symmetry. The diode reactors are subject to modest forces under normal conditions but large lateral forces when adjacent paths are lost

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D. R., E-mail: dmikkelsen@pppl.gov; Bitter, M.; Delgado-Aparicio, L.; Hill, K. W. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543 (United States); Greenwald, M.; Howard, N. T.; Hughes, J. W.; Rice, J. E. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); Reinke, M. L. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Podpaly, Y. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); AAAS S and T Fellow placed in the Directorate for Engineering, NSF, 4201 Wilson Blvd., Arlington, Virginia 22230 (United States); Ma, Y. [MIT Plasma Science and Fusion Center, 175 Albany St., Cambridge, Massachusetts 02139 (United States); ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Candy, J.; Waltz, R. E. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2015-06-15

    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.

  2. 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.

  3. 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.

  4. Comparison of Scrape-off Layer Turbulence in Alcator C-Mod with Three Dimensional Gyrofluid Computations

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S. J.; Scott, B. D.; Terry, J. L.; LaBombard, B.; Hughes, J. W.; Stotler, D. P.

    2009-09-01

    This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod [S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)] and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed. 2009 American Institute of Physics.

  5. Validation of full-wave simulations for mode conversion of waves in the ion cyclotron range of frequencies with phase contrast imaging in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Tsujii, N., E-mail: tsujii@k.u-tokyo.ac.jp [Graduate School of Frontier Sciences, The University of Tokyo, Chiba 277-8561 (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 02139 (United States); Jaeger, E. F. [XCEL Engineering, Inc., Oak Ridge, Tennessee 37830 (United States); Green, D. L. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Harvey, R. W. [CompX, Del Mar, California 92014 (United States)

    2015-08-15

    Mode conversion of fast waves in the ion cyclotron range of frequencies (ICRF) is known to result in current drive and flow drive under optimised conditions, which may be utilized to control plasma profiles and improve fusion plasma performance. To describe these processes accurately in a realistic toroidal geometry, numerical simulations are essential. Quantitative comparison of these simulations and the actual experimental measurements is important to validate their predictions and to evaluate their limitations. The phase contrast imaging (PCI) diagnostic has been used to directly detect the ICRF waves in the Alcator C-Mod tokamak. The measurements have been compared with full-wave simulations through a synthetic diagnostic technique. Recently, the frequency response of the PCI detector array on Alcator C-Mod was recalibrated, which greatly improved the comparison between the measurements and the simulations. In this study, mode converted waves for D-{sup 3}He and D-H plasmas with various ion species compositions were re-analyzed with the new calibration. For the minority heating cases, self-consistent electric fields and a minority ion distribution function were simulated by iterating a full-wave code and a Fokker-Planck code. The simulated mode converted wave intensity was in quite reasonable agreement with the measurements close to the antenna, but discrepancies remain for comparison at larger distances.

  6. Validation of full-wave simulations for mode conversion of waves in the ion cyclotron range of frequencies with phase contrast imaging in Alcator C-Mod

    International Nuclear Information System (INIS)

    Mode conversion of fast waves in the ion cyclotron range of frequencies (ICRF) is known to result in current drive and flow drive under optimised conditions, which may be utilized to control plasma profiles and improve fusion plasma performance. To describe these processes accurately in a realistic toroidal geometry, numerical simulations are essential. Quantitative comparison of these simulations and the actual experimental measurements is important to validate their predictions and to evaluate their limitations. The phase contrast imaging (PCI) diagnostic has been used to directly detect the ICRF waves in the Alcator C-Mod tokamak. The measurements have been compared with full-wave simulations through a synthetic diagnostic technique. Recently, the frequency response of the PCI detector array on Alcator C-Mod was recalibrated, which greatly improved the comparison between the measurements and the simulations. In this study, mode converted waves for D-3He and D-H plasmas with various ion species compositions were re-analyzed with the new calibration. For the minority heating cases, self-consistent electric fields and a minority ion distribution function were simulated by iterating a full-wave code and a Fokker-Planck code. The simulated mode converted wave intensity was in quite reasonable agreement with the measurements close to the antenna, but discrepancies remain for comparison at larger distances

  7. 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.

  8. 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.

  9. Formation and evolution of internal transport barriers in Alcator C-Mod

    International Nuclear Information System (INIS)

    Central fueling of Alcator C-Mod plasmas with lithium and deuterium pellets often leads to a strong reduction of core energy and particle transport. These transient modes, which typically persist for a few energy confinement times, are characterized by the development, during the post-pellet reheat, of a very steep pressure gradient (scale length Ip ≤ a/5) in the inner third of the plasma. Inside the transport barrier, the ion thermal diffusivity drops to values close to those predicted from neoclassical theory. The global energy confinement time shows an increase of about 30% relative to L-mode scaling. Sawtooth suppression is typical, but is not observed in all cases. The addition of up to 3 MW of ICRF auxiliary heating, shortly after the pellet injection, leads to high fusion reactivity, with D-D neutron rates enhanced by a factor of about 10 over L-mode discharges with similar input powers. The measured current density profile shows that a region of reversed magnetic shear exist at the plasma core. The change in current profile is consistent with the calculated bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasma are near both the n = ∞ and the n = 1 marginal stability limits. (author). 12 refs, 5 figs

  10. Internal Transport Barriers in Alcator C-Mod Ohmic H-Mode Plasmas

    Science.gov (United States)

    Wolfe, S. M.; Fiore, C. L.; Bonoli, P. T.; Greenwald, M. J.; Hubbard, A. E.; Marmar, E. S.; Rice, J. E.; Wukitch, S. J.; Redi, M.

    2002-11-01

    Ohmic H-mode operation in Alcator C-Mod is often accompanied by a spontaneous peaking of the central density. The resulting electron density profile shows a ``foot'' at an r/a = 0.5, similar to most double transport barrier modes in this device. Analysis of the transport in these plasmas shows a reduction of the core thermal transport, decline of central toroidal rotataion velocity, and decrease of ηe in the barrier region. This is similar to the ITBs that are induced using off axis ICRF injection.(C. L. Fiore, et al., Phys. Plasmas), 8 2023.^,(S. J. Wukitch, et al., Phys. Plasmas), 9 2149.^,(J.E. Rice, et al., Nuclear Fusion), 42 510. The results of transport and stability code analysis of these plasmas will be presented. Recent experiments have been done to test the response and the performance of such Ohmic H-mode ITBs under the addition of varying amounts of central rf power. The results of these experiments will be also be presented.

  11. 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.

  12. Comparison of edge turbulence imaging at two different poloidal locations in the scrape-off layer of Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S. J.; Davis, W. M.; Diallo, A.; Ellis, R. A.; Stotler, D. P. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Terry, J. L.; Golfinopoulos, T.; Hughes, J. W.; LaBombard, B.; Landreman, M. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Agostini, M. [Consorzio RFX, Associazione Euratom-ENEA sulla fusione, C.so Stati Uniti 4, I-3512 Padova (Italy); Grulke, O. [Max Planck Institute for Plasma Physics, EURATOM Association, D-17491 Greifswald (Germany); Myra, J. R. [Lodestar Research Corporation, 2400 Central Ave., Boulder, Colorado 80301 (United States); Pace, D. C. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2013-07-15

    This paper describes 2D imaging measurements of plasma turbulence made in the scrape-off layer of the Alcator C-Mod tokamak simultaneously at two different poloidal locations, one near the outer midplane and the other near the divertor X-point region. These images were made with radial and poloidal resolution using two gas puff imaging diagnostics not directly connected along a B field line. The turbulence correlation structure has a significantly different tilt angle with respect to the local flux surfaces for the midplane and X-regions, and a slightly different ellipticity and size. The time-averaged turbulence velocities can be different in the midplane and X-regions, even within the same flux surface in the same shot. The structures are partially consistent with a magnetic flux tube mapping model, and the velocities are compared with various models for turbulence flow.

  13. Measurements of LHCD current profile and efficiency for simulation validation on Alcator C-Mod

    Science.gov (United States)

    Mumgaard, Robert T.

    2014-10-01

    Lower hybrid current drive (LHCD) is an effective tool to significantly modify the magnetic equilibrium by driving off-axis, non-inductive current. On Alcator C-Mod, an upgraded Motional Stark Effect (MSE) diagnostic enables the current profile to be accurately reconstructed during plasmas with strong LHCD and a hard X-ray camera measures the fast electron Bremsstrahlung profile. LHCD is applied for >4 current relaxation times, producing fully-relaxed magnetic equilibria in plasmas with non-inductive current fraction up to unity at currents up to 1.0 MA. C-Mod has developed an extensive database of LHCD performance, spanning a wide range in plasma current, launched n||, LHCD power, Te and plasma density. This dataset provides a unique platform for validation of LHCD current drive simulations with the plasma shape, density, field and LH frequency range envisioned for ITER and future reactors. In these conditions the measured current drive efficiencies are similar to that assumed for ITER with values up to 0.4*1020A/Wm2 despite being in a weak single-pass absorption regime. The driven current is observed to be off-axis, broadening the current profile, raising q0 above 1, suppressing sawteeth, decreasing/reversing the magnetic shear and sometimes destabilizing MHD modes and/or triggering internal transport barriers. Measurements indicate increased efficiency at increased temperature and plasma current but with a complicated dependence on launched n||. The MSE-constrained reconstructions show a loss in current drive efficiency as the plasma density is increased above =1.0×1020 m-3 consistent with previous observations of a precipitous drop in hard x-ray emission. Additionally, the measured driven current profile moves radially outward as the density is increased. Ray tracing simulations using GENRAY-CQL3D qualitatively reproduce these trends showing the rays make many passes through the plasma at high density and predicting a narrower current and HXR profile with than

  14. H-mode regimes and observations of central toroidal rotation in ALCATOR C-mod

    International Nuclear Information System (INIS)

    The Enhanced Dα or EDA H-mode regime in Alcator C-Mod has been investigated and compared in detail to ELM-free plasmas. (In this paper, ELM-free will refer to discharges with no type I ELMs and with no sign of EDA, though technically, most EDA plasmas are ELM-free as well.) 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. EDA plasmas are seen to be more likely at low plasma current (q>3.7-4), for moderate plasma shaping, (0.35-0.55), and for high neutral pressures. No obvious trends were observed with input power or pressure (β). In both H-mode regimes, and in ICRF heated L-modes, central impurity toroidal rotation has been deduced, from the Doppler shifts of argon x-ray lines. Rotation velocities up to 1.3x105 m/s in the co-current direction have been observed in H-mode discharges that had no direct momentum input. There is a strong correlation between the increase in the central impurity rotation velocity and the increase in the plasma stored energy, induced by ICRF heating. In otherwise similar discharges with the same stored energy increase, plasmas with lower current rotate faster. The ion pressure gradient is an unimportant contributor to the central impurity rotation and the presence of a substantial core radial electric field is inferred during the ICRF pulse. An inward shift of ions induced by ICRF waves could give rise to a non-ambipolar electric field in the plasma core. Comparisons with a neo-classical ion orbit shift model show good agreement with the observations, both in magnitude, and in the scaling with plasma current. (author)

  15. Observations of counter-current toroidal rotation in Alcator C-Mod LHCD plasmas

    International Nuclear Information System (INIS)

    Following the application of lower hybrid current drive (LHCD) power, the core toroidal rotation in Alcator C-Mod L- and H-mode plasmas is found to increment in the counter-current direction, in conjunction with a decrease in the plasma internal inductance, li. Along with the drops in li and the core rotation velocity, there is peaking of the electron and impurity density profiles, as well as of the ion and electron temperature profiles. The mechanism generating the counter-current rotation is unknown, but it is consistent in sign with an inward shift of energetic electron orbits, giving rise to a negative core radial electric field. The peaking in the density, toroidal rotation (in the counter-current direction) and temperature profiles occurs over a time scale similar to the current relaxation time but slow compared with the energy and momentum confinement times. Most of these discharges exhibit sawtooth oscillations throughout, with the inversion radius shifting inward during the LHCD and profile evolution. The magnitudes of the changes in the internal inductance and the central rotation velocity are strongly correlated and found to increase with increasing LHCD power and decreasing electron density. The maximum effect is obtained with a waveguide phasing of 60 deg. (a launched parallel index of refraction n|| ∼ 1.5), with a significantly smaller magnitude at 120 deg. (n|| ∼ 3.1), and with no effect for negative or heating (180 deg.) phasing. Regardless of the plasma parameters and launched n|| of the waves, there is a strong correlation between the rotation velocity and li changes, possibly providing a clue for the underlying mechanism.

  16. Fluctuation statistics in the scrape-off layer of Alcator C-Mod

    Science.gov (United States)

    Kube, R.; Theodorsen, A.; Garcia, O. E.; LaBombard, B.; Terry, J. L.

    2016-05-01

    We study long time series of the ion saturation current and floating potential, sampled by Langmuir probes dwelled in the outboard mid-plane scrape-off layer and embedded in the lower divertor baffle of Alcator C-Mod. A series of ohmically heated L-mode plasma discharges is investigated with line-averaged plasma density ranging from {{\\bar{n}}\\text{e}}/{{n}\\text{G}}=0.15 to 0.42, where n G is the Greenwald density. All ion saturation current time series that are sampled in the far scrape-off layer are characterized by large-amplitude burst events. Coefficients of skewness and excess kurtosis of the time series obey a quadratic relationship and their histograms coincide partially upon proper normalization. Histograms of the ion saturation current time series are found to agree well with a prediction of a stochastic model for the particle density fluctuations in scrape-off layer plasmas. The distribution of the waiting times between successive large-amplitude burst events and of the burst amplitudes are approximately described by exponential distributions. The average waiting time and burst amplitude are found to vary weakly with the line-averaged plasma density. Conditional averaging reveals that the radial blob velocity, estimated from floating potential measurements, increases with the normalized burst amplitude in the outboard mid-plane scrape-off layer. For low density discharges, the conditionally averaged waveform of the floating potential associated with large amplitude bursts at the divertor probes has a dipolar shape. In detached divertor conditions the average waveform is random, indicating electrical disconnection of blobs from the sheaths at the divertor targets.

  17. 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

  18. 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

  19. Mechanisms for ITB formation and control in Alcator C-Mod identified through gyrokinetic simulations of TEM turbulence

    International Nuclear Information System (INIS)

    Internal particle and thermal energy transport barriers are produced in Alcator C-Mod with off-axis ICRF heating, with core densities exceeding 1021 m-3, without core fueling, and with little change in the temperature profile. Applying on-axis ICRF heating controls the core density gradient and rate of rise. The present study employs linear and nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence to explore mechanisms for ITB formation and control in Alcator C-Mod ITB experiments. Anomalous pinches are found to be negligible in our simulations; further, the collisional Ware pinch is sufficient to account for the slow density rise, lasting many energy confinement times. The simulations have revealed new nonlinear physics of TEM turbulence. The critical density gradient for onset of TEM turbulent transport is nonlinearly up-shifted by zonal flows. As the density profile peaks, during ITB formation, this nonlinear critical gradient is eventually exceeded, and the turbulent particle diffusivity from GS2 gyrokinetic simulations matches the particle diffusivity from transport analysis, within experimental errors. A stable equilibrium is then established when the TEM turbulent diffusion balances the Ware pinch in the ITB. This equilibrium is sensitive to temperature through gyroBohm scaling of the TEM turbulent transport, and the collisionality dependence of the neoclassical pinch, providing for control of the density rate of rise with on-axis RF heating. With no core particle fueling, and ∼1 mm between density spatial channels, the C-Mod experiments provide a nearly ideal test bed for particle transport studies. The pure TEM is the only unstable drift mode in the ITB, producing particle transport driven by the density gradient. (author)

  20. 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.

  1. 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

  2. Design of off-midplane launcher (LH3) based on velocity space synergy for Alcator C-Mod

    International Nuclear Information System (INIS)

    A new LH launcher (LH3) is designed for Alcator C-Mod to increase the net LHCD power to 2 MW. With the existing launcher (LH2), LH3 aims to enhance the single pass power absorption to improve LHCD efficiency at high density. For this purpose, launcher design parameters are surveyed to maximize the synergistic effect between LH2. Ray-tracing is extensively used and the launcher location and the launched N∥ = c/v∥ are optimized. At the line averaged density of 1.4x1020 m3, it is predicted that the combination of two launchers can reduce the parasitic edge loss dramatically, and therefore can increase the LH driven current by about 50 - 60% compared to cases in which two launchers are used separately. Based on the parameter survey, an off-midplane launcher (LH3) is designed to be located 15 cm above mid-plane. LH3 employs 8-way power splitters to improve the antenna-plasma coupling. The antenna-plasma coupling simulation is performed using COMSOL and ALOHA codes, predicting 90% of RF power coupling when the launcher front density is high enough (≥ ncutoff). (author)

  3. Estimate of convective radial transport due to SOL turbulence as measured by GPI in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Zweben, S.J., E-mail: szweben@pppl.gov [PPPL, P.O. Box 451, Princeton, NJ 08540 (United States); Terry, J.L.; LaBombard, B. [PSFC, MIT, Cambridge, MA 02139 (United States); Agostini, M. [Consorzio RFX, Padova (Italy); Greenwald, M. [PSFC, MIT, Cambridge, MA 02139 (United States); Grulke, O. [IPP, Garching (Germany); Hughes, J.W. [PSFC, MIT, Cambridge, MA 02139 (United States); D' Ippolito, D.A. [Lodestar Research, Boulder, CO 80301 (United States); Krasheninnikov, S.I. [UCSD, San Diego, CA 92093 (United States); Myra, J.R.; Russell, D.A. [Lodestar Research, Boulder, CO 80301 (United States); Stotler, D.P. [PPPL, P.O. Box 451, Princeton, NJ 08540 (United States); Umansky, M. [LLNL, Livermore, CA 94550 (United States)

    2011-08-01

    The convective radial transport effects of SOL turbulence have been estimated using recent turbulence data from the gas puff imaging (GPI) camera diagnostic on Alcator C-Mod. The average radial turbulence speed within the region 1-2 cm outside the separatrix near the outer was calculated by a 2-D cross-correlation technique to be V{sub r} {approx} 0.2-0.3 km/s. Assuming this to be the local convective plasma velocity, the density SOL width {lambda}{sub n} was evaluated using a simple convective model to be {lambda}{sub n} {approx} 4-7 cm, which is {approx}2-3 times higher than that measured using a Langmuir probe. This convective velocity was also {approx}2-3 times lower than the velocities estimated from analytic blob models, but showed a similar scaling with plasma current at constant q{sub 95}. The measured blob speeds were lower than both the convective speeds and the analytic blob model speeds.

  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. Characterization of core and edge turbulence in L- and enhanced Dα H-mode Alcator C-Mod plasmas

    International Nuclear Information System (INIS)

    The recently upgraded phase-contrast imaging (PCI) diagnostic is used to characterize the transition from the low (L) to the enhanced Dα (EDA) high (H) confinement mode in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] plasmas. PCI yields information on line integrated density fluctuations along vertical chords. The number of channels has been increased from 12 to 32 and the sampling rate from 1 MHz to 10 MHz. This expansion of diagnostic capabilities is used to study broadband turbulence in L and EDA H mode and to analyze the quasicoherent (QC) mode associated with EDA H mode. Changes in broadband turbulence at the transition from L to EDA H mode can be interpreted as an effect of the Doppler rotation of the bulk plasma. Additional fluctuation measurements of Dα light and the poloidal magnetic field show features correlated with PCI in two different frequency ranges at the transition. The backtransition from EDA H to L mode, the so-called enhanced neutron (EN) mode, is investigated by new high frequency (132 and 140 GHz) reflectometer channels operating in the ordinary (O) mode. This additional hardware has been installed in an effort to study localized turbulence associated with internal transport barriers (ITBs). The EN mode is a suitable candidate for this study, since an ITB exists transiently as the outer density decreases much faster than the core density in this mode. The fact that the density decays from the outside inward allows us to study fluctuations progressing towards the plasma core. Our results mark the first localized observation of the QC mode at medium density: 2.2x1020 m-3 (132 GHz). Correlating the reflectometry measurements with other fluctuating quantities provides some insight regarding the causality of the EN-mode development

  7. Cross-field plasma transport and main-chamber recycling in diverted plasmas on Alcator C-Mod

    International Nuclear Information System (INIS)

    Cross-field particle transport increases sharply with distance into the SOL and plays a dominant role in the 'main-chamber recycling' regime in Alcator C-Mod, a regime in which most of the plasma particle efflux recycles on the main-chamber walls rather than flows into the divertor volume. This observation has potentially important implications for a reactor: contrary to the ideal picture of divertor operation, a tightly baffled divertor may not offer control of the neutral density in the main-chamber such that charge exchange heat losses and sputtering of the main-chamber walls can be reduced. The conditions that give rise to the main-chamber recycling regime can be understood by considering the plasma-neutral particle balance: when the flux surface averaged neutral density exceeds a critical value, flows to the divertor can no longer compete with the ionization source and particle fluxes must increase with distance into the SOL. This critical neutral density condition can be recast into a critical cross-field plasma flux condition: particle fluxes must increase with distance into the SOL when the plasma flux crossing a given flux surface exceeds a critical value. Thus, the existence of the main-chamber recycling regime is intrinsically tied to the level of anomalous cross-field particle transport. Direct measurement of the effective cross-field particle diffusivities Deff in a number of ohmic L mode discharges indicates that Deff near the separatrix strongly increases as plasma collisionality increases. Convected heat fluxes correspondingly increase, implying that there exists a critical plasma density (or perhaps collisionality) beyond which no steady state plasma can be maintained, even in the absence of radiation. (author)

  8. 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

  9. 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...

  10. 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.

  11. The dependence of core rotation on magnetic configuration and the relation to the H-mode power threshold in Alcator C-Mod plasmas with no momentum input

    International Nuclear Information System (INIS)

    The observed toroidal rotation in Alcator C-Mod Ohmic L-mode plasmas has been found to depend strongly on magnetic configuration. For standard discharges with a lower single null and the ion Bx∇B drift downward, and with BT = 5.4 T, IP = 0.8 MA and ne = 1.4 x 1020/m3, the core toroidal rotation is measured to be in the range of 10-20 km/s (counter-current). Similar plasmas with upper single null have significantly stronger counter-current rotation, in the range of 30-50 km/s. The rotation depends very sensitively on the distance between the primary and secondary separatrices in near double null plasmas, with changes of ∼25 km/s occurring over a variation of a few millimeters in this distance. Application of ICRF power has been found to increase the rotation in the co-current direction. The transition to H-mode is seen to occur in these standard plasmas when the core rotation reaches a characteristic value, near 0 km/s, hence higher input power is needed to induce the transition in the upper single null configuration. (author)

  12. Small ELM regimes with good confinement on JET and comparison to those on ASDEX Upgrade, Alcator C-mod, and JT-60U

    International Nuclear Information System (INIS)

    Since it is uncertain if ITER operation is compatible with type-I ELMs, the study of alternative H-mode pedestals is an urgent issue. This paper reports on experiments on JET aiming to find scenarios with small ELMs and good confinement, such as the type-II ELMs in ASDEX Upgrade, the enhanced D-alpha H-mode in Alcator C-mod or the grassy ELMs in JT-60U. The study includes shape variations, especially the closeness to a double-null configuration, variations of q95, density and beta poloidal. H-mode pedestals without type-I ELMs have been observed only at the lowest currents (≤ 1.2 MA), showing similarities to the observations in the devices mentioned above. These are discussed in detail on the basis of edge fluctuation analysis. For higher currents, only the mixed type-I/II scenario is observed. Although the increased inter-ELM transport reduces the type-I ELM frequency, a single type-I ELM is not significantly reduced in size. Obviously, these results do question the accessibility of such small ELM scenarios on ITER, except perhaps the high beta-poloidal scenario at higher q95, which could not be tested at higher currents at JET due to limitations in heating power. (author)

  13. Mean Flows and Blob Velocities in Scrape-Off Layer (SOLT) Simulations of an L-mode discharge on Alcator C-Mod

    Science.gov (United States)

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

    2015-11-01

    Two-dimensional scrape-off layer turbulence (SOLT) code simulations are compared with an L-mode discharge on Alcator C-Mod. Density and temperature profiles for the simulations were obtained by smoothly fitting Thomson scatter and mirror Langmuir probe (MLP) data from the shot. Simulations differing in turbulence intensity were obtained by varying a dissipation parameter. Mean flow profiles and density fluctuation amplitudes are consistent with those measured by MLP in the experiment. Blob velocities in the simulations were determined from the correlation function for density fluctuations, as in the analysis of gas-puff-imaging (GPI) blobs in the experiment. In the simulations, it was found that larger blobs moved poloidally with the ExB flow velocity, vE , in the near-SOL, while smaller fluctuations moved with the group velocity of the dominant linear (interchange) mode, vE + 1/2 vdi, where vdi is the ion diamagnetic drift velocity. Comparisons are made with the measured GPI correlation velocity. The saturation mechanisms operative in the simulation of the discharge are explored. Work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences, under Agreement DE-FC02-99ER54512, Contract DE-AC02-09CH11466, and Princeton Plasma Physics Laboratory Subcontract S013429-U.

  14. Non-local heat transport, rotation reversals and up/down impurity density asymmetries in Alcator C-Mod ohmic L-mode plasmas

    International Nuclear Information System (INIS)

    Several seemingly unrelated effects in Alcator C-Mod ohmic L-mode plasmas are shown to be closely connected: non-local heat transport, core toroidal rotation reversals, energy confinement saturation and up/down impurity density asymmetries. These phenomena all abruptly transform at a critical value of the collisionality. At low densities in the linear ohmic confinement regime, with collisionality ν* ⩽ 0.35 (evaluated inside of the q = 3/2 surface), heat transport exhibits non-local behaviour, core toroidal rotation is directed co-current, edge impurity density profiles are up/down symmetric and a turbulent feature in core density fluctuations with kθ up to 15 cm−1 (kθρs ∼ 1) is present. At high density/collisionality with saturated ohmic confinement, electron thermal transport is diffusive, core rotation is in the counter-current direction, edge impurity density profiles are up/down asymmetric and the high kθ turbulent feature is absent. The rotation reversal stagnation point (just inside of the q = 3/2 surface) coincides with the non-local electron temperature profile inversion radius. All of these observations suggest a possible unification in a model with trapped electron mode prevalence at low collisionality and ion temperature gradient mode domination at high collisionality. (paper)

  15. 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.

  16. 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

  17. 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

  18. Scaling of H-mode pedestal characteristics in DIII-D and C-Mod

    International Nuclear Information System (INIS)

    Since the H-mode edge pedestal effectively sets the boundary conditions for energy transport throughout the core, a better understanding of the pedestal region is necessary in order to fully predict H-mode performance. Pedestal characteristics in the DIII-D and Alcator C-Mod tokamaks are described, and scalings of the pedestal width with various plasma parameters are shown. The pedestal width in both tokamaks varies in an inverse sense with plasma current, and is independent of toroidal field. Other similarities, as well as differences, are discussed. It is also found that the pedestal widths of the various physical quantities involved (Te, Ti, ne, ni) may be different. (author)

  19. 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

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. Energy and impurity transport in the Alcator C tokamak

    International Nuclear Information System (INIS)

    Energy and impurity confinement times, tausub(E) and tausub(I), have been investigated in Ohmically heated Alcator C discharges in the following parameter range: density n-bar20 m-3, current Isub(p)20 m-3, the results of size-scaling experiments are summarized by the relation tausub(E) is proportional to asup(0.8) Rsup(2.3). At high density, the inferred ion thermal conductivity chisub(i) apparently exceeds the neoclassical value of Hinton and Hazeltine by a factor which varies from 2 to about 4 depending on geometry and plasma parameters. Other interpretations of the data are, however, possible. Impurity confinement has also been studied in these discharges using the laser blow-off technique. The principal new results concerning tausub(I) are (1) independence of the charge and mass of the injected impurity ion; (2) a linear dependence on asub(l); (3) a rapid deterioration as a function of the amplitude of the m=2,3 tearing modes; and (4) an approximately linear dependence on the effective charge. (author)

  6. 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.

  7. 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

  8. Ion and electron parameters in the alcator C tokamak scrape-off region

    International Nuclear Information System (INIS)

    Janus is a bi-directional, multi-functional edge probe used to diagnose the ion and electron parameters in the Alcator C tokamak scrape-off region. Two mirror image sets of diagnostics are aligned to face the electron and ion sides along magnetic field lines. Each set of diagnostics consists of a retarding-field energy analyzer (RFEA), a Langmuir probe, and a calorimeter. The RFEA can alternatively sample both the ion and electron parallel energy distribution functions during a tokamak discharge. From the Langmuir probe, one can infer electron temperature, density, and the plasma floating potential. Simple Langmuir probe theory is found to yield the best agreement between the measured Langmuir probe characteristics and the RFEA-inferred T/sub e/. The calorimeter independently detects the total parallel heat flux incident to an electrically floating plate. The measured sheath transmission coefficient, however, is typically lower than the theoretically predicted value by a factor of approx.3. Together these diagnostics enable detailed, localized edge plasma characterization on Alcator C

  9. 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.

  10. Studies of photonuclear neutron emission during the start-up phase of the Alcator C tokamak

    International Nuclear Information System (INIS)

    Alcator C operations commenced with discharge cleaning and tokamak operation using hydrogen filling gas. Prior to and during these experiments no deuterium gas was allowed into the device. The earliest operation resulted in dosimeter readings of a few Roentgen per shot in the vicinity of the limiter and a localized source of neutron emission of up to 109 neutrons/shot. A strong correlation of the neutron emissions with hard x-ray emissions from the limiter and nonthermal features on the synchrotron emissions was observed during these discharges. Gamma energy spectroscopy of the activated limiter after removal from Alcator allowed identification of 16 radioisotopes which were consistent with photonuclear processes (γ,n , γ,p , γ,α reactions) arising in the limiter. After seven months of hydrogen operation conditions were achieved that resulted in substantially less non-thermal activity. Typical neutron emission rates of equal to or less than 106 n/sec were observed, i.e., about four orders of magnitude less than the expected D-D thermonuclear neutron emission rates for the same type of discharge if D2 was used as the filling gas

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. 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/.

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. 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) greater than or equal to 8, and with rf powers up to 1 MW at 4.6 GHz. The temperature measurements from rf-driven discharges are compared with those from similar ohmic discharges (identical current and density). 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 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 1013 cm-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. To correctly simulate the experimentally observed temperatures, the electron thermal diffusivity must be increased with rf power and/or density relative to its ohmic values

  10. Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

    International Nuclear Information System (INIS)

    The emphasis of this year's ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod

  11. 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.

  12. 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.

  13. Quasilinear evolution of non-thermal distributions in ion cyclotron resonance heating of tokamak plasmas

    International Nuclear Information System (INIS)

    The AORSA global-wave solver is combined with the CQL3D bounce-averaged Fokker-Planck code to simulate the quasilinear evolution of non-thermal distributions in ion cyclotron resonance heating of tokamak plasmas. A novel re-formulation of the quasilinear operator enables calculation of the velocity space diffusion coefficients directly from the global wave fields. To obtain self-consistency between the wave fields and particle distribution function, AORSA and CQL3D have been iteratively coupled using Python. The combined selfconsistent model is applied to minority ion heating in the Alcator C-Mod tokamak. Results show the formation of a 70 keV ion tail near the minority ion cyclotron resonance layer in approximate agreement with measurements from charge exchange neutral particle analyzers

  14. 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.

  15. Neutron calibration techniques for comparison of tokamak results

    International Nuclear Information System (INIS)

    A workshop on 1--3 August 1989 reviewed the techniques, uncertainties, and experiences of neutron calibration on PLT, TFTR, JET, Tore Supra, JT-60, JIPPT-IIU, Alcator C-Mod, ATF, FT, ASDEX, Textor, and DIII-D. In the summary session, the workshop participants discussed possible consensus neutron calibration techniques appropriate to D-D plasmas in tokamaks. The application of such techniques would facilitate a more accurate comparison of neutron yields from different devices, and also allow new calibration techniques to relate their precision to a reference value. General agreement was reached on the suitability of two techniques: (1) a 252Cf source calibration of epithermal neutron detectors, and (2) threshold neutron activation of Ni foils placed vertically above or below the plasma. This paper will present details on detector positioning, neutron transport calculations, and interlab normalization needed to accomplish the standardized calibration using a Cf neutron source

  16. 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

  17. 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.

  18. Computer Modeling of the C-MOD Phase Contrast Imaging System

    Science.gov (United States)

    Shugart, A. J.; Hallock, G. A.; Rowan, W.; Mazurenko, A.; Nelson-Melby, E.; Wukitch, S. J.; Porkolab, M.

    2000-10-01

    Phase Contrast Imaging (PCI) is used on Alcator C-Mod to measure electron density fluctuations related to plasma turbulence, quasi-coherent modes, Ion-Cyclotron Range of Frequencies (ICRF) and Ion-Bernstein waves (IBW). A simplified model of a PCI system has been developed using a computer code written in the Interactive Data Language (IDL). This code is implemented as a series of modules that form a small IDL library which models components of the optical setup. Using this library of routines one can account for the diffraction of light in both the near and far fields as well as focusing effects due to lenses and mirrors and the effects of the PCI phase plate. One can also determine the amplitude and phase of a cross section of the laser at any point in the optical system. We used this code to model PCI measurements of ICRF waves in C-Mod. The results of this analysis show the sensitivity and resolution of the system.

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

    International Nuclear Information System (INIS)

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

  20. 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.

  1. Flutter interpretation of thermal conduction with application to Alcator

    International Nuclear Information System (INIS)

    A previously presented magnetic fluctuation interpretation of anomalous thermal conduction in Tokamaks is extended to the high field, high density regime of Alcator. Density fluctuations are fully taken into account but found not to significantly alter the predictions. If the electron parallel thermal conductivity is corrected for high Knudsen number effects, it is found that the mean squared relative magnetic fluctuation level must scale as (density)-2, to fit Alcator scaling. (author)

  2. 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.

  3. Alcator C vertical viewing electron cyclotron emission diagnostic

    International Nuclear Information System (INIS)

    Electron cyclotron emission measured vertically through the center of a tokamak plasma yields detailed information about the electron velocity distribution. A diagnostic developed for this purpose on Alcator C tokamak uses specialized focusing optics to obtain a well collimated viewing chord, a compact viewing dump made of pyrex or Macor to reduce the effects of wall reflection and depolarization, and a rapid-scan polarizing Michelson interferometer - InSb detector system for the spectrum measurement; all constrained by the limited access and the compact size of Alcator C. Results of diffraction analysis are used to evaluate the theoretical performance of the optical system

  4. 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.

  5. 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)

  6. Velocity fields of edge/Scrape-Off-Layer turbulence in Alcator C-Mod

    International Nuclear Information System (INIS)

    Using high speed movies of edge/Scrape-Off-Layer emission, velocity fields of the turbulent fluctuations have been measured in a small region of the poloidal/radial plane at the plasma's outboard midplane. Gas-puff-imaging is used to localize the emission in the toroidal dimension. The measured D α line emission responds to the underlying s patio-temporal dynamics of density and temperature. Time-delay cross-correlation analysis is applied to the 64 x 64 pixel movie images and yields time-averaged velocity fields for the emission perturbations. The spatial resolution of the velocity determinations is a few mm over the viewed region. It is found that the velocities inside the separatrix are almost purely poloidal. In the Sol the radial velocity component is outward, and significant radial acceleration is observed. The poloidal and radial components of the SOL velocities vary with the discharge conditions. The magnitudes of all the velocity components measured this way range from 0 to ∼1000 m/s

  7. 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 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.

  8. 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)

  9. Measured MHD equilibrium in Alcator C

    International Nuclear Information System (INIS)

    A method of processing data from a set of partial Rogowski loops is developed to study the MHD equilibrium in Alcator C. Time dependent poloidal fields in the vicinity of the plasma are calculated from measured currents, with field penetration effects being accounted for. Fields from eddy currents induced by the plasma in the tokamak structure are estimated as well. Each of the set of twelve B/sub θ/ measurements can then be separated into a component from the plasma current and a component from currents external to the pickup loops. Harmonic solutions to Maxwell's equations in toroidal coordinates are fit to these measurements in order to infer the fields everywhere in the vacuum region surrounding the plasma. Using this diagnostic, plasma current, position, shape, and the Shafranov term Λ = β/sub p/ + l/sub i//2 - 1 may be computed, and systematic studies of these plasma parameters are undertaken for Alcator C plasmas

  10. 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.

  11. 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.

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

    Science.gov (United States)

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

    2016-05-01

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

  13. 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.

  14. Quantitative comparison of electron temperature fluctuations to nonlinear gyrokinetic simulations in C-Mod Ohmic L-mode discharges

    Science.gov (United States)

    Sung, C.; White, A. E.; Mikkelsen, D. R.; Greenwald, M.; Holland, C.; Howard, N. T.; Churchill, R.; Theiler, C.

    2016-04-01

    Long wavelength turbulent electron temperature fluctuations (kyρs 0.8) of Ohmic L-mode plasmas at Alcator C-Mod [E. S. Marmar et al., Nucl. Fusion 49, 104014 (2009)] with a correlation electron cyclotron emission diagnostic. The relative amplitude and frequency spectrum of the fluctuations are compared quantitatively with nonlinear gyrokinetic simulations using the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] in two different confinement regimes: linear Ohmic confinement (LOC) regime and saturated Ohmic confinement (SOC) regime. When comparing experiment with nonlinear simulations, it is found that local, electrostatic ion-scale simulations (kyρs ≲ 1.7) performed at r/a ˜ 0.85 reproduce the experimental ion heat flux levels, electron temperature fluctuation levels, and frequency spectra within experimental error bars. In contrast, the electron heat flux is robustly under-predicted and cannot be recovered by using scans of the simulation inputs within error bars or by using global simulations. If both the ion heat flux and the measured temperature fluctuations are attributed predominantly to long-wavelength turbulence, then under-prediction of electron heat flux strongly suggests that electron scale turbulence is important for transport in C-Mod Ohmic L-mode discharges. In addition, no evidence is found from linear or nonlinear simulations for a clear transition from trapped electron mode to ion temperature gradient turbulence across the LOC/SOC transition, and also there is no evidence in these Ohmic L-mode plasmas of the "Transport Shortfall" [C. Holland et al., Phys. Plasmas 16, 052301 (2009)].

  15. X-ray observations of 2l-nl' transitions and configuration-interaction effects from Kr, Mo, Nb and Zr in near neon-like charge states from tokamak plasmas

    International Nuclear Information System (INIS)

    X-ray spectra of 2l-3l' transitions in the elements krypton (Z = 36), zirconium (Z = 40), niobium (Z = 41) and molybdenum (Z = 42) from charge states around neon-like have been observed from Alcator C and Alcator C-Mod plasmas. Accurate wavelengths (±0.5 mA) have been determined with reference to neighbouring aluminium, silicon, sulfur, chlorine and argon lines with relatively well known wavelengths. Line identifications have been made by comparison to ab initio atomic structure calculations, using a fully relativistic, parametric potential code. Calculated wavelengths and oscillator strengths are presented for 2l-3l' transitions in neon-like ions and neighbouring Na-, Mg-, F- and O-like satellites. Electric quadrupole transitions in neon-like ions of the form 2s-3d and 2p-3p are found to be relatively bright, and their measured intensities agree well with the results of collisional-radiative modelling. The 2p-3s magnetic quadrupole transition is also relatively intense in the absence of collisional de-excitation at tokamak densities (1013-1015 cm-3). For krypton, transitions with upper levels up to n=9 have been observed, and the measured wavelengths and intensities have been compared with calculations. For some neon-like transitions with nearly degenerate upper levels, there can be substantial configuration interaction which alters the line intensities. (author)

  16. 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.

  17. 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.

  18. Fast wave scrape-off layer losses of tokamak plasmas in minority, mid/high harmonic, and helicon heating regimes

    International Nuclear Information System (INIS)

    This paper examines fast wave propagation and power loss in the scrape-off layer (SOL) of tokamak plasmas by using the full wave code AORSA, with the edge plasma beyond the last closed flux surface (LCFS) included in the solution domain and with a collisional damping parameter used as a proxy to represent the real, and most likely nonlinear, damping processes. In (1), 2D and 3D AORSA results for the low aspect ratio National Spherical Torus eXperiment (NSTX), show 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. This result is consistent with previous NSTX observations and it will be further verified in the upcoming NSTX-Upgrade (NSTX-U) experimental campaign. Here, full wave simulations have been extended to 'conventional' tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show behavior similar to that found in NSTX and NSTX-U, and consistent with previous DIII-D experimental observations. In contrast, a different behavior is found for Alcator C-Mod and EAST, which unlike NSTX/NSTX-U and DIII-D that operate in the mid/high harmonic regime, operate in the minority heating regime. In the minority heating regime AORSA results indicate lower SOL power losses with increasing density in front of the antenna, in agreement with the experimental observation that increasing the density in front of the antenna leads to better antenna-plasma coupling. The effect of the pitch angle of the magnetic field and a comparison of the minority heating and mid/high harmonic heating regimes are presented. It is found that for NSTX-U scenarios the behavior of the RF field in the SOL region changes with plasma current. Finally, the impact of the SOL region on the evaluation of helicon current drive efficiency in DIII-D is presented and compared to the other heating regimes mentioned above. (author)

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. Neoclassical islands, β-limits, error fields, and ELMS in reactor-scale tokamaks

    International Nuclear Information System (INIS)

    An assessment is presented of the impact of recent magnetohydrodynamic research results on performance projections for reactor-scale tokamaks as exemplified by the ITER Final Design Report facility. For nominal ELMy H-mode operation, the presence and amplitude of neoclassical tearing modes governs the achievable β-value. Recent work finds that the scaling of β at which such modes onset agrees well with a polarization drift model, with the consequence that, with reasonable assumptions regarding seed-island width, the mode onset β will be lower in reactor-scale tokamaks than in contemporary devices. Confinement degradation by such modes, on the other hand, depends on relative saturated island size which is governed by principally by β and secondarily by ν*-effects on bootstrap current density. Relative saturated island size should be comparable in present and reactor devices. DT ITER Demonstration Discharges in JET exhibited no confinement degradation at the planned ITER operating value of βN =2.2. Theory indicates that Electron Cyclotron Current Drive can either stabilize these modes or appreciably reduce saturated island size. Turning to operation in candidate steady-state, reverse-shear, high-bootstrap-fraction configurations, wall stabilization of external kink modes is effective while the plasma is rotating but (so far) rotation has not been maintained. Recent error field observations in JET imply an error-field size scaling that leads to a projection that ITER/FDR will be somewhat more tolerant to error fields than thought previously. ICRF experiments on JET and Alcator C-Mod indicate that plasmas heated by central energetic particles have benign ELMs compared to the usual type 1 ELM of NBI-heated discharges. (author)

  7. Characterization of density fluctuations during the search for an I-mode regime on the DIII-D tokamak

    Science.gov (United States)

    Marinoni, A.; Rost, J. C.; Porkolab, M.; Hubbard, A. E.; Osborne, T. H.; White, A. E.; Whyte, D. G.; Rhodes, T. L.; Davis, E. M.; Ernst, D. R.; Burrell, K. H.

    2015-09-01

    The I-mode regime, routinely observed on the Alcator C-Mod tokamak, is characterized by an edge energy transport barrier without an accompanying particle barrier and with broadband instabilities, known as weakly coherent modes (WCM), believed to regulate particle transport at the edge. Recent experiments on the DIII-D tokamak exhibit I-mode characteristics in various physical quantities. These DIII-D plasmas evolve over long periods, lasting several energy confinement times, during which the edge electron temperature slowly evolves towards an H-mode-like profile, while maintaining a typical L-mode edge density profile. During these periods, referred to as I-mode phases, the radial electric field at the edge also gradually reaches values typically observed in H-mode. Density fluctuations measured with the phase contrast imaging diagnostic during I-mode phases exhibit three features typically observed in H-mode on DIII-D, although they develop progressively with time and without a sharp transition: the intensity of the fluctuations is reduced; the frequency spectrum is broadened and becomes non-monotonic; two dimensional space-time spectra appear to approach those in H-mode, showing phase velocities of density fluctuations at the edge increasing to about 10 km s-1. However, in DIII-D there is no clear evidence of the WCM. Preliminary linear gyro-kinetic simulations are performed in the pedestal region with the GS2 code and its recently upgraded model collision operator that conserves particles, energy and momentum. The increased bootstrap current and flow shear generated by the temperature pedestal are shown to decrease growth rates, thus possibly generating a feedback mechanism that progressively stabilizes fluctuations.

  8. Tokamak operation with high-Z plasma facing components

    International Nuclear Information System (INIS)

    , as observed by spectroscopic influx measurements of WI. The penetration of tungsten over the ETB is highly affected by ELMs, showing a clear reduction of the core tungsten content with rising ELM frequency. This effect is explained by the fact that the ELMs expel a major fraction of the tungsten ions which accumulate inside the ETB during the inter-ELM phase. Core neoclassical ion transport acts as a multiplier for the pedestal density, which depends on the temperature and density gradients in the core plasma. The ASDEX Upgrade results will be complemented by results obtained with molybdenum in Alcator C-Mod and with a tungsten limiter in TEXTOR. (Author)

  9. 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.

  10. 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.

  11. 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.

  12. Full-wave Electromagnetic Field Simulations of Lower Hybrid Waves in Tokamaks

    International Nuclear Information System (INIS)

    The most common method for treating wave propagation in tokamaks in the lower hybrid range of frequencies (LHRF) has been toroidal ray tracing, owing to the short wavelengths (relative to the system size) found in this regime. Although this technique provides an accurate description of 2D and 3D plasma inhomogeneity effects on wave propagation, the approach neglects important effects related to focusing, diffraction, and finite extent of the RF launcher. Also, the method breaks down at plasma cutoffs and caustics. Recent adaptation of full-wave electromagnetic field solvers to massively parallel computers has made it possible to accurately resolve wave phenomena in the LHRF. One such solver, the TORIC code, has been modified to simulate LH waves by implementing boundary conditions appropriate for coupling the fast electromagnetic and the slow electrostatic waves in the LHRF. In this frequency regime the plasma conductivity operator can be formulated in the limits of unmagnetized ions and strongly magnetized electrons, resulting in a relatively simple and explicit form. Simulations have been done for parameters typical of the planned LHRF experiments on Alcator C-Mod, demonstrating fully resolved fast and slow LH wave fields using a Maxwellian non-relativistic plasma dielectric. Significant spectral broadening of the injected wave spectrum and focusing of the wave fields have been found, especially at caustic surfaces. Comparisons with toroidal ray tracing have also been done and differences between the approaches have been found, especially for cases where wave caustics form. The possible role of this diffraction-induced spectral broadening in filling the spectral gap in LH heating and current drive will be discussed

  13. 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.

  14. 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

  15. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  16. 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.

  17. 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.

  18. 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.

  19. Extraordinary mode absorption at the electron cyclotron harmonic frequencies as a Tokamak plasma diagnostic

    International Nuclear Information System (INIS)

    Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission

  20. Investigation of Aditya Tokamak plasmas with lithiumized wall

    International Nuclear Information System (INIS)

    The lithium coating on plasma facing components of tokamak leads to better plasma properties through the reduction in impurities and controlling the hydrogen recycling. In Aditya tokamak, lithiumization of vacuum vessel wall is regularly carried out prior to its daily operation using lithium rod exposed to overnight glow discharge-cleaning plasma. Spectroscopic studies of Aditya tokamak plasmas shows the reduction of hydrogen (Hα at 656.3 nm) and oxygen (O II at 441.6 nm) as compared to discharges without the lithium coated walls. This clearly indicates reduction of recycling and impurity influxes from the wall, respectively. After Li coating, plasma stored energy increases significantly and plasmas with higher electron densities are obtained. Estimation of energy confinement time shows that it increases after lithimization and becomes comparable to the values predicated by Neo-Alcator scaling for ohmically heated tokamak plasma. Further analysis indicates that recycling must be low to achieve better plasma confinement. (author)

  1. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  2. Observation of self-generated flows in tokamak plasmas with lower-hybrid-driven current.

    Science.gov (United States)

    Ince-Cushman, A; Rice, J E; Reinke, M; Greenwald, M; Wallace, G; Parker, R; Fiore, C; Hughes, J W; Bonoli, P; Shiraiwa, S; Hubbard, A; Wolfe, S; Hutchinson, I H; Marmar, E; Bitter, M; Wilson, J; Hill, K

    2009-01-23

    In Alcator C-Mod discharges lower hybrid waves have been shown to induce a countercurrent change in toroidal rotation of up to 60 km/s in the central region of the plasma (r/a approximately current redistribution time (approximately 100 ms) but longer than the energy and momentum confinement times (approximately 20 ms). A comparison of the co- and countercurrent injected waves indicates that current drive (as opposed to heating) is responsible for the rotation profile modifications. Furthermore, the changes in central rotation velocity induced by lower hybrid current drive (LHCD) are well correlated with changes in normalized internal inductance. The application of LHCD has been shown to generate sheared rotation profiles and a negative increment in the radial electric field profile consistent with a fast electron pinch. PMID:19257362

  3. 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......). Two full wave codes, a massively-parallel-processor (MPP) version of the TORIC-2D finite Larmor radius code [M. Brambilla, Plasma Phys. Controlled Fusion 41, 1 (1999)] and also an all orders spectral code AORSA2D [E. F. Jaeger , Phys. Plasmas 9, 1873 (2002)], have been developed which for the first...... 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)] to...

  4. One-dimensional transport code modeling of the divertor-limiter region in tokamaks

    International Nuclear Information System (INIS)

    A model of the diverter-limiter scrapeoff region has been incorporated into the BALDUR one-dimensional tokamak transport code. Simulations of the proposed Toroidal Fusion Test Reactor (TFTR), and Poloidal Diverter (PDX) experiments and existing Alcator-A tokamak experiments have been carried out for ohmic and neutral beam heated cases. In particular it is studied how the edge conditions and energy-loss mechanisms in PDX depend upon plasma density, and results are compared with analytic estimates. The sensitivity of the results to changes in the transport coefficients and scrapeoff model is discussed with particular reference to the power loading on the TFTR limiter. 13 refs

  5. X-ray observations of 2l-nl' transitions in Mo30+--Mo33+ from tokamak plasmas

    International Nuclear Information System (INIS)

    X-ray spectra of 2p-nd transitions with 4≤n≤18 in molybdenum (Z=42) charge states around neonlike (Mo32+) have been observed from Alcator C-Mod plasmas. Accurate wavelengths (±0.1 mA) have been determined by comparison with neighboring argon and chlorine lines with well-known wavelengths. Line identifications have been made by comparison to ab initio atomic-structure calculations, using a fully relativistic, parametric-potential code, and agreement between measured and theoretical wavelengths is good. Calculated wavelengths and oscillator strengths are presented for the strongest transitions with upper levels n between 4 and 14 to the lower levels n=2 in the four charge states Mo30+--Mo33+. Effects of configuration interaction have been observed in the intensities of lines with nearly degenerate energy levels

  6. Testing of low Z coated limiters in tokamak fusion devices

    International Nuclear Information System (INIS)

    Extensive testing on a laboratory scale has been used to select those coatings most suitable for this environment. From this testing which included pulsed electron beam heating, low energy ion bombardment and arcing, chemical vapor deposited coating of TiB2 and TiC on Poco graphite substrates have been selected and tested as limiters in ISX. Both limiter materials gave clean, stable, reproducible tokamak discharges the first day of operation. After one weeks exposure, the TiC limiter showed only superficial damage with no coating failure. The TiB2 limiter had some small areas of coating failure. TiC coated graphite limiters have also been briefly tested in the tokamaks Alcator and PDX with favorable results

  7. 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.

  8. 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.

  9. 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.

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

    International Nuclear Information System (INIS)

    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

  11. 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

  12. THOR tokamak engineering design and experimental programme

    International Nuclear Information System (INIS)

    The THOR machine is an iron cored tokamak having a major radius of 0.52m and a minor radius of 0.17m giving an aspect ratio of 3.1. It has a low ripple toroidal field of 1T and a volt-second capability of 0.24. The maximum plasma current is expected to be in the region of 80 x 103A. Stabilisation of the plasma is achieved by means of a D.C. vertical field and a 1cm thick copper shell. The D.C. field is cancelled during the rise time of the plasma current by means of a pulsed reverse vertical field. Energy for the toroidal, ohmic heating and reverse vertical field systems is supplied from capacitors having a stored energy capability of 685kJ. The aims of the experimental programme include the control and study of the transition from the normal quasi-resistive tokamak regime to the low density slide-away condition. The non-Maxwellian character of the electron distribution function, typical of the slide-away regime, where strong emission around the ion plasma frequency and consequent ion heating have been observed (Alcator), will be studied by a combined transversal and tangential Thomson scattering experiment

  13. X-ray diagnostics of tokamak plasmas

    International Nuclear Information System (INIS)

    In this review, the authors venture into a new arena for work in atomic X-ray spectroscopy which can be dubbed tokamak X-ray spectroscopy diagnostics (TOXRASD). Not only is the experimental development exciting, but the measurements explore areas of atomic and plasma physics which have been inaccessible until just recently. Even though much of the present experimental effort is oriented towards obtaining a TOXRASD for fusion conditions, the new results touch upon some very basic atomic physics questions as well. Much effort has gone into the study of few electron systems, i.e., H-, He-, and Li-like ions with the purpose of utilizing the characteristic X-ray emission for diagnosing laboratory and astronomical plasmas or with the aim of developing the diagnostics, i.e., extending our understanding of the relationship between X-ray line emission and the plasma conditions under which the ions are formed and their ground states excited. However, the emission from highly charged heavy ions, such as neon-like molybdenum, has also attracted interest. Here the focus of the discussion is around results of the latest vintage from the TOXRASD project at the Alcator C tokamak at MIT. 38 references, 14 figures

  14. 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.))

  15. Massachusetts Institute of Technology, Plasma Fusion Center FY97--FY98 work proposal

    International Nuclear Information System (INIS)

    Alcator C-Mod is the high-field, high-density divertor tokamak in the world fusion program. It is one of five divertor experiments capable of plasma currents exceeding one megamp. Because of its compact dimensions, Alcator C-Mod investigates an essential area in parameter space, which complements the world's larger experiments, in establishing the tokamak physics database. Three key areas of investigation have been called out in which Alcator C-Mod has a vital role to play: (1) divertor research on C-Mod takes advantage of the advanced divertor shaping, the very high scrap-off-layer power density, unique abilities in impurity diagnosis, and the High-Z metal wall, to advance the physics understanding of this critical topic; (2) in transport studies, C-Mod is making critical tests of both empirical scalings and theoretically based interpretations of tokamak transport, at dimensional parameters that are unique but dimensionless parameters often comparable to those in much larger experiments; (3) in the area of Advanced Tokamak research, so important to concept optimization, the high-field design of the device also provides long pulse length, compared to resistive skin time, which provides an outstanding opportunity to investigate the extent to which enhanced confinement and stability can be sustained in steady-state, using active profile control. In addition to these main programmatic emphasis, important enabling research is being performed in MHD stability and control, which has great significance for the immediate design of ITER, and in the physics and engineering of ICRF, which is the main auxiliary heating method on C-Mod

  16. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab

  17. Varennes Tokamak

    International Nuclear Information System (INIS)

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  18. 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.

  19. Overview of the Microwave Tokamak Experiment operation and developments

    International Nuclear Information System (INIS)

    At Lawrence Livermore National Laboratory (LLNL), we assembled and presently operate the Microwave Tokamak Experiment (MTX) to demonstrate the feasibility of using intense microwave pulses (up to 6 GW peak power) from a free electron laser (FEL) to provide electron cyclotron heating (ECH) for use in tokamaks, particularly high field machines. The MTX consists primarily of the ALCATOR C tokamak and power supplies from MIT, along with FEL; the FEL is made up of the ETA-II linear induction accelerator and the IMP steady-state wiggler. A four-barrel pellet injector was added to the tokamak to produce peaked density profiles. The tokamak operations started in November 1988, with full duration plasmas being obtained at a toroidal field of both 5 and 9 tesla. Initial results were obtained with the single pulse 140 GHz FEL at peak power levels of 200 to 400 MW late in 1989. Due to excessive transverse electron beam motion, and arcing in the accelerator cells, the accelerator was modified. These modifications have been successfully tested on a small portion of the rebuilt accelerator and have been incorporated in the remaining portion of the accelerator. A 140 GHz, 400 kW gyrotron was used to perform preliminary heating experiments during the fall of 1990. This same gyrotron system is serving as the master oscillator for the burst mode FEL. The new IMP steady state wiggler will be used to produce the high power microwaves for the burst mode. The FEL construction has been completed, and it will be used for heating experiments scheduled for this fall. This paper describes the recent experimental operations. It also briefly outlines the additions and improvements to the experiment, which are described in more detail in companion papers at this conference

  20. Termoska pro tokamak

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2014-01-01

    Roč. 7, prosinec (2014), s. 16-17 Institutional support: RVO:61389021 Keywords : fusion * tokamak * cryostat * ITER Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1604-termoska-pro-tokamak

  1. 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

  2. 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

  3. PPPL tokamak program

    International Nuclear Information System (INIS)

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  4. 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....

  5. 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

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

    International Nuclear Information System (INIS)

    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 νφ 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

  7. A review of ELMs in divertor tokamaks

    International Nuclear Information System (INIS)

    This paper reviews what is known about edge localized modes (ELMs), with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M,JT-60U, and TCV tokamaks and are reported here. At least three types of ELMs have been identified and their salient features determined. Type-1 giant ELMs can cause the sudden loss of up to 10-15% of the plasma stored energy but their amplitude (ΔW/W) does not increase with increasing power. Type- 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1-3% of the stored energy). All ELMs increase the scrape- off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger that the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-Mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of Type-1 ELM amplitude and frequency have been measured in several tokamaks and has recently been applied to predictions of the ELM Size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with a very small type-3 ELMs can be maintained with heating power sell above the H-mode threshold, where giant type-1 ELMs can be maintained with heating power well above the H-mode threshold, where Giant type-1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H- mode discharges should not be overlooked

  8. 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

  9. 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

  10. Lower hybrid current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bare - 1020m-3, ALCATOR-C) and the highest current drive efficiency (ηCD = 3.5x1019 m-2A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

  11. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. 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 of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  12. 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

  13. 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

  14. 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.

  15. Ohmic discharges with improved confinement in Tokamak Aditya

    International Nuclear Information System (INIS)

    ADITYA (R0 = 75 cm, a = 25 cm), an ohmically heated circular limiter tokamak is regularly being operated to carry out several experiments related to controlled thermonuclear fusion research. In recent experimental schedule, special efforts are made to enhance the plasma parameters to achieve Ohmic discharges with improved confinement. Repeatable plasma discharges of maximum plasma current of ∼ 160 kA and discharge duration beyond ∼ 250 ms with plasma current flattop duration of ∼ 140 ms has been obtained for the first time in the first Indian tokamak ADITYA. The discharge reproducibility has been improved with Lithium wall conditioning and much-improved plasma discharges are obtained by precisely controlling the plasma position. Improved discharges are attempted over a wider parameter range to carry out various confinement scaling experiments. In these discharges, chord-averaged electron density 1.0 - 4.0 X 1019m-3 using multiple hydrogen gas puffs, plasma temperature of the order of ∼ 400 - 700 eV has been achieved. The measured confinement time matches quite well with ALCATOR scaling for most of the discharges. It is also observed that in new discharges, the confinement time crosses the L-mode scaling. Detailed analysis of these discharges along with the possible reasons for obtaining higher confinement times will be addressed in this paper. (author)

  16. 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

  17. Spectroscopic diagnostics of high temperature plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Moos, W.

    1990-01-01

    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 q{sub o} and local poloidal field measurements using Zeeman polarimetry.

  18. Spectroscopic diagnostics of high temperature plasmas. [Annual report

    Energy Technology Data Exchange (ETDEWEB)

    Moos, W.

    1990-12-31

    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 q{sub o} and local poloidal field measurements using Zeeman polarimetry.

  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. 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.

  1. 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

  2. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  3. 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)

  4. 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)

  5. 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). (1) The energy load associated with Type-I ELMs is of great concern for the lifetime of the ITER divertor target. In order to 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. (2) 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. (3) 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)

  6. The Thor tokamak experiment

    International Nuclear Information System (INIS)

    The main characteristics of the plasma produced in Thor tokamak discharges are described. The machine performances are outlined and the experimental results relevant to the equilibrium, the stability and the control of the discharge regimes are discussed in detail. (author)

  7. Modular tokamak magnetic system

    Science.gov (United States)

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  8. 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

  9. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs.

  10. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  11. Microwave Tokamak Experiment: An overview of the construction and checkout phase

    International Nuclear Information System (INIS)

    At Lawrence Livermore National Laboratory (LLNL) we constructed and presently operate the Microwave Tokamak Experiment (MTX) to demonstrate the feasibility of using microwave pulses produced from a free electron laser (FEL) to provide electron cyclotron heating (ECH) for use in tokamaks, particularly high-field machines. The MTX consists primarily of the ALCATOR C tokamak and power supplies that were documented and disassembled at the Massachusetts Institute of Technology (MIT) and shipped to LLNL in April 1987. We made many additions, including a new primary power system from the magnetic Fusion Test Facility (MFTF) substation, a new commutation system, substantially upgraded seismic support system for earthquake loading, a fast controls system for use with the FEL, a new data-acquisition system, and a new vault facility. We checked out these systems and put them into operation in October 1988; we achieved the first plasma in November 1988. We have also constructed and installed the microwave transmission system and the local microwave system to be used with the FEL. These systems transmit the microwaves to MTX quasi-optically through an evacuated tube. The ongoing plasma operations, both with and without FEL heating, are described in a companion paper. 12 refs., 2 figs., 2 tabs

  12. 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

  13. 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 co-ordinated 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 Co-ordinated Research Project is presented. (author)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. IFS Numerical Laboratory Tokamak

    International Nuclear Information System (INIS)

    A numerical laboratory of a tokamak plasma is being developed. This consists of the backbone (the overall manager in terms of the MPPL programming language), and the modularized components that can be plugged in or out for a particular run and their hierarchical arrangement. The components include various metrics for overall geometry various dynamics, field calculations, and diagnoses. 2 refs

  3. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  4. Transport in gyrokinetic tokamaks

    International Nuclear Information System (INIS)

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ''gyrokinetic tokamak'' is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/ρs approx-gt 64) with minor radius, with current, and with a/ρs are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (Nk ∼ 10) of k dominate the transport, and for each, only a handful (Np ∼ 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients Mkpq governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time

  5. 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

  6. 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

  7. 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

  8. 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.)

  9. 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)

  10. 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

  11. [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

  12. 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.

  13. Extraordinary mode absorption at the electron cyclotron harmonic frequencies as a tokamak electron temperature diagnostic

    International Nuclear Information System (INIS)

    An experimental study of extraordinary mode absorption at the semi-opaque second and third cyclotron harmonic frequencies has been performed on the Alcator C tokamak. A narrow beam of submillimetre laser radiation was used to illuminate the plasma in the horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission. Experimental electron cyclotron absorption (ECA) data have been found to agree with theoretical results for lowest significant order in finite density and finite Larmor radius. The ECA data have been used together with data for the tokamak electron density and magnetic field to determine local electron temperatures in the range 75 ≤ Te(eV) ≤ 3300, with a spatial resolution of ≤ 1 cm. Transmission during lower hybrid radiofrequency (LHRF) heating and current drive remained unaffected by suprathermal electrons because of their low density and the relativistic downshift of the absorbing harmonics away from the laser frequency. Therefore, this permitted the ECA technique to be used to measure the bulk plasma temperature during LHRF heating. A density dependent non-resonant attenuation was observed which, when present, was taken into account in the data analysis. (author)

  14. 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

  15. Polarization spectroscopy of tokamak plasmas

    International Nuclear Information System (INIS)

    Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs

  16. 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

  17. 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.

  18. 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

  19. Tore Supra tokamak

    International Nuclear Information System (INIS)

    This part of the electricity uses chapter of the Engineers Techniques collection is entirely devoted to the technical description of Tore Supra tokamak. A thermonuclear fusion device with magnetic confinement control such as Tore Supra concentrates a huge amount of high power electro-technical and electronic equipments. These power systems play a major role and are sometimes boosted to their extreme limits. From these equipments we can find: big superconducting magnets, big cooled copper magnets, high-voltage power supplies with thyristors (320 MVA installed), several MW hyper-frequency sources, several MW accelerated atom injectors, cryogenic, heat extraction, high-vacuum pumping systems, etc.. The components developed for these applications are numerous and frequently original: superconductor for variable magnetic field, DC static circuit breaker with high switch-off capability (0.7 GVA), 2 MW tetrodes, 500 kW klystrons, 500 kW gyrotrons, very low temperature (3 deg. K) electromechanical pumps, etc.. Tore Supra is a good example of the various applications of electricity and a testimony of the constant progress of the techniques mastered by electricians. This chapter is divided in 5 parts. Part 1 gives some general informations about thermonuclear fusion research, tokamak principles and electrotechnical systems of fusion research devices. Part 2 describes the Tore Supra tokamak, its aims and specificities, its internal components, the poloidal field system and the plasma heating systems. Part 3 concerns the power pulse sources: distribution network, poloidal field power supply, plasma heating systems, and ergodic divertor power supply. Part 4 describes the permanent electric power supplies for the auxiliary systems: toroidal field, cryogenic installation, cooling-drying loops. The last chapter briefly summarizes the perspectives of nuclear fusion research. (J.S.)

  20. 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

  1. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  2. Understanding disruptions in tokamaks

    International Nuclear Information System (INIS)

    This paper describes progress achieved since 2007 in understanding disruptions in tokamaks, when the effect of plasma current sharing with the wall was introduced into theory. As a result, the toroidal asymmetry of the plasma current measurements during vertical disruption event (VDE) on the Joint European Torus was explained. A new kind of plasma equilibria and mode coupling was introduced into theory, which can explain the duration of the external kink 1/1 mode during VDE. The paper presents first results of numerical simulations using a free boundary plasma model, relevant to disruptions.

  3. Tokamak instrumentation and controls

    Energy Technology Data Exchange (ETDEWEB)

    Becraft, W. R.; Bettis, E. S.; Houlberg, W. A.; Onega, R. J.; Stone, R. S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine.

  4. 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.

  5. 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.

  6. 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.)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. Bootstrap current in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kessel, C.E.

    1994-03-01

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

  12. Bootstrap current in a tokamak

    International Nuclear Information System (INIS)

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

  13. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] [and others]. E-mail: ludwig@plasma.inpe.br

    1999-07-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  14. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    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)

  15. 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...

  16. 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

  17. Stellarator - tokamak configurations

    International Nuclear Information System (INIS)

    The stellarator configuration and tokamak configuration with helical fields have been studied both from an equilibrium and stability point of view. The model was restricted to a surface current model with a sharp boundary between plasma and vacuum. A general derivation of equilibrium and stability based on the Energy Principle is given. Physically the unstable modes are identified as external global modes. Detailed numerical results in different parameter regimes are presented and discussed. Critical β-limits for equilibrium and stability are obtained and in particular it is shown that in certain parameter ranges there exist a high-β as well as a low-β-region of stability. 7 refs., 14 figs

  18. 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

  19. 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.

  20. 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.

  1. Magnetohydrodynamic instability, feedback stabilization, and disruption study for the Korea superconducting tokamak advanced research tokamak

    International Nuclear Information System (INIS)

    Passive and active feedback stabilization schemes being considered in Korea Superconducting Tokamak Advanced Research (KSTAR) device for the stabilization of the resistive magnetohydrodynamic modes such as the resistive wall and the neoclassical tearing are briefly introduced. A short summary is also presented on the tokamak simulation results of disruption dynamics and load in the KSTAR tokamak obtained using the tokamak simulation code (TSC)

  2. 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.

  3. High-energy-ion depletion in the charge exchange spectrum of Alcator C

    International Nuclear Information System (INIS)

    A three-dimensional, guiding center, Monte Carlo code is developed to study ion orbits in Alcator C. The highly peaked ripple of the magnetic field of Alcator is represented by an analytical expression for the vector potential. The analytical ripple field is compared to the resulting magnetic field generated by a current model of the toroidal plates; agreement is excellent. Ion-Ion scattering is simulated by a pitch angle and an energy scattering operator. The equations of motion are integrated with a variable time step, extrapolating integrator. The code produces collisionless banana and ripple trapped loss cones which agree well with present theory. Global energy distributions have been calculated and show a slight depletion above 8.5 keV. Particles which are ripple trapped and lost are at energies below where depletion is observed. It is found that ions pitch angle scatter less as energy is increased. The result is that, when viewed in velocity space, ions form probability lobes the shape of mouse ears which are fat near the thermal energy. Therefore, particles enter the loss cone at low energies near the bottom of the core. Recommendations for future work include improving the analytic model of the ripple field, testing the effect of del . B not equal to 0 on ion orbits, and improving the efficiency of the code by either using a spline fit for the magnetic fields or by creating a vectorized Monte Carlo code

  4. The upgradation of Aditya Tokamak

    International Nuclear Information System (INIS)

    Aditya Tokamak is the first Indian tokamak, indigenously built and commissioned at the Institute for Plasma Research, Gandhinagar, Gujarat, India, in September, 1989. Aditya Tokamak has been in operation since more than 25 years. More than 30,000 discharges are taken and a large number of experiments are carried out, with plasma current ranging from 50 KA to 150 KA, lasting for 100 to 250 milliseconds. Various types of wall conditioning techniques and different hot plasma diagnostics are tested and operated on Aditya Tokamak. The experiments for turbulent particle transport and turbulence in the edge plasma, gas puffing, lithium coating, mitigation, plasma disruption, limiter and electron biasing, runaway discharges etc. led to many interesting results contributing immensely to the world of thermonuclear fusion. Experiments on Pre-ionization and Plasma heating by ICRH and ECRH are also worked out. The scientific objectives of Aditya tokamak Upgrade include Low loop voltage plasma start-up with strong pre-ionization having a good plasma control system. The upgrade is designed keeping in mind the experiments, disruption mitigation studies relevant to future fusion devices, runway mitigation studies, demonstration of Radio-frequency heating and current drive etc. This upgraded Aditya tokamak will be used for basic studies on plasma confinement and scaling to larger devices, development and testing of new diagnostics etc. This machine will be easily accessible compared to SST-1 and will be very useful for generation of technical and scientific expertise for future fusion devices. In this paper, especial features of the upgrade including various aspects of designing of new components for Aditya Upgrade tokamak is presented

  5. 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

  6. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    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

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

  12. 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.)

  13. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  14. 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

  15. 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

  16. 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

  17. 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.

  18. Progress Report for C-Mod Collaboration

    International Nuclear Information System (INIS)

    The aims of the collaboration have not changed. The report describes progress in the areas of FRCECE system, charge exchange recombination spectroscopy, Beam-Emission spectroscopy (BES), as well as other contributions. A significant number of resulting publications are listed.

  19. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254. ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  20. RF preionization in Tokamak thor

    International Nuclear Information System (INIS)

    During the study of the RF preionization in Tokamak Thor was observed that the starting of the plasma and its time behaviour were correlated with the presence of resonance conditions both at the electron cyclotron frequency Ωsub(deg) and at its sub-harmonics Ωsub(deg)/n. These results are supported by a simple qualitative calculation

  1. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

  2. Edge plasma diagnostics in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Brotánková, Jana; Hron, Martin; Adámek, Jiří; Ďuran, Ivan; Van Oost, G.; Peleman, P.; Gunn, J.; Devynck, P.; Martines, E.; Schrittwieser, R.; Kocan, M.

    Kudowa Zdrój : -, 2006, s. 910-935. [Sixth International Workshop and Summer School Towards Fusion Energy - Plasma Physics, Diagnostics, Spin-offs. Kudowa Zdrój (PL), 18.09.2006-22.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * diagnostics * heating Subject RIV: BL - Plasma and Gas Discharge Physics

  3. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  4. System impedance and high-speed transient behavior of the Alcator-C magnet-power-supply system

    International Nuclear Information System (INIS)

    The commutation inductance is one of the parameters which determines the voltage regulation, ac bus notch depth, inversion margin angle and converter conduction mode for the TF power supply. It was felt that measuring the actual commutation inductance would provide valuable information on this parameter and a chance to become familiar with the details of the Alcator power supplies. Although initial work indicated that the commutation inductance varied during an Alcator experimental run, the measurements reported here show the inductance to be almost constant. While measuring the commutation inductance, two seemingly unrelated phenomena were observed; lightly damped ringing in the commutation notches and growing oscillations of the alternator voltage envelope. Theoretical models and simulations have been developed to explain the ringing and envelope oscillations. Two possible methods of damping the resonance (and thus eliminating the ringing and envelope oscillation) are proposed and the relative merits of each solution have been outlined

  5. 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)

  6. 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.

  7. Microwave Tokamak Experiment: Overview and status

    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. 3 figs., 3 tabs

  8. 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.

  9. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

  10. Frascati Tokamak transformer switching system

    International Nuclear Information System (INIS)

    Plasma ionization and heating, in the Frascati Tokamak, is obtained generating an emf along the plasma column, by switching the dc current flowing in the Tokamak transformer. 30 kA flowing in the 60 mH transformer inductance must be commutated into a resistance to generate 40 kV across the transformer itself. Studies and tests to solve this problem have been conducted, on different types of breakers, in cooperation between Tecnomasio Italiano Brown Boveri, Milan and Laboratori Gas Ionizzati, Frascati. Satisfactory results have finally been obtained using a DLF commercial air blast breaker in a chopper type circuit. A capacitor bank in parallel to the breaker is discharged immediately after the contacts separation and the arc in the switching element is extinguished at the first current zero. A saturable reactance in series with the breaker reduces the current decay rate to allow sufficient deionization time

  11. Burn Control Mechanisms in Tokamaks

    Science.gov (United States)

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  12. 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.

  13. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  14. 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

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

    International Nuclear Information System (INIS)

    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

  16. Quasilinear Evolution of Multiple Non-thermal Ion Distributions in ICRF Heating

    International Nuclear Information System (INIS)

    The AORSA global-wave solver is combined with the CQL3D bounce-averaged Fokker-Planck code to simulate the quasilinear evolution of non-thermal distributions in ion cyclotron resonance heating of tokamak plasmas. A novel re-formulation of the quasilinear operator enables calculation of the velocity space diffusion coefficients directly from the global wave fields. To obtain self-consistency between the wave fields and particle distribution function, AORSA and CQL3D have been iteratively coupled using Python. The combined self-consistent model is applied to minority ion heating in the Alcator C-Mod tokamak. Results show the formation of a 70 keV ion tail near the minority ion cyclotron resonance layer in approximate agreement with measurements from charge exchange neutral particle analyzers.

  17. Metrology measurements for Aditya tokamak upgradation

    International Nuclear Information System (INIS)

    After 25 years of Aditya tokamak (midsized, air-core, R0= 75 cm, a = 25 cm) operation achieving high temperature circular plasmas in limiter configuration, upgrading it to Aditya-U tokamak with divertor configuration has been planned and the upgradation is under progress. The upgradation process include dismantling of the existing Aditya tokamak to its base level and re-erect it by placing new subsystems like new vacuum vessel of circular cross-section, new buckling cylinder etc. Apposite alignment of subsystems, mainly all the magnetic coil systems in all grades and scales of tokamak is very crucial and essential, as misaligned magnetic coil system scan generate error magnetic fields, which can significantly impact the plasma formation and sustainment in a tokamak. With this motivation, position and alignment measurement of the existing magnetic coils and structural components of ADITYA tokamak is carried out for the very first time with the optical metrology instrument. Prior to carrying out measurement exercise, machine datum has been transferred to the reference on the wall of tokamak hall using five-point laser and the machine center has been transformed to the four wall of tokamak hall. All position measurements are done with respect to machine major axis in cylindrical geometry. Measurement includes existing radial (R) and elevation (Z) positions of all magnetic coils and various structural components within the accuracy of ± 1 mm. More than 5000 data points are recorded using optical metrology instrument. Again the elevation references are transferred to the primary network established and the angular references are transformed on the floor of the tokamak hall. These results will serve as ready reference for reassembly and alignment of Aditya - Upgrade tokamak. In this paper detailed position measurements of different subsystems of old Aditya tokamak and the relocation of them along with new ones using the optical metrology instruments will be presented

  18. Theory of high-beta tokamaks

    International Nuclear Information System (INIS)

    The theoretical researches on high beta tokamak are reviewed. The ballooning mode instability is thought to be the most serious problem for the high beta tokamaks, and the theoretical results on the ballooning mode instability are discussed in detail. The experimental results in high beta belt pinch devices are also discussed. (author)

  19. 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

  20. 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

  1. 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

  2. Physics of compact ignition tokamak designs

    International Nuclear Information System (INIS)

    Models for predicting plasma performance in compact ignition experiments are constructed on the basis of theoretical and empirical constraints and data from tokamak experiments. Emphasis is placed on finding transport and confinement models which reproduce results of both ohmically and auxiliary heated tokamak data. Illustrations of the application of the models to compact ignition designs are given

  3. 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

  4. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  5. Natural current profiles in a tokamak

    International Nuclear Information System (INIS)

    In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described

  6. Extending Spectroscopic Capabilities for Mo PFC Erosion Rate Diagnostics

    Science.gov (United States)

    Loch, S. D.; Ennis, D. A.; Pindzola, M. S.; Johnson, C. A.; Hartwell, G. J.; Maurer, D. A.; Griffin, D. C.; Ballance, C. P.; Reinke, M.; Lipschultz, B.; Soukhanovskii, V.

    2015-11-01

    The use of ionizations per photon coefficients (SXB) provides a useful means of measuring wall erosion rates. Two problems hindering the use of such diagnostics for high-Z materials are a lack of accurate atomic data and determining which lines from the complex spectral features should be used for accurate erosion measurements. We present a new approach for generating and selecting SXB coefficients for high-Z materials. The theoretical spectra show strong agreement with spectra from the Alcator C-Mod and Compact Toroidal Hybrid experiments. Mo II spectral features are identified, including a line ratio suitable for electron temperature measurements which constrains the SXB implementation. Applications of the new SXBs to NSTX-U edge plasmas is described and future plans for Mo and W influx diagnostics are outlined. Work supported by US DoE Cooperative agreement DE-FC02-99ER54512 at MIT using the Alcator C-Mod tokamak, a DOE Office of Science user facility. This work is also supported by US Department of Energy Grant No. DE-FG02-00ER54610 to Auburn University.

  7. Improved LHCD simulation model and implication for future experiments

    Science.gov (United States)

    Shiraiwa, S.; Wallace, G.; Baek, S.; Bonoli, P.; Faust, I.; Parker, R.; Labombard, B.; White, A.; Wukitch, S.

    2015-11-01

    The simulation model for LHCD using the raytracing/FokkerPlanck (GENRAY/CQL3D) code has been improved. Including realistic 2D SOL profiles resolves the discrepancy previously observed at high density (ne > 1 ×1020m-3). Impact of nonlinear interaction in front of the launcher is investigated. It is shown that the distortion of launch n| | spectrum is rather small (up to 10% of injected power). These simulation results suggest that improvement of current drive observed on Alcator C-Mod is indeed caused by realizing preferable SOL plasma profiles. Implication of these results to future experiments will be discussed. In order to minimize edge parasitic losses, realizing high single pass absorption and reducing prompt losses in front of launcher are both crucial. The advantage of LH launch from low field side (LFS) and high field side (HFS) is compared in this regards. A compact LH launcher suitable to test LH wave launch from HFS on a small scale device is designed and its plasma coupling characteristic will be presented. 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.

  8. Potential turbulence in tokamak plasmas

    International Nuclear Information System (INIS)

    Microscopic potential turbulence in tokamak plasmas are investigated by a multi-sample-volume heavy ion beam probe. The wavenumber/frequency spectra S(k,ω) of the plasmas potential fluctuation as well as density fluctuation are obtained for the first time. The instantaneous turbulence-driven particle flux, calculated from potential and density turbulence has oscillations of which amplitude is about 100 times larger than the steady-state outwards flux, showing sporadic behaviours. We also observed large-scale coherent potential oscillations with the frequency around 10-40 kHz. (author)

  9. The bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    The properties of the Hirshman equation for the bootstrap in the tokamak and the difference between it and the simpler Hinton-Hazeltine equation are discussed. The Hirshman model, which takes into account finite-aspect-ratio effects, is used to calculate the bootstrap current in the plasma in a circular cross section with Te = Ti. Approximate upper and lower bounds on the bootstrap current are obtained. These restrict the range of variation of the current as the temperature and density profiles vary. 16 refs., 9 figs

  10. 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

  11. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír

    2012-01-01

    Roč. 87, č. 12 (2012), s. 2238-2241. ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006

  12. Anomalous particle pinch in Tokamaks

    International Nuclear Information System (INIS)

    The diffusion coefficient in phase space usually varies with the particle energy. A consequence is the dependence of the fluid particle flux on the temperature gradient. If the diffusion coefficient in phase space decreases with the energy in the bulk of the thermal distribution function, the particle thermodiffusion coefficient which links the particle flux to the temperature gradient is negative. This is a possible explanation for the inward particle pinch that is observed in tokamaks. A quasilinear theory shows that such a thermodiffusion is generic for a tokamak electrostatic turbulence at low frequency. This effect adds to the particle flux associated with the radial gradient of magnetic field. This behavior is illustrated with a perturbed electric potential, for which the trajectories of charged particle guiding centers are calculated. The diffusion coefficient of particles is computed and compared to the quasilinear theory, which predicts a divergence at low velocity. It is shown that at low velocity, the actual diffusion coefficient increases, but remains lower than the quasilinear value. Nevertheless, this differential diffusion between cold and fast particles leads to an inward flux of particles. (author)

  13. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  14. Cluster storage for COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pisacka, J., E-mail: pisacka@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Hron, M., E-mail: hron@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University, V Holesovickach 2, 180 00 Praha 8 (Czech Republic); Panek, R., E-mail: panek@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer New data storage system needed for the COMPASS tokamak. Black-Right-Pointing-Pointer Distributed, fault-tolerant, parallel, scalable, non-proprietary. Black-Right-Pointing-Pointer GlusterFS selected for testing on a small test bed. Black-Right-Pointing-Pointer Aggregated reading throughput reached 300 MiB/s for 6 clients - very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  15. Cluster storage for COMPASS tokamak

    International Nuclear Information System (INIS)

    Highlights: ► New data storage system needed for the COMPASS tokamak. ► Distributed, fault-tolerant, parallel, scalable, non-proprietary. ► GlusterFS selected for testing on a small test bed. ► Aggregated reading throughput reached 300 MiB/s for 6 clients – very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  16. Predictive Modeling of Tokamak Configurations*

    Science.gov (United States)

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  17. Tokamak Physics Experiment divertor design

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  18. 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.)

  19. 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.

  20. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

  1. 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.

  2. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.)

  3. Power and particle exhaust in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER`s nominal design positions; important directions for further research are identified.

  4. 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.

  5. 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.)

  6. 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

  7. Synchrotron radiation in inhomogeneous tokamak plasmas

    International Nuclear Information System (INIS)

    Synchrotron emission in a tokamak configuration with inhomogeneous plasma parameters is considered to investigate the effects of the temperature profile and vertical elongation on the radiation loss. Using the numerical solution of the transfer equation for ITER-like plasma parameters, several new results on the radiated energy in a Maxwellian plasma have been derived. In particular: (i) synchrotron loss is profile dependent, namely, at constant average thermal energy, the emitted radiation increases with the peak temperature, (ii) an analytical formula of the global loss in inhomogeneous tokamak plasmas with arbitrary vertical elongation is established, (iii) the maximum of the frequency emission spectrum is a linear function of the volume average temperature, (iiii) high frequency synchrotron radiation is entirely due to electrons with energy much greater than the thermal energy. The need for experimental investigations on synchrotron emission in present-day large tokamaks to determine the effect of reflections of the complex tokamak first wall is stressed

  8. Edge plasma studies on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Hron, Martin; Peleman, P.; Spolaore, M.; Martines, E.; Hronová-Bilyková, Olena; Dejarnac, Renaud; Devynck, P.; Brotánková, Jana; Sentkerestiová, Jana; Ďuran, Ivan; Gunn, J.; Stöckel, Jan; Van Oost, G.; Adámek, Jiří; van de Peppel, L.; Štěpán, Michal

    Krakow : Euratom - IPPLM Association, 2006 - (Zagorski, R.), - [IEA Large Tokamak IA Workshop on Edge Transport in Fusion plasmas. Kraków (PL), 11.09.2006-13.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * scrape-off layer * turbulence * interchange instability Subject RIV: BL - Plasma and Gas Discharge Physics http://www.etfp2006.ifpilm.waw.pl/presentations.html

  9. 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)

  10. 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

  11. Plasma diagnostics using synchrotron radiation in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Fidone, I.; Giruzzi, G.; Granata, G.

    1995-09-01

    This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs.

  12. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aBtx of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  13. Epoxide insulation for Tokamak coils

    International Nuclear Information System (INIS)

    The construction and testing of 12-tonne toroidal-field electromagnets for the Joint European Torus by Brown Boveri and Cie (Mannheim) are described. The principle of Tokamak confinement of a plasma which acts as the secondary winding of a transformer is explained. The Cu conductors are sanded and coated with epoxide adhesive before being wrapped in 7mm thick woven glass fibre, dried by heating under vacuum, impregnated and encapsulated in 1.2 tonnes of Araldite, which is solidified under pressure of 4 atmospheres and hardened for ten hours at 1500C. The prototype withstood tests involving 25,000 flexure cycles at 1.1 MN and 2 Hz, 2,000 quarter-hour 10kA heating cycles between 840 and 200C, and exposure to 500 million rads. 32 such coils were constructed at the rate of one every three weeks. (M.B.D.)

  14. 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

  15. JT-60SA project for JA-EU broader approach satellite tokamak and national centralized tokamak

    International Nuclear Information System (INIS)

    JT-60 Super Advanced (JT-60SA) project is the joint project of ITER satellite tokamak by Japan and EU with Japanese Tokamak. The background, objects, device design, management of JT-60SA is stated. It consists of six chapters: the first chapter describes introduction, the second chapter states the objects of tokamak device complementing ITER, the third chapter contains research subjects and device performance such as plasma performance and demand for devices, operation scenario, control of MHD instability, and control of heat and particles, the forth chapter design of devices, the fifth chapter management and the sixth conclusion. In order to realize prototype reactor, improvement research of tokamak, development of reactor engineering technology, fusion reactor researches, tokamak theory and simulation, and social and environment safety research has to be advanced. (S.Y.)

  16. Particle and energy balances in tokamak plasmas

    International Nuclear Information System (INIS)

    Computational and experimental studies on particle and energy balances in tokamak plasmas are described. Firstly, concerning the modeling of tokamak plasmas, the particle balance considering diffusion and recycling, and the energy balance considering transport and energy losses due to impurities are discussed. Production mechanisms of gaseous and metallic impurities, which play important role in tokamak plasmas, are also discussed from a viewpoint of plasma-wall interactions. Scaling laws of density, temperature and energy confinement time are shown on the basis of recent data. Secondarily, tokamak plasmas are simulated with the above model, and anomalous diffusion and electron thermal conduction are indicated. Characteristics of a future tokamak plasma are also simulated. Stationary impurity density distributions and related energy losses, such as bremsstrahlung, ionization and excitation, are calculated taking into account diffusion and ionization processes. Edge cooling by oxygen impurities is described quantitatively compared with experiments. Permissible impurity levels of carbon, oxygen and iron in future large tokamaks are estimated. Thirdly, experimental studies on surface cleaning methods of the first wall are described; discharge cleaning in JFT-2, baking effect on the outgassing rates of wall materials, surface treatment of high-temperature molybdenum by oxygen and hydrogen gases, and in-situ coating of molybdenum by a coaxial magnetron sputter method. Lastly, problems in future large tokamaks aiming at break-even or self-ignited plasma are discussed quantitatively, such as trapped particle instabilities, impurities and additional heating. It is predicted that new conceptions will be necessary to overcome the problems and attain the fusion goal. (auth.)

  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. Plasma Physics Regimes in Tokamaks with Li Walls

    International Nuclear Information System (INIS)

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors

  19. Small tokamaks for fusion technology testing

    International Nuclear Information System (INIS)

    Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R0/a, H-D-fueled tokamaks with R0 ∼ 0.6--0.75 m, R0/a ∼ 1.8--2.5, and Bt0 ∼ 1.4--2.2 T can be driven with Ptot ∼ 4.5 MW to maintain Ip ∼ 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q∼1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R0 ∼ 0.8 m, R0/a ∼ 2, and Bt0 ∼ 4 T can be driven with Ptot ∼ 15 MW to maintain Ip ∼ 4.6 MA and produce an peak neutron wall load WL ∼ 1 MW/m2. Such devices appear possible if the plasma properties at the power R0/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%

  20. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  1. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region.

  2. 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)

  3. 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.)

  4. THOR tokamak magnetic field system

    International Nuclear Information System (INIS)

    The THOR Machine is an iron cored Tokamak having a major radius of 0.52 m and a minor radius of 0.17 m giving an aspect ratio of 3:1. It has a low ripple toroidal field of 1 T and an iron core giving 0.24 Vs. The maximum plasma current is expected to be in the region of 80x103 A. The maximum toroidal field ripple on axis is of the order of 0.01% and 2.5% at the plasma edge. The equilibrium of the plasma is achieved by means of a D.C. vertical field and a 1 cm thick copper shell. The D.C. field is cancelled during the rise time of the plasma current by means of pulsed reverse vertical field windings placed between the copper shell and the vacuum vessel. The design of this field system represents a compromise between obtaining adequate field penetration through the relatively thin vacuum vessel and maintaining the mechanical strength necessary to withstand the transient magnetic forces. Energy for the toroidal field system is supplied by a 15 kV 600 kJ capacitor bank and for the ohmic heating and reverse vertical fields by 5 kV 25 kJ and 50 kJ banks respectively. The problems encountered in the design, development and manufacture of these field systems are discussed. (author)

  5. Stability analysis of tokamak plasmas

    International Nuclear Information System (INIS)

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  6. 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.

  7. Tokamak x ray diagnostic instrumentation

    International Nuclear Information System (INIS)

    Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/ΔE is greater than 104 and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally

  8. Simulation of runaway electrons in Tokamak disruptions

    International Nuclear Information System (INIS)

    Self-consistent modelling of the generation of runaway electrons and the evolution of the toroidal electric field during tokamak disruptions is presented. The process of runaway generation is analysed by combining a relativistic kinetic equation for the electrons with Maxwell's equations for the electric field. Such modelling allows for a quantitative assessment of the runaway generation during disruptions in present day tokamak experiments, and to extrapolate to future tokamaks like ITER. It is found that the current profile can change dramatically during a disruption, such that the post disruption current, carried mainly by the runaway electrons, is significantly more peaked than the current profile before the disruption. In fact, it is found that the central current density can increase in spite of a reduction in the total current. (authors)

  9. Activation analysis of the compact ignition tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak.

  10. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

  11. 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)

  12. 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)

  13. Current drive by spheromak injection into a tokamak

    OpenAIRE

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

    1990-01-01

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

  14. Imaging with Spherically Bent Crystals or Reflectors

    International Nuclear Information System (INIS)

    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.

  15. 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

  16. 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.

  17. Neutral transport simulations of gas puff imaging experiments

    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

  18. 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.

  19. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs

  20. 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

  1. 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.

  2. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Full text: SST-1, a steady state superconducting tokamak, is at advanced stage of erection at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation and triangularity. The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 T at plasma center and a plasma current of 220 kA. Hydrogen gas will be used and plasma discharge duration will be 1000 s. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors having ports and 16 rings with D- shaped cross-section, which are welded in-situ during the SST-1 assembly. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as Sc magnets and cryostat, to minimize the radiation losses at the Sc magnets. In SST-1 tokamak, the auxiliary current drive will be based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. The assembly of the SST-1 tokamak is nearing completion. The cool down of the Superconducting magnets is scheduled to start by middle of year 2004

  3. 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.

  4. 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

  5. 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)

  6. 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)

  7. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

  8. 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

  9. Tokamak Spectroscopy for X-Ray Astronomy

    Science.gov (United States)

    Fournier, Kevin B.; Finkenthal, M.; Pacella, D.; May, M. J.; Soukhanovskii, V.; Mattioli, M.; Leigheb, M.; Rice, J. E.

    2000-01-01

    This paper presents the measured x-ray and Extreme Ultraviolet (XUV) spectra of three astrophysically abundant elements (Fe, Ca and Ne) from three different tokamak plasmas. In every case, each spectrum touches on an issue of atomic physics that is important for simulation codes to be used in the analysis of high spectral resolution data from current and future x-ray telescopes. The utility of the tokamak as a laboratory test bed for astrophysical data is demonstrated. Simple models generated with the HULLAC suite of codes demonstrate how the atomic physics issues studied can affect the interpretation of astrophysical data.

  10. 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

  11. Tokamak power systems studies, FY 1985

    International Nuclear Information System (INIS)

    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

  12. A need for non-tokamak approaches to magnetic fusion energy

    International Nuclear Information System (INIS)

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

  13. 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.

  14. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    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

  15. Microinstabilities in weak density gradient tokamak systems

    International Nuclear Information System (INIS)

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient

  16. High βp bootstrap tokamak reactor

    International Nuclear Information System (INIS)

    Basic characteristics of a steady state tokamak fusion reactor is presented. The minimum required energy multiplication factor Q is found to be 20 to 30 for the feasibility of the fusion reactor. Such a high Q steady state tokamak operation is possible, within our present knowledge of the operational constraints and the current drive physics, when a large fraction of the plasma current is carried by the bootstrap current. Operation at high βp (≥2.0) and high qψ (=4-5) with relatively small εβp (3) and fusion output power (2.5 GW) and is consistent with the present knowledges of the plasma physics of the tokamak, namely the Troyon limit, the energy confinement scalings, the bootstrap current, the current drive efficiency (NB current drive with the total power of 70 MW and the beam energy of 1 MeV) with a favorable aspect on the formation of the cold and dense diverter plasma-condition. From the economical aspect of the tokamak fusion reactor, a more compact reactor is favorable. The use of the high field magnet with Bmax = 16T (for example Ti-doped Nb3Sn conductor) enables to reduce the total machine size to 50% of the above-described conventional design, namely Rp = 7m, Vp = 760m-3, PF = 2.8 GW. (author)

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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)

  3. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  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. Material erosion and migration in tokamaks

    International Nuclear Information System (INIS)

    The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the subsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasma-facing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined

  6. 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.

  7. 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

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. 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

  13. Disruption generated secondary runaway electrons in present day tokamaks

    International Nuclear Information System (INIS)

    An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

  14. 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

  15. Design and construction of the KSTAR tokamak

    International Nuclear Information System (INIS)

    The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and 'advanced tokamak' physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long-pulse current-drive and heating systems. The 'advanced tokamak' aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. (author)

  16. The tokamak - an imperfect frame of refernce

    International Nuclear Information System (INIS)

    It is attempted to assess the suitability of tokamaks for fusion power plants on the basis of existing design studies by reference to the reality of energy production in fission power plants. A definition of suitability criteria and a discussion of their relation to the most important features of power plants are followed by a comparative treatment. For example, the mean volumetric net electric power density in the nuclear islands of tokamak power plant designs is only 2,5 to 4 E of the value common today in light water reactor nuclear islands. In addition, configuration problems, auxiliary power requirements and energy payback time are discussed and taken into account in the assessment. (orig.)

  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. The physics of tokamak start-up

    International Nuclear Information System (INIS)

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

  19. Microinstability theory in tokamaks: a review

    International Nuclear Information System (INIS)

    Significant investigations in the area of tokamak microinstability theory are reviewed. Emphasis is given to the work covering the period from 1970 through 1976. Special attention is focused on low-frequency electrostatic drift-type modes, which are generally believed to be the dominant tokamak microinstabilities under normal operating conditions. The basic linear formalism including electromagnetic (finite beta) modifications is presented along with a general survey of the numerous papers investigating specific linear and nonlinear effects on these modes. Estimates of the associated anomalous transport and confinement times are discussed, and a summary of relevant experimental results is given. Studies of the nonelectrostatic and high-frequency instabilities associated with the presence of high energy ions from neutral beam injection (or with the presence of alpha particles from fusion reactions) are also surveyed

  20. 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.

  1. Runaway acceleration during magnetic reconnection in tokamaks

    International Nuclear Information System (INIS)

    In this paper, the basic theory of runaway electron production is reviewed and recent progress is discussed. The mechanisms of primary and secondary generation of runaway electrons are described and their dynamics during a tokamak disruption is analysed, both in a simple analytical model and through numerical Monte Carlo simulation. A simple criterion for when these mechanisms generate a significant runaway current is derived, and the first self-consistent simulations of the electron kinetics in a tokamak disruption are presented. Radial cross-field diffusion is shown to inhibit runaway avalanches, as indicated in recent experiments on JET and JT-60U. Finally, the physics of relativistic post-disruption runaway electrons is discussed, in particular their slowing down due to emission of synchrotron radiation, and their ability to produce electron-positron pairs in collisions with bulk plasma ions and electrons

  2. 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.

  3. 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

  4. 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)

  5. 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)

  6. Global migration of impurities in tokamaks

    International Nuclear Information System (INIS)

    The migration of impurities in tokamaks has been studied with the help of tracer-injection (13C and 15N) experiments in JET and ASDEX Upgrade since 2001. We have identified a common pattern for the migrating particles: scrape-off layer flows drive impurities from the low-field side towards the high-field side of the vessel. Migration is also sensitive to the density and magnetic configuration of the plasma, and strong local variations in the resulting deposition patterns require 3D treatment of the migration process. Moreover, re-erosion of the deposited particles has to be taken into account to properly describe the migration process during steady-state operation of the tokamak. (paper)

  7. KTM Tokamak operation scenarios software infrastructure

    International Nuclear Information System (INIS)

    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.

  8. 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.

  9. 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.

  10. Electromagnetic effects of plasma disruptions in tokamaks

    International Nuclear Information System (INIS)

    The tokamak is modeled as typically 100 mutually-coupled toroidal circuits. The self and mutual inductances and the currents and voltages are calculated. Using the calculated currents, the poloidal magnetic field and the electromagnetic forces as functions of space and time are calculated. The major conclusion of the analysis is that the torus sectors should be electrically connected to each other near the plasma. Such connections reduce the structural loads, eliminate arcing, and reduce the induced potentials in the poloidal field coils

  11. Confinement scaling and ignition in tokamaks

    International Nuclear Information System (INIS)

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 1015 cm-3, high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition

  12. 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

  13. Comparison of tokamak burn cycle options

    International Nuclear Information System (INIS)

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options

  14. Microwave correllation reflectometry for tokamak CASTOR

    Czech Academy of Sciences Publication Activity Database

    Nanobashvili, S.; Žáček, František; Zajac, Jaromír

    2005-01-01

    Roč. 55, č. 6 (2005), s. 701-719. ISSN 0011-4626 R&D Projects: GA AV ČR IAA1043101 Grant ostatní: GA EU(EU) INTAS ´2001 1B-2056 Institutional research plan: CEZ:AV0Z20430508 Keywords : microwaves * tokamak * plasma * turbulence * reflectometry Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.360, year: 2005

  15. Bifurcated Helical Core Equilibrium States in Tokamaks

    International Nuclear Information System (INIS)

    Full text: Tokamaks with weak to moderate reversed central magnetic shear in which the minimum of the inverse rotational transform qmin is in the neighbourhood of unity can trigger bifurcated MagnetoHydroDynamic (MHD) equilibrium states. In addition to the standard axisymmetric branch that can be obtained with standard Grad-Shafranov solvers, a novel branch with a three-dimensional (3D) helical core has been computed with the ANIMEC code, an anisotropic pressure extension of the VMEC code. The solutions have imposed nested magnetic flux surfaces and are similar to saturated ideal internal kink modes. The difference in energy between both possible branches is very small. Plasma elongation, current and β enhance the susceptibility for bifurcations to occur. An initial value nonlinear ideal MHD evolution of the axisymmetric branch compares favourably with the helical core equilibrium structures calculated. Peaked prescribed pressure profiles reproduce the 'snake' structures observed in many tokamaks which has led to a new explanation of the snake as a bifurcated helical equilibrium state that results from a saturated ideal internal kink in which pellets or impurities induce a hollow current profile. Snake equilibrium structures are computed in free boundary TCV tokamak simulations. Magnetic field ripple and resonant magnetic perturbations in MAST free boundary calculations do not alter the helical core deformation in a significant manner when qmin is near unity. These bifurcated solutions constitute a paradigm shift that motivates the application of tools developed for stellarator research in tokamak physics investigations. The examination of fast ion confinement in this class of equilibria is performed with the VENUS code in which a coordinate independent noncanonical phase-space Lagrangian formulation of guiding centre drift orbit theory has been implemented. (author)

  16. 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

  17. Frascati Tokamak Upgrade (FTU): Results and developments

    International Nuclear Information System (INIS)

    In the present note the relation is examined between the FTU experimental programme and the most important issues in controlled thermonuclear fusion researches. FTU is a high-density, high magnetic field tokamak devoted to the study of plasma heating and current drive, energy and particle confinement and plasma-wall interaction. The most important FTU results and their relevance for ITER will be discussed

  18. 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." PMID:26636854

  19. Tokamak with liquid metal toroidal field coil

    Science.gov (United States)

    Ohkawa, Tihiro; Schaffer, Michael J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.

  20. Lower hybrid heating of tokamaks to ignition

    International Nuclear Information System (INIS)

    The incorporation of a quasi-linear collisional wave damping model of lower hybrid electron heating into a radial transport code reveals favourable prospects for heating tokamak plasmas to ignition. The RF frequencies considered here are such that the wave interaction is primarily with the electrons. For a particular test reactor design, 30 MW of lower hybrid power used in conjunction with a programmed plasma density start-up suffices to initiate a self-sustained thermonuclear burn. (author)

  1. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  2. Development of Atomic Beam Probe for tokamaks

    Czech Academy of Sciences Publication Activity Database

    Berta, M.; Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S.; Havlíček, Josef; Háček, Pavel

    2013-01-01

    Roč. 88, č. 11 (2013), s. 2875-2880. ISSN 0920-3796 R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ABP * Plasma diagnostics * COMPASS tokamak * Current density * Plasma density profile measurement Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613005048#

  3. 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."

  4. Tokamak exhaust process for the ITER project

    International Nuclear Information System (INIS)

    The ITER project calls for an unprecedented amount of hydrogen isotopes to be processed. To facilitate environmental responsibility and economic application of fusion technology, the re-use of hydrogen isotopes is vital. The US ITER Project Office (USIPO) is responsible for the front end of the ITER Tritium Plant, the Tokamak Exhaust Processing (TEP) System. The TEP system must separate the Tokamak exhaust gases into a stream containing only hydrogen isotopes and a stream containing only non-hydrogen gases. The USIPO has selected the Savannah River National Laboratory (SRNL) in partnership with the Los Alamos National Laboratory (LANL) to complete the TEP portion of the project. SRNL's participation builds on the laboratory's decades of work with hydrogen and its isotopes deuterium and tritium - providing the applied research and development that supports the Savannah River Site's handling of tritium. SRNL's experience and expertise in large-scale tritium processing systems and its track record of effective project execution are a unique combination that is key to the success of the ITER project. LANL brings to the partnership experience and expertise in tritium processing technologies specific to the fusion program. This knowledge and understanding were gained through the development and operation of the Tritium Systems Test Assembly at Los Alamos for over 20 years starting in the late 1970's. The US's implementation of the tokamak exhaust processing (TEP) system will provide a technically mature, robust, and cost-effective solution for the separation of hydrogen isotopes from the tokamak exhaust stream. The TEP technology, design challenges, and project status will be presented. (orig.)

  5. MHD stability of an almost circular tokamak

    International Nuclear Information System (INIS)

    In a tokamak, the ratio β between the plasma pressure and that of the magnetic field is limited by the appearance of instabilities. The magnetic field in a tokamak reactor will always be limited by technological constraints. It is therefore crucial to know what factors have an effect on the β limit, since a zero resistivity plasma fluid model allows for theoretical reproduction of the β limits observed experimentally. Theoretical studies have shown that the distributions of pressure and current density may have a substantial effect on the β limit. The effect of the current density and pressure distributions on the β limit has been studied for tokamak with a circular core section. The best results are obtained when the current density is concentrated in the centre of the section and is nil at the periphery. But the second region of stability against ballooning modes cannot be obtained in a circular tokamak owing to the destabilisation of the universal modes. This study was then extended to the stability of plasmas the section of which is almost circular and has a point of reflection. Such configurations are vital for fusion since they allow systems in which the confinement time does not deteriorate with an increase in the additional heating power. The β limit was calculated for different positions of the reflection point. The results show that when it is displaced from the interior towards the exterior of the torus, the stability of the overall modes is progressively improved until it is vertical. But if the point of reflection is further displaced from this vertical position towards the exterior of the torus, localised modes close to the edge of the plasma are destabilised and bring about a drop in the β limit. (author) figs., tabs., 80 refs

  6. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    International Nuclear Information System (INIS)

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

  7. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    Science.gov (United States)

    Azizov, E. A.

    2012-02-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.

  8. 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)

  9. Management and protection system for superconducting tokamak

    Science.gov (United States)

    Juszczyk, B.; Wojenski, A.; Zienkiewicz, P.; Kasprowicz, G.; Pozniak, K.; Romaniuk, R.

    2015-09-01

    This paper describes system for a diagnostics of a high-voltage power supply section of tokamaks. System is designed to assure reliability and safety of power supply subsystems. It is divided into two main components: remote and local. Remote part is located near tokamak, whereas local part can be localised away from the tokamak area. The remote side consists of custom, standalone devices. On the other hand, the local device is based on the uTCA.4 architecture. Components are connected with an optic fibre over a link-layer protocol which provides high throughput, low latency and transmission redundancy. All main operations ie. data processing, transmission etc. are performed on the FPGA devices. At the local side there is one device treated as a master device. It implements sort of a routing table which connects consecutive system inputs and outputs. It also provides possibility for some user defined data processing. This document contains general system overview, short description of hardware used in the project and gateware implementation.

  10. 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.

  11. Relativistic runaway electrons in tokamak plasmas

    International Nuclear Information System (INIS)

    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)

  12. The minimum dissipation state for tokamaks

    International Nuclear Information System (INIS)

    The principle of minimum dissipation rate subject to helicity and energy balance is applied to tokamaks with an arbitrary aspect ratio. We solved the resulting Euler-Lagrange equations analytically and numerically. It is found that for low and general aspect ratio tokamaks, there exists different typical minimum dissipation state, corresponding to the typical experimental current profile respectively. It is also found that there exist different types of relaxed states in different regions of the parameter space for a selected device. Three forms of current profile are presented under different experimental conditions for a low aspect ratio tokamak like NSTX. The first peaks in the edge region of the high field side similar as the typical experimental form. The second peaks in the central region on the equatorial plane. The third may have a hole or reverse in the central part. E0/ηB0 is the key parameter in determining the final relaxed state; both the second and the third states could be obtained violently by increasing it to be above a critical value. (author)

  13. Electron cyclotron emission from tokamak plasmas

    International Nuclear Information System (INIS)

    Emitted electron radiation can be used as a diagnostic signal to measure the electron temperature of a thermonuclear plasma. This type of diagnostics is well established in tokamak physics. In ch. 2 of this thesis the development, calibration and special design features are treated of a six-channel prototype of a twelve-channel grating spectrometer which is built for JET at Culham for electron cyclotron emission (ECE) measurements. In order to test this prototype measurements have been performed with the T-10 tokamak at the Kurchatov Institute in Moscow. With this prototype nearly half of the temperature profile of the T-10 could be measured. Detailed observations of sawteeth instabilities have been performed. Plasma heating by electron cyclotron resonance heating experiments was studied. A detailed description of these measurements and results is given in ch. 3. Often ECE spectra from tokamaks showed non-thermal features. In order to interprete them a computer code Notec has been developed. This code that calculates the ECE radiation emerging from the plasma for a 3-D configuration, is described in ch. 4. Some preliminary results and applications are presented. (Auth.)

  14. 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 Mexico, as part of an agreement between the Faculties of Mech.-Elect. Eng. and Phy. Sci.-Maths. The main objective for having MEDUSA is to train students in plasma physics & technical related issues, aiming a full design of a medium size device (e.g. Tokamak-T). Details of technical modifications and a preliminary scientific programme will be presented. MEDUSA-MX will also benefit any developments in the existing Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  15. Nonlinear simulation studies of tokamaks and STs

    International Nuclear Information System (INIS)

    The multilevel physics, massively parallel plasma simulation code, M3D has been used to study spherical toris (STs) and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX under strong toroidal flow is explained. Internal reconnection events in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-β disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g. through a fast momentum source. Normally, however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion driven n = 1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n = 0. (author)

  16. Nonlinear simulation studies of tokamaks and ST's

    International Nuclear Information System (INIS)

    The multilevel physics, massively parallel plasma simulation code, M3D has been used to study ST's and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX under strong toroidal flow is explained. IRE's in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-β disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g., through a fast momentum source. Normally however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion-driven n=1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n=0. (author)

  17. Collisionless microtearing modes in standard tokamak configurations

    International Nuclear Information System (INIS)

    Microtearing Modes (MTM) are electromagnetic microinstabilities occurring in magnetically confined fusion plasmas driven by parallel electron current and collisions in the presence of electron temperature gradient. MTMs were first predicted to occur in such plasmas in early 70s. Collisional MTMs have recently gathered attention in Spherical Tokamak configurations and RFPs. Very recently collisional MTMs have been reported in configuration relevant to standard tokamak, namely ASDEX-U. Perhaps for the first time, we show the existence of MTMs in purely collisionless limit and in large aspect ratio tokamak configurations using fully gyrokinetic full radius linear calculations. The physics of both electron scale as well as minor radius scale are resolved in the studies. Results of the studies, such as the 2-D structure of the mode and the dependence of growth rates on plasma pressure, perpendicular (to B0) wavelength spectrum and the effect of Landau damping and magnetic drift resonance will be presented. A comparison with another electromagnetic mode, namely Kinetic Ballooning Mode, which is driven by ion temperature gradient will also be shown. (author)

  18. Design of the ITER tokamak assembly tools

    International Nuclear Information System (INIS)

    ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values

  19. 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

  20. 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