WorldWideScience

Sample records for control core plasma

  1. Plasma core reactor applications

    International Nuclear Information System (INIS)

    Latham, T.S.; Rodgers, R.J.

    1976-01-01

    Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

  2. Proceeding of JSPS-CAS Core University Program seminar on production and control of high performance plasmas with advanced plasma heating and diagnostic systems

    International Nuclear Information System (INIS)

    Gao Xiang; Morita, Shigeru

    2011-02-01

    The JSPS-CAS Core University Program (CUP) seminar on 'Production and control of high performance plasmas with advanced plasma heating and diagnostic systems' took place in Guilin Bravo Hotel, Guilin, China, 1-4 November 2010. This seminar was organized in the framework of CUP in the field of plasma and nuclear fusion. Two special talks and 46 oral talks were presented in the seminar including 36 Chinese, 18 Japanese and 4 Korean attendees. Production and control of high performance plasmas is a crucial issue for realizing an advanced nuclear fusion reactor in addition to developments of advanced plasma heating and diagnostics. This seminar was motivated along the issues. Results in the field of fusion experiments obtained through CUP activities during recent two years were summarized. Possible direction of future collaboration and further encouragement of scientific activity of younger scientists were also discussed in this seminar with future experimental plans in both countries. (author)

  3. Proceeding of JSPS-CAS core university program seminar on production and control of high performance plasmas with advanced plasma heating and diagnostic systems

    International Nuclear Information System (INIS)

    Gao Xiang; Morita, Shigeru

    2009-01-01

    The JSPS-CAS Core University Program (CUP) seminar on 'Production and control of high performance plasmas with advanced plasma heating and diagnostic systems' took place in Shiner hotel, Lijiang, China, 4-7 November 2008. This seminar was organized in the framework of CUP in the field of plasma and nuclear fusion. One special talk and 34 oral talks were presented in the seminar including 16 Japanese attendees. Production and control of high performance plasmas is a crucial issue for realizing an advanced nuclear fusion reactor in addition to developments of advanced plasma heating and diagnostics. This seminar was motivated along the issues. Results obtained from CUP activities during recent four years were summarized. Several crucial issues to be resolved near future were also extracted in this seminar. The 31 of the papers are indexed individually. (J.P.N.)

  4. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    Energy Technology Data Exchange (ETDEWEB)

    H.W.Kugel, M.G.Bell, H.Schneider, J.P.Allain, R.E.Bell, R Kaita, J.Kallman, S. Kaye, B.P. LeBlanc, D. Mansfield, R.E. Nygen, R. Maingi, J. Menard, D. Mueller, M. Ono, S. Paul, S.Gerhardt, R.Raman, S.Sabbagh, C.H.Skinner, V.Soukhanovskii, J.Timberlake, L.E.Zakharov, and the NSTX Research Team

    2010-01-25

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  5. Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Schneider, H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington, Seattle, WA 98195 (United States); Sabbagh, S. [Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

  6. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Schneider, H.; Allain, J.P.; Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D.; Nygen, R.E.; Maingi, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S.; Raman, R.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Zakharov, L.E.; NSTX Research Team

    2010-01-01

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  7. Research on plasma core reactors

    International Nuclear Information System (INIS)

    Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

  8. Reactor core control device

    International Nuclear Information System (INIS)

    Sano, Hiroki

    1998-01-01

    The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

  9. Hollow core plasma channel generation

    International Nuclear Information System (INIS)

    Quast, Heinrich Martin

    2018-03-01

    The use of a hollow plasma channel in plasma-based acceleration has beneficial properties for the acceleration of electron and positron bunches. In the scope of the FLASHForward facility at DESY, the generation of such a plasma structure is examined. Therefore, the generation of a ring-shaped laser intensity profile with different techniques is analyzed. From the obtained intensity profiles the electron density of a hollow plasma channel is simulated in the focal region. Different parameters are scanned to understand their influence on the electron density distribution - an important parameter being, for example, the radius of the central region of the channel. In addition to the simulations, experiments are presented, during which a laser pulse is transformed into a hollow beam with a spiral phase plate. Subsequently, it forms a plasma during the interaction with hydrogen, where the plasma is imaged with interferometry. For energies above 0.9 mJ a hollow plasma structure can be observed at the location of first plasma formation.

  10. Plasma position control device

    International Nuclear Information System (INIS)

    Takase, Haruhiko.

    1987-01-01

    Purpose: To conduct position control stably to various plasmas and reduce the burden on the control coil power source. Constitution: Among the proportional, integration and differentiation controls, a proportional-differentiation control section and an integration control section are connected in parallel. Then, a signal switching circuit is disposed to the control signal input section for the proportional-differentiation control section such that either a present position of plasmas or deviation between the present plasma position and an aimed value can be selected as a control signal depending on the control procedures or the state of the plasmas. For instance, if a rapid response is required for the control, the deviation between the present plasma position and the aimed value is selected as the input signal to conduct proportional, integration and differentiation controls. While on the other hand, if it is intended to reduce the burden on the control coil power source, it is adapted such that the control signal inputted to the proportional-differentiation control section itself can select the present plasma position. (Yoshihara, H.)

  11. Modeling of impurity transport in the core plasma

    International Nuclear Information System (INIS)

    Hulse, R.A.

    1992-01-01

    This paper presents a brief overview of computer modeling of impurity transport in the core region of controlled thermonuclear fusion plasmas. The atomic processes of importance in these high temperature plasmas and the numerical formulation of the model are described. Selected modeling examples are then used to highlight some features of the physics of impurity behavior in large tokamak fusion devices, with an emphasis on demonstrating the sensitivity of such modeling to uncertainties in the rate coefficients used for the atomic processes. This leads to a discussion of current requirements and opportunities for generating the improved sets of comprehensive atomic data needed to support present and future fusion impurity modeling studies

  12. Plasma control device

    International Nuclear Information System (INIS)

    Matsutomi, Akiyoshi.

    1995-01-01

    Plasma position and shape estimation values are outputted based on measured values of coil current. When the measured values of the position and the shape are judged to be abnormal, position and shape estimation values estimated by a plasma position and shape estimation means are outputted as position and shape feed back values to a plasma position and shape control means instead of the position and shape measured values. Since only a portion of the abnormal position and shape measured values may also be replaced with the position and shape estimation values, errors in the plasma position and shape feed back values can be reduced as a whole. In addition, even if the position and shape measured values are abnormal or if they can not be measured, plasma control can be continued by alternative position and shape estimation values, thereby enabling to avoid interruption of plasma control. Since the position and shape estimation values are obtained not only with the measured values of coil current but also with the position and shape estimation values, the accuracy is improved. Further, noises superposed on the position and shape measured values are filtered, and the device is stabilized compared with a prior art device. (N.H.)

  13. Plasma control device

    International Nuclear Information System (INIS)

    Takase, Haruhiko.

    1987-01-01

    Purpose: To obtain the optimum controllability for the plasmas and the thermonuclear device by selectively executing control operation for proportion, integration and differentiation (PID) by first and second controllers respectively based on selection instruction signals. Constitution: Deviation between a vertical direction equilibrium position: Zp as the plasma status amount measured in a measuring section and an aimed value Zref thereof is inputted to a first PID selection controller. The first controller selectively executes one of the PID control operations in accordance with the first selection signal instruction instructed by a PID control operation instruction circuit. Further, Zp is also inputted to a second PID selection controller, which selectively executes one of the PID control operations in accordance with the second selection instruction signal in the same manner as in the first controller. The deviation amount u between operations signals u1 and u2 from the first and second PID selection controllers is inputted to a power source to thereby supply a predetermined current value to control coils that generate equilibrium magnetic fields for making the vertical direction equilibrium position of plasmas constant. (Kamimura, M.)

  14. Transport of recycled deuterium to the plasma core in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Bell, M.G.; Budny, R.V.; Jassby, D.L.; Park, H.; Ramsey, A.T.; Stotler, D.P.; Strachan, J.D.

    1997-10-01

    The authors report a study of the fueling of the plasma core by recycling in the Tokamak Fusion Test Reactor (TFTR). They have analyzed discharges fueled by deuterium recycled from the limiter and tritium-only neutral beam injection. In these plasmas, the DT neutron rate provides a measure of the deuterium influx into the core plasma. They find a reduced influx with plasmas using lithium pellet conditioning and with plasmas of reduced major (and minor) radius. Modeling with the DEGAS neutrals code shows that the dependence on radius can be related to the penetration of neutrals through the scrape-off layer

  15. Combined core/boundary layer plasma transport simulations in tokamaks

    International Nuclear Information System (INIS)

    Prinja, A.K.; Schafer, R.F. Jr.; Conn, R.W.; Howe, H.C.

    1987-01-01

    Significant new numerical results are presented from self-consistent core and boundary or scrape-off layer plasma simulations with 3-D neutral transport calculations. For a symmetric belt limiter it is shown that, for plasma conditions considered here, the pump limiter collection efficiency increases from 11% to 18% of the core efflux as a result of local reionization of blade deflected neutrals. This hitherto unobserved effect causes a significant amplification of upstream ion flux entering the pump limiter. Results from coupling of an earlier developed two-zone edge plasma model ODESSA to the PROCTR core plasma simulation code indicates that intense recycling divertor operation may not be possible because of stagnation of upstream flow velocity. This results in a self-consistent reduction of density gradient in an intermediate region between the central plasma and separatrix, and a concomitant reduction of core-efflux. There is also evidence of increased recycling at the first wall. (orig.)

  16. Research program on plasma core assembly

    International Nuclear Information System (INIS)

    Bernard, W.; Helmick, H.H.; Jarvis, G.A.; Plassmann, E.A.; White, R.H.

    1975-05-01

    An operating critical assembly having a 1.024-m-diam by 1.055-m-high cavity reflected by 0.48 m of beryllium has been constructed of residual beryllium reflector segments, control drums, drive motors, and control console from the solid core nuclear rocket development program. The critical mass for uranium distributed throughout the cavity is 16.9 kg U(93.2), which is high because of neutron undermoderation in the reflector due to porosity and poison contaminants in the beryllium and graphite reflector components. A flux trapping beryllium annulus, 0.546-m-i.d. by 0.89-m-o.d., centered in the cavity, reduced the critical mass to 6.7 kg U(93.2). This arrangement will permit operation with a UF 6 zone inside the flux trap driven by an outer uranium-graphite fuel zone. The fission density inside the flux trap is appreciably higher than in the outer fuel zone. (U.S.)

  17. Applications of plasma core reactors to terrestrial energy systems

    International Nuclear Information System (INIS)

    Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

  18. Reactor-core-reactivity control device

    International Nuclear Information System (INIS)

    Miura, Teruo; Sakuranaga, Tomonobu.

    1983-01-01

    Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

  19. Sawtooth control in fusion plasmas

    International Nuclear Information System (INIS)

    Graves, J P; Angioni, C; Budny, R V; Buttery, R J; Coda, S; Eriksson, L-G; Gimblett, C G; Goodman, T P; Hastie, R J; Henderson, M A; Koslowski, H R; Mantsinen, M J; Martynov, An; Mayoral, M-L; Mueck, A; Nave, M F F; Sauter, O; Westerhof, E

    2005-01-01

    Clear observations of early triggering of neo-classical tearing modes by sawteeth with long quiescent periods have motivated recent efforts to control, and in particular destabilize, sawteeth. One successful approach explored in TCV utilizes electron cyclotron heating in order to locally increase the current penetration time in the core. The latter is also achieved in various machines by depositing electron cyclotron current drive or ion cyclotron current drive close to the q = 1 rational surface. Crucially, localized current drive also succeeds in destabilizing sawteeth which are otherwise stabilized by a co-existing population of energetic trapped ions in the core. In addition, a recent reversed toroidal field campaign at JET demonstrates that counter-neutral beam injection (NBI) results in shorter sawtooth periods than in the Ohmic regime. The clear dependence of the sawtooth period on the NBI heating power and the direction of injection also manifests itself in terms of the toroidal plasma rotation, which consequently requires consideration in the theoretical interpretation of the experiments. Another feature of NBI, expected to be especially evident in the negative ion based neutral beam injection (NNBI) heating planned for ITER, is the parallel velocity asymmetry of the fast ion population. It is predicted that a finite orbit effect of asymmetrically distributed circulating ions could strongly modify sawtooth stability. Furthermore, NNBI driven current with non-monotonic profile could significantly slow down the evolution of the safety factor in the core, thereby delaying sawteeth

  20. Sawtooth control in fusion plasmas

    Science.gov (United States)

    Graves, J. P.; Angioni, C.; Budny, R. V.; Buttery, R. J.; Coda, S.; Eriksson, L.-G.; Gimblett, C. G.; Goodman, T. P.; Hastie, R. J.; Henderson, M. A.; Koslowski, H. R.; Mantsinen, M. J.; Martynov, An; Mayoral, M.-L.; Mück, A.; Nave, M. F. F.; Sauter, O.; Westerhof, E.; Contributors, JET–EFDA

    2005-12-01

    Clear observations of early triggering of neo-classical tearing modes by sawteeth with long quiescent periods have motivated recent efforts to control, and in particular destabilize, sawteeth. One successful approach explored in TCV utilizes electron cyclotron heating in order to locally increase the current penetration time in the core. The latter is also achieved in various machines by depositing electron cyclotron current drive or ion cyclotron current drive close to the q = 1 rational surface. Crucially, localized current drive also succeeds in destabilizing sawteeth which are otherwise stabilized by a co-existing population of energetic trapped ions in the core. In addition, a recent reversed toroidal field campaign at JET demonstrates that counter-neutral beam injection (NBI) results in shorter sawtooth periods than in the Ohmic regime. The clear dependence of the sawtooth period on the NBI heating power and the direction of injection also manifests itself in terms of the toroidal plasma rotation, which consequently requires consideration in the theoretical interpretation of the experiments. Another feature of NBI, expected to be especially evident in the negative ion based neutral beam injection (NNBI) heating planned for ITER, is the parallel velocity asymmetry of the fast ion population. It is predicted that a finite orbit effect of asymmetrically distributed circulating ions could strongly modify sawtooth stability. Furthermore, NNBI driven current with non-monotonic profile could significantly slow down the evolution of the safety factor in the core, thereby delaying sawteeth.

  1. Controlled fusion and plasma heating

    International Nuclear Information System (INIS)

    1990-06-01

    The contributions presented in the 17th European Conference on Controlled Fusion and Plasma Heating were focused on Tore Supra investigations. The following subjects were presented: ohmic discharges, lower hybrid experiments, runaway electrons, Thomson scattering, plasma density measurements, magnetic fluctuations, polarization scattering, plasma currents, plasma fluctuation measurements, evaporation of hydrogen pellets in presence of fast electrons, ripple induced stochastic diffusion of trapped particles, tearing mode stabilization, edge effects on turbulence behavior, electron cyclotron heating, micro-tearing modes, divertors, limiters

  2. The control of TCV plasmas

    International Nuclear Information System (INIS)

    Lister, J.B.; Hofmann, F.; Moret, J.M.

    1996-07-01

    The general control of tokamak plasmas has evolved considerably over the last few years with an increase in the plasma pulse length, an increase in the control of additional heating and fuelling and an increase in the degree to which the shape of the plasma can be varied. The TCV tokamak is specifically designed to explore the operational benefits of plasma shaping over a wide variety of plasma shapes. Consequently, considerable attention has been given to the control of the poloidal field coil currents which impose the desired shape. This paper deals with all aspects of the control of TCV plasmas, from the diagnostic measurements to the power supplies, via control algorithms and overall supervision. (author) 44 figs., tabs., 25 refs

  3. The control of TCV plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lister, J B; Hofmann, F; Moret, J M [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP); and others

    1996-07-01

    The general control of tokamak plasmas has evolved considerably over the last few years with an increase in the plasma pulse length, an increase in the control of additional heating and fuelling and an increase in the degree to which the shape of the plasma can be varied. The TCV tokamak is specifically designed to explore the operational benefits of plasma shaping over a wide variety of plasma shapes. Consequently, considerable attention has been given to the control of the poloidal field coil currents which impose the desired shape. This paper deals with all aspects of the control of TCV plasmas, from the diagnostic measurements to the power supplies, via control algorithms and overall supervision. (author) 44 figs., tabs., 25 refs.

  4. Feedback control of plasma position in the HL-1 tokamak

    International Nuclear Information System (INIS)

    Yuan Baoshan; Jiao Boliang; Yang Kailing

    1991-01-01

    In the HL-1 tokamak with a thick copper shell, the control of plasma position is successfully performed by a feedback-feedforward system with dual mode regulator and the equilibrium field coils outside the shell. The plasma position can be controlled within ±2 mm in both vertical and horizontal directions under the condition that the iron core of transformer is not saturated

  5. Plasma control and utilization

    International Nuclear Information System (INIS)

    Ensley, D.L.

    1976-01-01

    A plasma is confined and heated by a microwave field resonant in a cavity excited in a combination of the TE and TM modes while responding to the resonant frequency of the cavity as the plasma dimensions change to maintain operation at resonance. The microwave field is elliptically or circularly polarized as to prevent the electromagnetic confining field from going to zero. A high Q chamber having superconductive walls is employed to minimize wall losses while providing for extraction of thermonuclear energy produced by fusion of nuclei in the plasma. 24 claims, 15 figures

  6. Chapter 8: Plasma operation and control

    Science.gov (United States)

    ITER Physics Expert Group on Disruptions, Control, Plasma, and MHD; ITER Physics Expert Group on Energetic Particles, Heating, Current and Drive; ITER Physics Expert Group on Diagnostics; ITER Physics Basis Editors

    1999-12-01

    Wall conditioning of fusion devices involves removal of desorbable hydrogen isotopes and impurities from interior device surfaces to permit reliable plasma operation. Techniques used in present devices include baking, metal film gettering, deposition of thin films of low-Z material, pulse discharge cleaning, glow discharge cleaning, radio frequency discharge cleaning, and in situ limiter and divertor pumping. Although wall conditioning techniques have become increasingly sophisticated, a reactor scale facility will involve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficient removal of tritium incorporated into co-deposited layers on plasma facing components and their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized. Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactor scale plasmas will rely on the same principles, but will face additional challenges, exemplified by the ITER/FDR design. The absolute positioning of outermost flux surface and divertor strike points will have to be precise and reliable in view of the high heat fluxes at the separatrix. Long pulses will require minimal control actions, to reduce accumulation of AC losses in superconducting PF and TF coils. To this end, more complex feedback controllers are envisaged, and the experimental validation of the plasma equilibrium response models on which such controllers are designed is encouraging. Present simulation codes provide an adequate platform on which equilibrium response techniques can be validated. Burning plasmas require kinetic control in addition to traditional magnetic shape and position control. Kinetic control refers to measures controlling density, rotation and temperature in the plasma core as

  7. Analysis of core plasma heating and ignition by relativistic electrons

    International Nuclear Information System (INIS)

    Nakao, Y.

    2002-01-01

    Clarification of the pre-compressed plasma heating by fast electrons produced by relativistic laser-plasma interaction is one of the most important issues of the fast ignition scheme in ICF. On the basis of overall calculations including the heating process, both by relativistic hot electrons and alpha-particles, and the hydrodynamic evolution of bulk plasma, we examine the feature of core plasma heating and the possibility of ignition. The deposition of the electron energy via long-range collective mode, i.e. Langmuir wave excitation, is shown to be comparable to that through binary electron-electron collisions; the calculation neglecting the wave excitation considerably underestimates the core plasma heating. The ignition condition is also shown in terms of the intensity I(h) and temperature T(h) of hot electrons. It is found that I(h) required for ignition increases in proportion to T(h). For efficiently achieving the fast ignition, electron beams with relatively 'low' energy (e.g.T(h) below 1 MeV) are desirable. (author)

  8. Plasma physics for controlled fusion

    International Nuclear Information System (INIS)

    Miyamoto, K.

    2010-01-01

    The primary objective of this lecture note is to present the theories and experiments of plasma physics for recent activities of controlled fusion research for graduate and senior undergraduate students. Chapters 1-6 describe the basic knowledge of plasma and magnetohydrodynamics (MHD). MHD instabilities limit the beta ratio (ratio of plasma pressure to magnetic pressure) of confined plasma. Chapters 7-9 provide the kinetic theory of hot plasma and discuss the wave heating and non-inductive current drive. The dispersion relation derived by the kinetic theory are used to discuss plasma waves and perturbed modes. Landau damping is the essential mechanism of plasma heating and the stabilization of perturbation. Landau inverse damping brings the amplification of waves and the destabilization of perturbed modes. Chapter 10 explains the plasma transport due to turbulence, which is the most important and challenging subject for plasma confinement. Theories and simulations including subject of zonal flow are introduced. Chapters 11, 12 and 13 describe the recent activities of tokamak including ITER as well as spherical tokamak, reversed field pinch (RFP) and stellarator including quasi-symmetric configurations. Emphasis has been given to tokamak research since it made the most remarkable progress and the construction phase of 'International Tokamak Experimental Reactor' called ITER has already started. (author)

  9. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  10. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  11. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1978-01-01

    Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  12. Plasma control concepts for ITER

    International Nuclear Information System (INIS)

    Lister, J.B.; Nieswand, C.

    1997-01-01

    This overview paper skims over a wide range of issues related to the control of ITER plasmas. Although operation of the ITER project will require extensive developmental work to achieve the degree of control required, there is no indication that any of the identified problems will present overwhelming difficulties compared with the operation of present tokamaks. However, the precision of control required and the degree of automation of the final ITER plasma control system will present a challenge which is somewhat greater than for present tokamaks. In order to operate ITER optimally, integrated use of a large amount of diagnostic information will be necessary, evaluated and interpreted automatically. This will challenge both the diagnostics themselves and their supporting interpretation codes. The intervening years will provide us with the opportunity to implement and evaluate most of the new features required for ITER on existing tokamaks, with the exception of the control of an ignited plasma. (author) 7 figs., 7 refs

  13. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    1995-01-01

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC)

  14. Controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

  15. Controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

  16. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    1994-07-01

    40 papers are presented at this 21. conference on controlled fusion and plasma physics (JET). Titles are: effects of sawtooth crashes on beams ions and fusion product tritons; beta limits in H-modes and VH-modes; impurity induced neutralization of MeV energy protons in JET plasmas; lost α particle diagnostic for high-yield D-T fusion plasmas; 15-MeV proton emission from ICRF-heated plasmas; pulse compression radar reflectometry for density measurements; gamma-ray emission profile measurements during ICRH discharges; the new JET phase ICRH array; simulation of triton burn-up; parametric dependencies of JET electron temperature profiles; detached divertor plasmas; excitation of global Alfven Eigenmodes by RF heating; mechanisms of toroidal rotation; effect of shear in the radial electric field on confinement; plasma transport properties at the L-H transition; numerical study of plasma detachment conditions in JET divertor plasmas; the SOL width and the MHD interchange instability; non linear magnetic reconnection in low collisionality plasmas; topology and slowing down of high energy ion orbits; sawtooth crashes at high beta; fusion performances and alpha heating in future JET D-T plasmas; a stable route to high-beta plasmas with non-monotonic q-profiles; theory of propagation of changes to confinement; spatial distribution of gamma emissivity and fast ions during ICRF heating; multi-camera soft X-ray diagnostic; radiation phenomena and particle fluxes in the X-event; local measurement of transport parameters for laser injected trace impurities; impurity transport of high performance discharges; negative snakes and negative shear; neural-network charge exchange analysis; ion temperature anisotropy in helium neutral beam fuelling; impurity line emission due to thermal charge exchange in edge plasmas; control of convection by fuelling and pumping; VH mode accessibility and global H-mode properties; ion cyclotron emission by spontaneous emission; LHCD/ICRH synergy

  17. DOE Radiological Control Manual Core Training Program

    International Nuclear Information System (INIS)

    Scott, H.L.; Maisler, J.

    1993-01-01

    Over the past year, the Department of Energy (DOE) Office of Health (EH-40) has taken a leading role in the development of new standardized radiological control training programs for use throughout the DOE complex. The Department promulgated its Radiological Control (RadCon) Manual in June 1992. To ensure consistent application of the criteria presented in the RadCon Manual, standardized radiological control core training courses and training materials have been developed for implementation at all DOE facilities. In producing local training programs, standardized core courses are to be supplemented with site-specific lesson plans, viewgraphs, student handbooks, qualification standards, question banks, and wallet-sized training certificates. Training programs for General Employee Radiological Training, Radiological Worker I and II Training, and Radiological Control Technician Training have been disseminated. Also, training committees under the direction of the Office of Health (EH-40) have been established for the development of additional core training courses, development of examination banks, and the update of the existing core training courses. This paper discusses the current activities and future direction of the DOE radiological control core training program

  18. LIDAR Thomson scattering for ITER core plasma revisited

    International Nuclear Information System (INIS)

    Gowers, C.; Nielsen, P.; Salzmann, H.

    2016-01-01

    The authors have become aware that the development of the hitherto planned time-of-flight Thomson scattering system for the ITER core plasma is not proceeding and that conventional Thomson scattering set-ups are being discussed as an alternative. In this paper, we want to point out the advantages of LIDAR and show how criticized details of the original design can be improved. We present a design of the front optics, which in neutronics terms closely resembles a layout already previously accepted. The presented design does not require Raman scattering calibration for the density measurement. Comparison with the JET Core LIDAR system and simulations at higher temperatures both show that with the new design the specified accuracy can be met with a 1–2 J laser. Current laser and detector technology is reviewed. A strategy for how to proceed is presented

  19. Plasma position and shape control for ITER

    International Nuclear Information System (INIS)

    Portone, A.; Gribov, Y.; Huguet, M.

    1995-01-01

    Key features and main results about the control of the plasma shape in ITER are presented. A control algorithm is designed to control up to 6 gaps between the plasma separatrix and the plasma facing components during the reference burn phase. Nonlinear simulations show the performances of the controller in the presence of plasma vertical position offsets, beta drops and power supply voltage saturation

  20. Controlling marginally detached divertor plasmas

    Science.gov (United States)

    Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

    2017-06-01

    A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

  1. Coherent control of plasma dynamics

    Science.gov (United States)

    He, Zhaohan

    2014-10-01

    The concept of coherent control - precise measurement or determination of a process through control of the phase of an applied oscillating field - has been applied to numerous systems with great success. Here, we demonstrate the use of coherent control on plasma dynamics in a laser wakefield electron acceleration experiment. A tightly focused femtosecond laser pulse (10 mJ, 35 fs) was used to generate electron beams by plasma wakefield acceleration in the density down ramp. The technique is based on optimization of the electron beam using a deformable mirror adaptive optical system with an iterative evolutionary genetic algorithm. The image of the electrons on a scintillator screen was processed and used in a fitness function as direct feedback for the optimization algorithm. This coherent manipulation of the laser wavefront leads to orders of magnitude improvement to the electron beam properties such as the peak charge and beam divergence. The laser beam optimized to generate the best electron beam was not the one with the ``best'' focal spot. When a particular wavefront of laser light interacts with plasma, it can affect the plasma wave structures and trapping conditions of the electrons in a complex way. For example, Raman forward scattering, envelope self-modulation, relativistic self-focusing, and relativistic self-phase modulation and many other nonlinear interactions modify both the pulse envelope and phase as the pulse propagates, in a way that cannot be easily predicted and that subsequently dictates the formation of plasma waves. The optimal wavefront could be successfully determined via the heuristic search under laser-plasma conditions that were not known a priori. Control and shaping of the electron energy distribution was found to be less effective, but was still possible. Particle-in-cell simulations were performed to show that the mode structure of the laser beam can affect the plasma wave structure and trapping conditions of electrons, which

  2. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  3. Core electron-root confinement (CERC) in helical plasmas

    International Nuclear Information System (INIS)

    Yokoyama, M.; Ida, K.; Maassbcrg, H.

    2006-10-01

    The improvement of core electron heat confinement has been realized in a wide range of helical devices such as CHS, LHD, TJ-II and W7-AS. Strongly peaked electron temperature profiles and large positive radial electric field, E r , in the core region are common fractures for this improved confinement. Such observations are consistent with a transition to the electron-root' solution of the ambipolarity condition for E r in the context of the neoclassical transport, which is unique to non-axisymmetric configurations. Based on this background, this improved confinement has been collectively dubbed 'core electron-root confinement' (CERC). The electron heat diffusivity is much reduced due to the electron-root E r compared to that with E r =0 assumed, which clearly demonstrates that 1/v ripple diffusion (ν being the collision frequency) in low-collisional helical plasmas could be overcome. The magnetic configuration properties play important roles in this transition, and thresholds are found for the collisionality and electron cyclotron heating (ECH) power. (author)

  4. Core-edge coupling and the effect of the edge on overall plasma performance

    International Nuclear Information System (INIS)

    Fichtmueller, M.; Corrigan, G.; Lauro-Taroni, L.

    1999-01-01

    Several attempts to model the entire plasma cross section have been reported in the last few years. Two possibilities are to either couple a core code to a scrape-off layer (SOL) code at a specified interface or to extend the computational region of an SOL-code all the way to the plasma centre. The most advanced global code is the code COCONUT which is based on the former principle and comprises the Monte-Carlo code NIMBUS, the 2D scrape-off layer code EDGE2D, the core transport code JETTO and the core impurity transport code SANCO. A main feature of COCONUT is its modular structure which ensures a high degree of flexibility and the capability to cover a large range of time-scales. The influence of the SOL on the core is illustrated with a range of global simulations carried out with COCONUT. The simulations show that the primary effect of the SOL is the control of the particle sources and sinks with a secondary effect on plasma dilution, radiation and perhaps pedestal temperatures. (author)

  5. Apparatus for controlling nuclear core debris

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    Disclosed is an apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling

  6. Apparatus for controlling nuclear core debris

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

  7. Development of compact toroids injector for direct plasma controls

    Energy Technology Data Exchange (ETDEWEB)

    Azuma, K. [Mitsubishi Heavy Industries Ltd., Takasago (Japan); Oda, Y. [Mitsubishi Heavy Industries Ltd., Takasago (Japan); Onozuka, M. [Mitsubishi Heavy Industries Ltd., Takasago (Japan); Uyama, T. [Himeji Inst. of Tech. (Japan); Nagata, M. [Himeji Inst. of Tech. (Japan); Fukumoto, N. [Himeji Inst. of Tech. (Japan)

    1995-12-31

    The application of the compact toroids injector for direct plasma controls has been investigated. The compact toroids injection can fuel particles directly into the core of the plasma and modify the plasma profiles at the desired locations. The acceleration tests of the compact toroids have been conducted at Himeji Institute of Technology. The tests showed that the hydrogen compact toroid was accelerated up to 80km/s and the plasma density of the compact toroid was compressed to 1.2 x 10{sup 21}m{sup -3}. (orig.).

  8. Development of compact toroids injector for direct plasma controls

    International Nuclear Information System (INIS)

    Azuma, K.; Oda, Y.; Onozuka, M.; Uyama, T.; Nagata, M.; Fukumoto, N.

    1995-01-01

    The application of the compact toroids injector for direct plasma controls has been investigated. The compact toroids injection can fuel particles directly into the core of the plasma and modify the plasma profiles at the desired locations. The acceleration tests of the compact toroids have been conducted at Himeji Institute of Technology. The tests showed that the hydrogen compact toroid was accelerated up to 80km/s and the plasma density of the compact toroid was compressed to 1.2 x 10 21 m -3 . (orig.)

  9. Plasma current sustainment after iron core saturation in the STOR-M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Ding, Y.; Hubeny, M.; Lu, Y.; Onchi, T.; McColl, D.; Xiao, C.; Hirose, A. [Plasma Physics Laboratory, University of Saskatchewan, 116 Science Place, Saskatoon, SK S7N 5E2 (Canada)

    2014-10-15

    Highlights: • Plasma current can be started up by small iron core without central solenoid. • Iron core removes central solenoid. • Plasma current can be maintained after iron core saturation. • Hysteresis curve shows the partial core saturation. • Image field from iron core is estimated during discharge. • Spherical tokamak reactor without CS is proposed using the small iron core. - Abstract: We propose to use of a small iron core transformer to start up the plasma current in a spherical tokamak (ST) reactor without central solenoid (CS). Taking advantage of the high aspect ratio of the STOR-M iron core tokamak, we have demonstrated that the plasma current up to 10–15 kA can be started up using the outer Ohmic heating (OH) coils without CS, and that the plasma current can be maintained further by increasing the outer OH coil current during iron core saturation phase. When the magnetizing current reaches 1.2 kA and the iron core becomes saturated, the third capacitor bank connected to the outer OH coils is discharged to maintain the plasma current. The plasma current is slightly increased and maintained for additional 5 ms as expected from numerical calculations. Core saturation has been clearly observed on the hysteresis curve. This is the first experimental demonstration of the feasibility of slow transition from the iron core to air core transformer phase without CS. The results implies that a plasma current can be initiated by a small iron core and could be ramped up by additional heating and vertical field after iron core saturation in future STs without CS.

  10. Open core control software for surgical robots.

    Science.gov (United States)

    Arata, Jumpei; Kozuka, Hiroaki; Kim, Hyung Wook; Takesue, Naoyuki; Vladimirov, B; Sakaguchi, Masamichi; Tokuda, Junichi; Hata, Nobuhiko; Chinzei, Kiyoyuki; Fujimoto, Hideo

    2010-05-01

    In these days, patients and doctors in operation room are surrounded by many medical devices as resulting from recent advancement of medical technology. However, these cutting-edge medical devices are working independently and not collaborating with each other, even though the collaborations between these devices such as navigation systems and medical imaging devices are becoming very important for accomplishing complex surgical tasks (such as a tumor removal procedure while checking the tumor location in neurosurgery). On the other hand, several surgical robots have been commercialized, and are becoming common. However, these surgical robots are not open for collaborations with external medical devices in these days. A cutting-edge "intelligent surgical robot" will be possible in collaborating with surgical robots, various kinds of sensors, navigation system and so on. On the other hand, most of the academic software developments for surgical robots are "home-made" in their research institutions and not open to the public. Therefore, open source control software for surgical robots can be beneficial in this field. From these perspectives, we developed Open Core Control software for surgical robots to overcome these challenges. In general, control softwares have hardware dependencies based on actuators, sensors and various kinds of internal devices. Therefore, these control softwares cannot be used on different types of robots without modifications. However, the structure of the Open Core Control software can be reused for various types of robots by abstracting hardware dependent parts. In addition, network connectivity is crucial for collaboration between advanced medical devices. The OpenIGTLink is adopted in Interface class which plays a role to communicate with external medical devices. At the same time, it is essential to maintain the stable operation within the asynchronous data transactions through network. In the Open Core Control software, several

  11. LIDAR TS for ITER core plasma. Part I: layout & hardware

    Science.gov (United States)

    Salzmann, H.; Gowers, C.; Nielsen, P.

    2017-12-01

    The original time-of-flight design of the Thomson scattering diagnostic for the ITER core plasma has been shown up by ITER. This decision was justified by insufficiencies of some of the components. In this paper we show that with available, present day technology a LIDAR TS system is feasible which meets all the ITER specifications. As opposed to the conventional TS system the LIDAR TS also measures the high field side of the plasma. The optical layout of the front end has been changed only little in comparison with the latest one considered by ITER. The main change is that it offers an optical collection without any vignetting over the low field side. The throughput of the system is defined only by the size and the angle of acceptance of the detectors. This, in combination with the fact that the LIDAR system uses only one set of spectral channels for the whole line of sight, means that no absolute calibration using Raman or Rayleigh scattering from a non-hydrogen isotope gas fill of the vessel is needed. Alignment of the system is easy since the collection optics view the footprint of the laser on the inner wall. In the described design we use, simultaneously, two different wavelength pulses from a Nd:YAG laser system. Its fundamental wavelength ensures measurements of 2 keV up to more than 40 keV, whereas the injection of the second harmonic enables measurements of low temperatures. As it is the purpose of this paper to show the technological feasibility of the LIDAR system, the hardware is considered in Part I of the paper. In Part II we demonstrate by numerical simulations that the accuracy of the measurements as required by ITER is maintained throughout the given plasma parameter range. The effect of enhanced background radiation in the wavelength range 400 nm-500 nm is considered. In Part III the recovery of calibration in case of changing spectral transmission of the front end is treated. We also investigate how to improve the spatial resolution at the

  12. Plasma and controlled thermonuclear reaction

    International Nuclear Information System (INIS)

    Kapitsa, P.

    1980-01-01

    The principle and prospects are given of three methods of achieving controlled thermonuclear reaction. The original and so far most promising TOKAMAK method is presented invented in the USSR. Another method is the heating of a sphere about 1 mm in diameter from a mixture of deuterium and tritium by focused laser light from all sides. The third method consists in continuous plasma heating. A rope-like plasma discharge at a temperature of more than a million K results in the gas from microwave oscillations. The discharge is placed in a magnetic field and the ion temperature is increased by magneto-acoustic waves. A reactor is proposed operating on this principle and problems are pointed out which will have to be resolved. (M.S.)

  13. Plasma and controlled thermonuclear reaction

    Energy Technology Data Exchange (ETDEWEB)

    Kapitsa, P

    1980-06-01

    The principle and prospects are given of three methods of achieving controlled thermonuclear reaction. The original and so far most promising TOKAMAK method is presented invented in the USSR. Another method is the heating of a sphere about 1 mm in diameter from a mixture of deuterium and tritium by focused laser light from all sides. The third method consists in continuous plasma heating. A rope-like plasma discharge at a temperature of more than a million K results in the gas from microwave oscillations. The discharge is placed in a magnetic field and the ion temperature is increased by magneto-acoustic waves. A reactor is proposed operating on this principle and problems are pointed out which will have to be resolved.

  14. The plasma position control of ITER EDA plasma

    International Nuclear Information System (INIS)

    Senda, Ikuo; Nishio, Satoshi; Tsunematsu, Toshihide; Nishino, Toru; Fujieda, Hirobumi.

    1994-09-01

    The study on the plasma position control of ITER EDA performed by Japan Home Team during the sensitivity study in 1994 is summarized. The controllabilities of plasmas in the Outline Design and elongated version are compared. The model used to describe the motion of the plasma is a rigid model. The PD feedback control is applied with respect to the displacements of the plasma from the equilibrium. Three types of fluctuations, which initiate the motion of the plasma, are examined, namely a finite horizontal fluctuation field, a small horizontal fluctuation field such that the motion of the plasma is governed by the passive structures and an abrupt change of the poloidal beta β p and internal inductance l i . In the simulations of finite horizontal fluctuation fields, controls depend on the strength of the fluctuations, for instance, 3-5V is needed for 5-10G of fluctuation fields in the Outline Design. When the fluctuation field is small and the plasma displacement grows in a characteristic time of the passive structures, a few volt of the control voltage is enough to obtain good controllability. It is shown that the control when (β p , l i ) changes simultaneously is demanding and a large control voltage is required to maintain satisfactory control. Comparing the elongated version with the Outline Design, the control voltage which is larger than the Outline Design by a factor of 2-3 is required to obtain the same controllability in the elongated version. (author)

  15. JT-60 plasma control system

    International Nuclear Information System (INIS)

    Kurihara, K.

    1988-01-01

    JT-60 plasma control can be performed by the supervisory controller, the measurement system and actuators such as the poloidal field coil power supplies, gas injectors, neutral beam injection (NBI) heating system and radio frequency (RF) heating system. One of the most important characteristics of this system is a perfect digital control one composed of mini-computers, fast array processors and CAMAC modules, and it has large flexibility and few troubles to adjust the system. This system started to be operated in April 1985, after the six-year-long design, construction and testing, and have been operated and improved many times for two years. In this paper, the final system specification and its performance are presented aiming at the technological aspect of hardware and software. In addition, and experienced troubles are also presented. (author)

  16. The stabilizing effect of core pressure on the edge pedestal in MAST plasmas

    International Nuclear Information System (INIS)

    Chapman, I.T.; Simpson, J.; Saarelma, S.; Kirk, A.; O'Gorman, T.; Scannell, R.

    2015-01-01

    The pedestal pressure measured in Mega Ampere Spherical Tokamak plasmas has been shown to increase as the global plasma pressure increases. By deliberately suppressing the transition into the high-confinement regime, the core plasma pressure was systematically altered at the time of the first edge localized mode. Stability analysis shows that the enhanced Shafranov shift at higher core pressure stabilizes the ballooning modes driven by the pedestal pressure gradient, consequently allowing the pedestal to reach higher pressures. (paper)

  17. Automatic plasma control in magnetic traps

    International Nuclear Information System (INIS)

    Samojlenko, Y.; Chuyanov, V.

    1984-01-01

    Hot plasma is essentially in thermodynamic non-steady state. Automatic plasma control basically means monitoring deviations from steady state and producing a suitable magnetic or electric field which brings the plasma back to its original state. Briefly described are two systems of automatic plasma control: control with a magnetic field using a negative impedance circuit, and control using an electric field. It appears that systems of automatic plasma stabilization will be an indispensable component of the fusion reactor and its possibilities will in many ways determine the reactor economy. (Ha)

  18. Core/shell PLGA microspheres with controllable in vivo release profile via rational core phase design.

    Science.gov (United States)

    Yu, Meiling; Yao, Qing; Zhang, Yan; Chen, Huilin; He, Haibing; Zhang, Yu; Yin, Tian; Tang, Xing; Xu, Hui

    2018-02-27

    Highly soluble drugs tend to release from preparations at high speeds, which make them need to be taken at frequent intervals. Additionally, some drugs need to be controlled to release in vivo at certain periods, so as to achieve therapeutic effects. Thus, the objective of this study is to design injectable microparticulate systems with controllable in vivo release profile. Biodegradable PLGA was used as the matrix material to fabricate microspheres using the traditional double emulsification-solvent evaporation method as well as improved techniques, with gel (5% gelatine or 25% F127) or LP powders as the inner phases. Their physicochemical properties were systemically investigated. Microspheres prepared by modified methods had an increase in drug loading (15.50, 16.72, 15.66%, respectively) and encapsulation efficiencies (73.46, 79.42, 74.40%, respectively) when compared with traditional methods (12.01 and 57.06%). The morphology of the particles was characterized by optical microscope (OM) and scanning electron microscopy (SEM), and the amorphous nature of the encapsulated drug was confirmed by differential scanning calorimetry (DSC) and X-ray diffraction (XRD) analysis. To evaluate their release behaviour, the in vitro degradation, in vitro release and in vivo pharmacodynamics were subsequently studied. Traditional microspheres prepared in this study with water as the inner phase had a relatively short release period within 16 d when compared with modified microspheres with 5% gelatine as the inner phase, which resulted in a smooth release profile and appropriate plasma LP concentrations over 21 d. Thus this type of modified microspheres can be better used in drugs requiring sustained release. The other two formulations containing 25% F127 and LP micropowders presented two-stage release profiles, resulting in fluctuant plasma LP concentrations which may be suitable for drugs requiring controlled release. All the results suggested that drug release rates from

  19. Edge plasma control using an LID configuration on CHS

    Energy Technology Data Exchange (ETDEWEB)

    Masuzaki, S.; Komori, A.; Morisaki, T. [National Inst. for Fusion Science, Oroshi, Toki (Japan)] [and others

    1997-07-01

    A Local Island Divertor (LID) has been proposed to enhance energy confinement through neutral particle control. For the case of the Large Helical Device (LHD), the separatrix of an m/n = 1/1 magnetic island, formed at the edge region, will be utilized as a divertor configuration. The divertor head is inserted in the island, and the island separatrix provides connection between the edge plasma region surrounding the core plasma and the back plate of the divertor head through the field lines. The particle flux and associated heat flux from the core plasma strike the back plate of the divertor head, and thus particle recycling is localized in this region. A pumping duct covers the divertor head to form a closed divertor system for efficient particle exhaust. The advantages of the LID are ease of hydrogen pumping because of the localized particle recycling and avoidance of the high heat load that would be localized on the leading edge of the divertor head. With efficient pumping, the neutral pressure in the edge plasma region will be reduced, and hence the edge plasma temperature will be higher, hopefully leading to a better core confinement region. A LID configuration experiment was done on the Compact Helical System (CHS) to confirm the effect of the LID. The typical effects of the LID configuration on the core plasma are reduction of the line averaged density to a half, and small or no reduction of the stored energy. In this contribution, the experimental results which were obtained in edge plasma control experiments with the LID configuration in the CHS are presented.

  20. Static and dynamic control of plasma equilibrium in a Tokamak

    International Nuclear Information System (INIS)

    Blum, J.; Dei Cas, R.

    1979-01-01

    We are dealing here with the problem of controlling the plasma boundary and its displacements. Static control consists in determining the currents in the external coils of the Tokamak so that the plasma boundary has certain fixed characteristics: radial position, vertical elongation, desired shape. A self-consistent method is proposed here, considering a free plasma boundary, and using the techniques of optimal control of distributed parameter systems to solve the problem. The dynamic control problem considered in the second part of the paper is the control of the plasma radial displacements. An elaborate system of preprogramming and feedback control has been developed to ensure equilibrium and stability of the horizontal plasma motions. Optimal control techniques have been used to calculate the optimal primary coils configuration, the preprogramming voltages and the feedback gains. A new stability diagrams has been obtained which takes into account the erosion of the plasma by the limiter. All these calculations have been applied successfully to TFR 600 where thin liner and the presence of an iron core make the problem of stabilization of the radial displacements very difficult

  1. Initial operation of NSTX with plasma control

    International Nuclear Information System (INIS)

    Gates, D.; Bell, M.; Ferron, J.; Kaye, S.; Menard, J.; Mueller, D.; Neumeyer, C.; Sabbagh, S.

    2000-01-01

    First plasma, with a maximum current of 300kA, was achieved on NSTX in February 1999. These results were obtained using preprogrammed coil currents. The first controlled plasmas on NSTX were made starting in August 1999 with the full 1MA plasma current achieved in December 1999. The controlled quantities were plasma position (R, Z) and current (Ip). Variations in the plasma shape are achieved by adding preprogrammed currents to those determined by the control parameters. The control system is fully digital, with plasma position and current control, data acquisition, and power supply control all occurring in the same four-processor real time computer. The system uses the PCS (Plasma Control Software) system designed at General Atomics. Modular control algorithms, specific to NSTX, were written and incorporated into the PCS. The application algorithms do the actual control calculations, with the PCS handling data passing. The control system, including planned upgrades, will be described, along with results of the initial controlled plasma operations. Analysis of the performance of the control system will also be presented

  2. Plasma automatic control in magnetic traps

    International Nuclear Information System (INIS)

    Samojlenko, Yu.I.; Chuyanov, V.A.

    1983-01-01

    Principles of constructing the systems providing a plasma equilibrium and stability in thermonuctear devices are laid down. Operation of the servo system to maintain a plasma equilibrium is described using the tokamak plasma filament as an example. Operation of the system to suppress a flute instability is also described. This system measures electric disturbances on the plasma body surface and controls charge distribution on external electrodes. It is pointed out that systems of automatic control of plasma equilibrium and stability become an essential element of a future thermonuclear reactor and the system potentialities would much determine the reactor economic efficiency

  3. Core flow control system for field applications; Sistema de controle de core-flow

    Energy Technology Data Exchange (ETDEWEB)

    Granzotto, Desiree G.; Adachi, Vanessa Y.; Bannwart, Antonio C.; Moura, Luiz F.M. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil); Sassim, Natache S.D.A. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Centro de Estudo do Petroleo (CEPETRO); Carvalho, Carlos H.M. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

    2008-07-01

    The significant heavy oil reserves worldwide and the presently high crude oil prices make it essential the development of technologies for heavy oil production and transportation. Heavy oils, with their inherent features of high viscosity (100- 10,000 cP) and density (below 20 deg API) require specific techniques to make it viable their flow in pipes at high flow rates. One of the simplest methods, which do not require use of heat or diluents, is provided by oil-water annular flow (core-flow). Among the still unsolved issues regarding core-flow is the two-phase flow control in order to avoid abrupt increases in the pressure drop due to the possible occurrence of bad water-lubricated points, and thus obtain a safe operation of the line at the lowest possible water-oil ratio. This work presents results of core flow tests which allow designing a control system for the inlet pressure of the line, by actuating on the water flow rate at a fixed oil flow rate. With the circuit model and the specified controller, simulations can be done to assess its performance. The experiments were run at core-flow circuit of LABPETRO-UNICAMP. (author)

  4. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1991-01-01

    On JET results were presented on additional heating power, on a recently discovered regime of enhanced pellet performance (PEP), on low-density H-mode plasma confinement with hot ions, bounds on very high electric currents by material limiters, the first experiments on lower hybrid current drive, on the L-H transition threshold dependence on the direction of the gradient-B drift, and on alpha-particle physics issues. The TFTR presentations focused on transport. Particle loss ramifications of the toroidal Alfven eigenmodes were found to be small, while their threshold of excitation is lower than theoretically predicted. On DIII-D a scaling study of transport with gyroradius as the only variable was reported, with approximately Bohm scaling emerging; but the effective heat diffusivity scaling could not be established due to profile consistency effects. While beta-limit investigations with DIII-D generally confirm the ideal, MHD limit found by Troyon, evidence of a reduction of the accessible range for the internal inductance with the safety factor seems to favour current-density control in a steady-state D-T burner. Onset of strongly sheared poloidal rotation in a thin layer during the L-H mode transition was experimentally shown, while a new, so-called VH (''very high'') confinement mode was discovered by boronization of the wall. The JT-90 tokamak has recently been upgraded to JT-60-U. Presentations by the ASDEX team summarized the lack of agreement with theory of L-mode confinement. With TEXTOR, an improved mode (I-mode) of confinement was found by boronization. Finally, reviews are included on the status of impurity transport and helium removal in tokamaks, on stellarators, alternative magnetic confinement systems, inertial confinement, and non-fusion plasma physics. 2 tabs

  5. Operation and control of ITER plasmas

    International Nuclear Information System (INIS)

    2001-01-01

    Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

  6. Operation and control of ITER plasmas

    International Nuclear Information System (INIS)

    1999-01-01

    Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

  7. Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions

    International Nuclear Information System (INIS)

    Sengoku, Seio

    1985-08-01

    Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)

  8. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  9. Longitudinal and radial MHD linear induction accelerator with hot conducting plasma core

    International Nuclear Information System (INIS)

    Denno, K.

    1985-01-01

    Conceptual design of linear induction accelerator is presented using for the core continuum a highly conductive plasma with sustained pumping velocity. Karlovitz criterion of boundary theory is employed in the process of design

  10. Plasma position control on Alcator C

    International Nuclear Information System (INIS)

    Pribyl, P.A.

    1981-05-01

    The Alcator C MHD equilibrium is investigated from the standpoint of determining the plasma position. A review of equilibrium theory is presented, indicating that the central flux surfaces of the plasma should be displaced about 1 to 2 cm from the outermost. Further, the plasma should have a slightly noncircular cross-section. A comparison is made between the observed and predicted profiles. Flux loops sensitive to plasma position generate the error signal for the feedback control circuit. This measurement agrees with other position-sensitive diagnostics, such as limiter heating, and centroids of density, soft x-ray, and electron cyclotron emission. A linear model is developed for the position control feedback system, including the vertical field SCR supply, plasma, and feedback electronics. Operation of the control system agrees well with that predicted by the model, with acceptable plasma position being maintained for the duration of the discharge. The feedback control system is in daily use for Alcator C runs

  11. Transitions to improved core electron heat confinement in JT-II plasmas

    International Nuclear Information System (INIS)

    Estrada, T.; Medina, F.; Ascasibar, E.; Balbin, R.; Castejon, F.; Hidalgo, C.; Lopez-Bruna, D.; Petrov, S.

    2008-01-01

    Transitions to improved core electron heat confinement are triggered by low order rational magnetic surfaces in TJ-II ECH plasmas. Transitions triggered by the rational surface n=4/m=2 show an increase in the ion temperature synchronized with the increase in the electron temperature. SXR measurements demonstrate that, under certain circumstances, the rational surface positioned inside the plasma core region precedes and provides a trigger for the transition. (author)

  12. Control method for thermonuclear plasma

    International Nuclear Information System (INIS)

    Azuma, Kingo; Oda, Yasushi.

    1997-01-01

    CT (Compact Troid) is a doughnut-like shaped plasmas having a toroidal current and a poloidal current at the inside and forming a poloidal magnetic fluxes and toroidal magnetic flux. The structure of the CT is collapsed at a time of stationary state, accordingly, when it is injected to thermonuclear plasmas, particles can be supplied locally, and the state of the plasmas to be supplied can be changed by changing the direction of the injection. If a CT which is reverse to the poloidal magnetic fields is injected, plasmas with excessive ions can be supplied locally thereby enabling to form magnetic field in the thermonuclear plasmas. If the magnetic fields are formed in the vicinity of the surface of the thermonuclear plasmas, fast ions which have come over the magnetic field structure can be returned to the central portion of the plasmas. Then, confining performance of thermonuclear plasmas can be greatly improved, the efficiency for fuel supply can be increased, and energy required for ignition can be suppressed. (N.H.)

  13. Enhancement of EAST plasma control capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Bingjia, E-mail: bjxiao@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei (China); Yuan, Qiping; Luo, Zhengping; Huang, Yao; Liu, Lei; Guo, Yong; Pei, Xiaofang; Chen, Shuliang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Humphreys, D.A.; Hyatt, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Mueller, Dennis [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Calabró, G.; Crisanti, F. [ENEA UnitàTecnicaFusione, C.R. Frascati, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Albanese, R.; Ambrosino, R. [CREATE, Università di Napoli Federicao II, Università di Cassino and Università di Napoli Parthenope, Via Claudio 19, 80125 Napoli (Italy)

    2016-11-15

    Highlights: • Parallel plasma equilibrium reconstruction using GPU for real-time control on EAST. • Vertical control using Bang-bang + PID method to improve the response and minimize the oscillation caused by the latency. • Quasi-snow flake divertor plasma configuration has been demonstrated on EAST. - Abstract: In order to improve the plasma control performance and enhance the capability for advanced plasma control, new algorithms such as PEFIT/ISOFLUX plasma shape feedback control, quasi-snowflake plasma shape development and vertical control under new vertical control power supply, have been implemented and experimentally tested and verified in EAST 2014 campaign. P-EFIT is a rewritten version of EFIT aiming at fast real-time equilibrium reconstruction by using GPU for parallelized computation. Successful control using PEFIT/ISOFLUX was established in dedicated experiment. Snowfldivertor plasma shape has the advantage of spreading heat over the divertor target and a quasi-snowflake (QSF) configuration was achieved in discharges with I{sub p} = 0.25 MA and B{sub t} = 1.8T, κ∼1.9, by plasma position feedback control. The shape feedback control to achieve QSF shape has been preliminary implemented by using PEFIT and the initial experimental test has been done. For more robust vertical instability control, the inner coil (IC) and its power supply have been upgraded. A new control algorithm with the combination of Bang-bang and PID controllers has been developed. It is shown that new vertical control power supply together with the new control algorithms results in higher vertical controllability.

  14. Observations on the W-transport in the core plasma of JET and ASDEX Upgrade

    Czech Academy of Sciences Publication Activity Database

    Pütterich, T.; Dux, R.; Neu, R.; Bernert, M.; Beurskens, M.N.A.; Bobkov, V.; Brezinsek, S.; Challis, C.; Coenen, J.W.; Coffey, I.; Czarnecka, A.; Giroud, C.; Jacquet, P.; Joffrin, E.; Kallenbach, A.; Lehnen, M.; Lerche, E.; De La Luna, E.; Marsen, S.; Matthews, G.; Mayoral, M.-L.; McDermott, R.M.; Meigs, A.; Mlynář, Jan; Sertoli, M.; van Rooij, G.

    2013-01-01

    Roč. 55, č. 12 (2013), s. 124036-124036 ISSN 0741-3335. [European Physical Society Conference on Plasma Physics/40./. Espoo, 01.07.2013-05.07.2013] Institutional support: RVO:61389021 Keywords : tokamak * impurity transport * core plasma * fusion * tungsten * ASDEX Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.386, year: 2013 http://iopscience.iop.org/0741-3335/55/12/124036/pdf/0741-3335_55_12_124036.pdf

  15. Apparatus for controlling molten core debris

    International Nuclear Information System (INIS)

    Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

    1972-01-01

    Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures

  16. Integrated core-SOL simulations of L-mode plasma in ITER and Indian demo

    International Nuclear Information System (INIS)

    Wisitsorasak, Apiwat; Onjun, Thawatchai; Kanjanaput, Wittawat

    2015-01-01

    Core-SOL simulations are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasma in ITER and Indian DEMO reactors operating in low confinement mode (L-Mode). In each simulation, the plasma current, temperature, and density profiles in both core and SOL region are evolved self-consistency. The SOL is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in SOL subsequently provide as the boundary conditions of core plasma region on low-confinement mode. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 2001 (MMM2001) and neoclassical transport calculated by NCLASS module together with the toroidal velocity based on the torque due to Neoclassical Toroidal Viscosity (NTV). In addition, a sensitivity analysis is explored by varying plasma parameters, such as plasma density and auxiliary heating power. Furthermore, the ignition tests are conducted to observed plasma response in each design after shutting down an auxiliary heating. (author)

  17. Combined plasma gas-phase synthesis and colloidal processing of InP/ZnS core/shell nanocrystals

    Science.gov (United States)

    Gresback, Ryan; Hue, Ryan; Gladfelter, Wayne L.; Kortshagen, Uwe R.

    2011-12-01

    Indium phosphide nanocrystals (InP NCs) with diameters ranging from 2 to 5 nm were synthesized with a scalable, flow-through, nonthermal plasma process at a rate ranging from 10 to 40 mg/h. The NC size is controlled through the plasma operating parameters, with the residence time of the gas in the plasma region strongly influencing the NC size. The NC size distribution is narrow with the standard deviation being less than 20% of the mean NC size. Zinc sulfide (ZnS) shells were grown around the plasma-synthesized InP NCs in a liquid phase reaction. Photoluminescence with quantum yields as high as 15% were observed for the InP/ZnS core-shell NCs.

  18. Transient behaviour in the plasma core of TJ-II stellarator and its relation with rational surfaces

    International Nuclear Information System (INIS)

    Estrada, T.; Luna, E. de la; Ascasibar, E; Jimenez, J.A.; Castejon, F.; Garcia-Cortes, I.; Lopez-Fraguas, A.; Sanchez, J.; Tribaldos, V.

    2002-01-01

    A transient behaviour is observed in the plasma core of TJ-II stellarator with fast drops in the electron temperature. Changes in the line-averaged density are observed synchronized with temperature drops. This phenomenon appears in plasmas created and heated using 300 kW of electron cyclotron heating with high power density. The transient behaviour resembles both, the electric pulsation discovered in CHS and the 'electron root' feature reported by the W7-AS team. The flexibility and low magnetic shear of TJ-II have permitted the identification of the plasma current as the control parameter for the appearance of this phenomenon. The results obtained during the magnetic configuration scans carried out in TJ-II points to the hypothesis that the transient behaviour is connected with the presence of a rational surface close to the plasma centre. Equilibrium calculations performed with the VMEC code reinforce this hypothesis. (author)

  19. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

    1998-01-01

    The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

  20. Recent plasma control progress on EAST

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, B.J., E-mail: bjxiao@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yuan, Q.P. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Humphreys, D.A.; Walker, M.L.; Hyatt, A.W.; Leuer, J.A.; Jackson, G.L. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Penaflor, B.G.; Pigrowski, D.A.; Johnson, R.D.; Welander, A.S. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Zhang, R.R.; Luo, Z.P.; Guo, Y.; Xing, Z.; Zhang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2012-12-15

    In recent 2 years, various algorithms to control plasma shape, current and density have been implemented or improved for EAST tokamak. These plasma control performances have been verified by either simulated or actual experimental operation, and thus plasma control basis has been established for the long pulse operation and high performance H-mode plasma operation with low hybrid wave (LHW) and ion cyclotron resonance frequency (ICRF) heating. Startup simulation has been done by using TOKSYS code for the plasma breakdown in either 3.1 Wb or 4.5 Wb initial poloidal flux state and the scenarios proved to be robust and used for routine operation. Various shape configurations have been well feedback controlled by using ISOFLUX limited, double-null or single null algorithms based on RTEFIT equilibrium reconstruction. For the long pulse operation, strike point control and magnetics drift compensation have been implemented in the plasma control system (PCS). To improve the operation safety and efficiency, the verification of magnetic diagnostics before plasma breakdown has been demonstrated adequate to prevent a discharge in case of key sensor failure.

  1. Plasma physics for controlled fusion

    CERN Document Server

    Miyamoto, Kenro

    2016-01-01

    This new edition presents the essential theoretical and analytical methods needed to understand the recent fusion research of tokamak and alternate approaches. The author describes magnetohydrodynamic and kinetic theories of cold and hot plasmas in detail. The book covers new important topics for fusion studies such as plasma transport by drift turbulence, which depend on the magnetic configuration and zonal flows. These are universal phenomena of microturbulence. They can modify the onset criterion for turbulent transport, instabilities driven by energetic particles as well as alpha particle generation and typical plasma models for computer simulation. The fusion research of tokamaks with various new versions of H modes are explained. The design concept of ITER, the international tokamak experimental reactor, is described for inductively driven operations as well as steady-state operations using non-inductive drives. Alternative approaches of reversed-field pinch and its relaxation process, stellator includi...

  2. Fast core prediction simulator for load follow control

    International Nuclear Information System (INIS)

    Yim, Man Sung; Lee, Sang Hoon; Lee, Un Chul

    1990-01-01

    An operator-assisting system for the reactor core control under power changing operating condition was developed. The system is consisted of core simulator routine and Xenon and Iodine initial condition generation routine. The initial condition generation routine, without exactly knowing the core status, is capable of providing accurate number densities and axial offset conditions of Xenon and Iodine after several hours of predictor- corrector calculations using the plant instrumentation signals of power level and power axial offset. The core simulator routine, even with the two node core model, gives equivalently accurate results as the one-dimensional model for the core behaviour simulation under power changing condition and can provide proper control strategies for load follow operation. The core simulator can also be used by the operator to develop remedial actions to restore the distorted power distribution by using its prediction capability

  3. Plasma Surface interaction in Controlled fusion devices

    International Nuclear Information System (INIS)

    1990-05-01

    The subjects presented in the 9th conference on plasma surface interaction in controlled fusion devices were: the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the effects observed in ergodic divertor experiments in Tore-Supra; the diffuse connexion induced by the ergodic divertor and the topology of the heat load patterns on the plasma facing components in Tore-Supra; the study of the influence of air exposure on graphite implanted by low energy high density deuterium plasma

  4. Flipped neutrino emissivity of hot plasma in supernova core

    Energy Technology Data Exchange (ETDEWEB)

    Goyal, A.; Dutta, S. (Department of Physics and Astrophysics, University of Delhi, Delhi 110 007 (India))

    1994-05-15

    We calculate the energy loss due to wrong-helicity sterile neutrinos produced due to the decay of plasmons into flipped neutrino pairs at relativistic temperatures and densities in the core of a nascent neutron star and compare our results with other processes.

  5. Power supply controlled for plasma torch generation

    International Nuclear Information System (INIS)

    Diaz Z, S.

    1996-01-01

    The high density of energy furnished by thermal plasma is profited in a wide range of applications, such as those related with welding fusion, spray coating and at the present in waste destruction. The waste destruction by plasma is a very attractive process because the remaining products are formed by inert glassy grains and non-toxic gases. The main characteristics of thermal plasmas are presented in this work. Techniques based on power electronics are utilized to achieve a good performance in thermal plasma generation. This work shown the design and construction of three phase control system for electric supply of thermal plasma torch, with 250 kw of capacity, as a part of the project named 'Destruction of hazard wastes by thermal plasma' actually working in the Instituto Nacional de Investigaciones Nucleares (ININ). The characteristics of thermal plasma and its generation are treated in the first chapter. The A C controllers by thyristors applied in three phase arrays are described in the chapter II, talking into account the power transformer, rectifiers bank and aliasing coil. The chapter III is dedicated in the design of the trigger module which controls the plasma current by varying the trigger angle of the SCR's; the protection and isolating unit are also presented in this chapter. The results and conclusions are discussed in chapter IV. (Author)

  6. Safety aspects of core power distribution surveillance and control

    International Nuclear Information System (INIS)

    Beraha, D.; Grumbach, R.; Hoeld, A.; Werner, W.

    1978-01-01

    The incentives for improved core surveillance and core control systems are outlined. An efficient code for evaluating the power distribution is indispensable for designing and testing such a system. The characteristics of the core simulator QUABOX/CUBBOX and the features required for off-line and on-line applications are described. The important role of the simulator for the safety assessment of a digital core control system is underlined. With regard to the safety aspects of core control, possible disturbances are classified. Simulation results are given concerning the failure of a control actuator. It is shown that means can be devised to prevent unstable behaviour of the control system and, furthermore, to contribute to a safe reactor operation by accounting for process disturbances. (author)

  7. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  8. Core-shell microparticles for protein sequestration and controlled release of a protein-laden core.

    Science.gov (United States)

    Rinker, Torri E; Philbrick, Brandon D; Temenoff, Johnna S

    2017-07-01

    Development of multifunctional biomaterials that sequester, isolate, and redeliver cell-secreted proteins at a specific timepoint may be required to achieve the level of temporal control needed to more fully regulate tissue regeneration and repair. In response, we fabricated core-shell heparin-poly(ethylene-glycol) (PEG) microparticles (MPs) with a degradable PEG-based shell that can temporally control delivery of protein-laden heparin MPs. Core-shell MPs were fabricated via a re-emulsification technique and the number of heparin MPs per PEG-based shell could be tuned by varying the mass of heparin MPs in the precursor PEG phase. When heparin MPs were loaded with bone morphogenetic protein-2 (BMP-2) and then encapsulated into core-shell MPs, degradable core-shell MPs initiated similar C2C12 cell alkaline phosphatase (ALP) activity as the soluble control, while non-degradable core-shell MPs initiated a significantly lower response (85+19% vs. 9.0+4.8% of the soluble control, respectively). Similarly, when degradable core-shell MPs were formed and then loaded with BMP-2, they induced a ∼7-fold higher C2C12 ALP activity than the soluble control. As C2C12 ALP activity was enhanced by BMP-2, these studies indicated that degradable core-shell MPs were able to deliver a bioactive, BMP-2-laden heparin MP core. Overall, these dynamic core-shell MPs have the potential to sequester, isolate, and then redeliver proteins attached to a heparin core to initiate a cell response, which could be of great benefit to tissue regeneration applications requiring tight temporal control over protein presentation. Tissue repair requires temporally controlled presentation of potent proteins. Recently, biomaterial-mediated binding (sequestration) of cell-secreted proteins has emerged as a strategy to harness the regenerative potential of naturally produced proteins, but this strategy currently only allows immediate amplification and re-delivery of these signals. The multifunctional, dynamic

  9. Control of Internal Transport Barriers in Magnetically Confined Fusion Plasmas

    Science.gov (United States)

    Panta, Soma; Newman, David; Sanchez, Raul; Terry, Paul

    2016-10-01

    In magnetic confinement fusion devices the best performance often involves some sort of transport barriers to reduce the energy and particle flow from core to edge. Those barriers create gradients in the temperature and density profiles. If gradients in the profiles are too steep that can lead to instabilities and the system collapses. Control of these barriers is therefore an important challenge for fusion devices (burning plasmas). In this work we focus on the dynamics of internal transport barriers. Using a simple 7 field transport model, extensively used for barrier dynamics and control studies, we explore the use of RF heating to control the local gradients and therefore the growth rates and shearing rates for barrier initiation and control in self-heated fusion plasmas. Ion channel barriers can be formed in self-heated plasmas with some NBI heating but electron channel barriers are very sensitive. They can be formed in self-heated plasmas with additional auxiliary heating i.e. NBI and radio-frequency(RF). Using RF heating on both electrons and ions at proper locations, electron channel barriers along with ion channel barriers can be formed and removed demonstrating a control technique. Investigating the role of pellet injection in controlling the barriers is our next goal. Work supported by DOE Grant DE-FG02-04ER54741.

  10. Tip Clearance Control Using Plasma Actuators

    Science.gov (United States)

    2007-03-01

    Clearance Control Using Plasma Actuators 4 posed by Denton (1993). A number of investigators have used partial shrouds, or " winglet " designs to...SDBD actuator Plasma enhanced aerodynamics has been demonstrated in a range of applications involving sepa- ration control, lift enhancement, drag... aerodynamic benefits of a squealer tip geometry. Specifically, the squealer tip is known to reduce the discharge coefficient of the tip gap, thereby

  11. Validation of ISTTOK Plasma Position Controller

    International Nuclear Information System (INIS)

    Valcarcel, D. F.; Carvalho, I. S.; Carvalho, B. B.; Fernandes, H.; Silva, C.; Duarte, P.; Duarte, A.; Carvalho, P. J.; Pereira, T.

    2008-01-01

    Active control of plasma position on the ISTTOK tokamak is of extreme importance due to the inherent instability caused by an unfavourable curvature of the external equilibrium magnetic field. The consequences of this instability can be suppressed by applying a dynamic equilibrium field. A digital real-time plasma position control system for ISTTOK has been developed to perform this task. This system uses magnetic measurements to determine the plasma position and feeds the control signal to power supplies that generate the equilibrium fields. After commissioning, the results obtained have shown some discrepancies between the magnetic plasma position reconstruction and several other diagnostics, such as tomography. This discrepancy at some extent is due to the effect of the external magnetic fields on the poloidal magnetic measurements. This work presents a study that addresses this issue. In a future work it will lead to the development of a corrected plasma position algorithm, aiming at obtaining improved performance of plasma discharges and controlled plasma column displacements

  12. The COMPASS Tokamak Plasma Control Software Performance

    Science.gov (United States)

    Valcarcel, Daniel F.; Neto, André; Carvalho, Ivo S.; Carvalho, Bernardo B.; Fernandes, Horácio; Sousa, Jorge; Janky, Filip; Havlicek, Josef; Beno, Radek; Horacek, Jan; Hron, Martin; Panek, Radomir

    2011-08-01

    The COMPASS tokamak has began operation at the IPP Prague in December 2008. A new control system has been built using an ATCA-based real-time system developed at IST Lisbon. The control software is implemented on top of the MARTe real-time framework attaining control cycles as short as 50 μs, with a jitter of less than 1 μs. The controlled parameters, important for the plasma performance, are the plasma current, position of the plasma current center, boundary shape and horizontal and vertical velocities. These are divided in two control cycles: slow at 500 μs and fast at 50 μs. The project has two phases. First, the software implements a digital controller, similar to the analog one used during the COMPASS-D operation in Culham. In the slow cycle, the plasma current and position are measured and controlled with PID and feedforward controllers, respectively, the shaping magnetic field is preprogrammed. The vertical instability and horizontal equilibrium are controlled with the faster 50-μs cycle PID controllers. The second phase will implement a plasma-shape reconstruction algorithm and controller, aiming at optimized plasma performance. The system was designed to be as modular as possible by breaking the functional requirements of the control system into several independent and specialized modules. This splitting enabled tuning the execution of each system part and to use the modules in a variety of applications with different time constraints. This paper presents the design and overall performance of the COMPASS control software.

  13. ISTTOK plasma control with the tomography diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, H.; Caralho, P.J.; Duarte, P.; Pereira, T.; Coelho, R.; Silva, C. [Association Euratom/IST, Institute of Plasmas and Nuclear Fusion, Technology Graduate Institute, P-1049-001 Lisbon (Portugal)

    2011-07-01

    A real-time plasma position control system is mandatory to achieve long duration (up to 250 ms), Alternating Current (AC) discharges on the ISTTOK tokamak. Such a system has been used for some time supported only on magnetic field diagnostic data. However, this system does not function accurately when the plasma current is low, rendering it inoperative during the plasma current reversal. A tomography diagnostic with 3 pinhole cameras and 8 silicone photodiode channels per camera was installed and customized to supply alternative plasma position to be used for plasma position control. As no filtering is applied, most of the radiation detected is in the visible/near-UV range. This system (i) executes a tomographic reconstruction, (ii) determines the average emissivity position from it, (iii) calculates the shift from the required position and (iv) supplies the vertical field power supply unit with the desired current value, all in less than 100 {mu}s. The horizontal magnetic field power supply unit is expected to be included in the system and will have no impact in the process time. This paper presents the tomography diagnostic architecture together with results of its scientific exploitation in ISTTOK AC discharges, where it has proven to be capable of supplying an accurate plasma position during the current reversal. The use of the tomography diagnostic for plasma position overcomes some limitations of the magnetic diagnostics, but poses challenges of its own such as blindness to plasma current direction. (authors)

  14. Electron internal transport barrier in the core of TJ-II ECH plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Estrada, T.; Hidalgo, C. [Laboratorio Nacional de Fusion por Confinamiento Magnetico. Asociacion EURATOM CIEMAT, Madrid (Spain); Dreval, N. [and others

    2003-07-01

    The influence of the magnetic topology on the formation of electron internal transport barriers (e-ITB) has been experimentally studied in the stellarator TJ-II. The formation of e-ITBs in electron cyclotron heated plasmas can be triggered by positioning a low order rational surface close to the plasma core region, while in configurations without any low order rational there are no indications of barrier formation within the available heating power. The e-ITB formation is characterized by an increase in the core electron temperature and plasma potential. Positive radial electric field increases in a factor of three in the plasma central region when the e-ITB forms. The results demonstrate that low order rational surfaces modify radial electric fields and electron heat transport. (orig.)

  15. Thermal and mechanical design of the plasma core CXRS diagnostics for the fusion reactor ITER; Thermische und mechanische Auslegung der Plasma Core CXRS Diagnostik des ITER Kernfusionsreaktors

    Energy Technology Data Exchange (ETDEWEB)

    Greza, H. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Neubauer, O.; Wolters, J. [Forschungszentrum Juelich GmbH (Germany)

    2009-07-01

    In the frame of the research project ITER (international thermonuclear experimental reactor) the plasma state is monitored using the plasma core diagnostics CXRS (charge exchange recombination spectroscopy).The authors describe the thermal and mechanical design of the first mirror of the CXRS diagnostics. The components of the first mirror are exposed to high heat and neutron irradiation. The surface temperature will be 300 to 400 deg C. The misalignment tolerance is plus or minus 0.1 degree. The maximum mechanical stresses in the mirror have to be minimized. The design calculations use the finite element code ANSYS. The results indicate that the heat input from the plasma can be removed by the coolant flow. Further calculation shave to concern the brazed joints between mirror and cooling block.

  16. Thermal and mechanical design of the plasma core CXRS diagnostics for the fusion reactor ITER; Thermische und mechanische Auslegung der Plasma Core CXRS Diagnostik des ITER Kernfusionsreaktors

    Energy Technology Data Exchange (ETDEWEB)

    Greza, H.; Knauff, R. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany); Neubauer, O.; Wolters, J.; Offermanns, G.; Biel, W. [Forschungszentrum Juelich GmbH (Germany)

    2011-07-01

    In the frame of the research project ITER (international thermonuclear experimental reactor) the plasma state is monitored using the plasma core diagnostics CXRS (charge exchange recombination spectroscopy).The authors describe the thermal and mechanical design of the first mirror of the CXRS diagnostics. The components of the first mirror are exposed to high heat and neutron irradiation. The surface temperature will be 300 to 400 deg C. The misalignment tolerance is plus or minus 0.1 degree. The maximum mechanical stresses in the mirror have to be minimized. The design calculations use the finite element code ANSYS. The results indicate that the heat input from the plasma can be removed by the coolant flow. Further calculation shave to concern the brazed joints between mirror and cooling block.

  17. Plasma-activated core-shell gold nanoparticle films with enhanced catalytic properties

    Energy Technology Data Exchange (ETDEWEB)

    Llorca, Jordi, E-mail: jordi.llorca@upc.edu; Casanovas, Albert; Dominguez, Montserrat; Casanova, Ignasi [Universitat Politecnica de Catalunya, Institut de Tecniques Energetiques (Spain); Angurell, Inmaculada; Seco, Miquel; Rossell, Oriol [Universitat de Barcelona, Departament de Quimica Inorganica (Spain)

    2008-03-15

    Catalytically active gold nanoparticle films have been prepared from core-shell nanoparticles by plasma-activation and characterized by high-resolution transmission electron microscopy, atomic force microscopy, and X-ray photoelectron spectroscopy. Methane can be selectively oxidized into formic acid with an O{sub 2}-H{sub 2} mixture in a catalytic wall reactor functionalized with plasma-activated gold nanoparticle films containing well-defined Au particles of about 3.5 nm in diameter. No catalytic activity was recorded over gold nanoparticle films prepared by thermal decomposition of core-shell nanoparticles due to particle agglomeration.

  18. Plasma-activated core-shell gold nanoparticle films with enhanced catalytic properties

    International Nuclear Information System (INIS)

    Llorca, Jordi; Casanovas, Albert; Dominguez, Montserrat; Casanova, Ignasi; Angurell, Inmaculada; Seco, Miquel; Rossell, Oriol

    2008-01-01

    Catalytically active gold nanoparticle films have been prepared from core-shell nanoparticles by plasma-activation and characterized by high-resolution transmission electron microscopy, atomic force microscopy, and X-ray photoelectron spectroscopy. Methane can be selectively oxidized into formic acid with an O 2 -H 2 mixture in a catalytic wall reactor functionalized with plasma-activated gold nanoparticle films containing well-defined Au particles of about 3.5 nm in diameter. No catalytic activity was recorded over gold nanoparticle films prepared by thermal decomposition of core-shell nanoparticles due to particle agglomeration

  19. Design and Architecture of SST-1 basic plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Kirit, E-mail: kpatel@ipr.res.in; Raju, D.; Dhongde, J.; Mahajan, K.; Chudasama, H.; Gulati, H.; Chauhan, A.; Masand, H.; Bhandarkar, M.; Pradhan, S.

    2016-11-15

    Highlights: • Reflective Memory network. • FPAG based Timing system for trigger distribution. • IRIG-B network for GPS time synchronization. • PMC based Digital Signal Processors and VME. • Simultaneous sampling ADC. - Abstract: Primary objective of SST-1 Plasma control system is to achieve Plasma position, shape and current profile control. Architecture of control system for SST-1 is distributed in nature. Fastest control loop time requirement of 100 μs is achieved using VME based simultaneous sampling ADCs, PMC based quad core DSP, Reflective Memory [RFM] based real-time network, VME based real-time trigger distribution network and Ethernet network. All the control loops for shape control, position control and current profile control share common signals from Magnetic diagnostic so it is planned to accommodate all the algorithms on the same PMC based quad core DSP module TS C-43. RFM based real-time data network replicate data from one node to next node in a ring network topology at sustained throughput rate of 13.4 MBps. Real-time Timing System network provides guaranteed trigger distribution in 3.8 μs from one node to all node of the network. Monitoring and configuration of different systems participating in the operation of SST-1 is done by Ethernet network. Magnetic sensors data is acquired using Pentek 6802 simultaneously sampling ADC card at the rate of 10KSPS. All the real-time raw data along with the control data will be archived using RFM network and SCSI HDD for the experiment duration of 1000 s. RFM network is also planned for real-time plotting of key parameter of Plasma during long experiment. After experiment this data is transferred to central storage server for archival purpose. This paper discusses the architecture and hardware implementation of the control system by describing all the involved hardware and software along with future plans for up-gradations.

  20. Status of DIII-D plasma control

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Penaflor, B.

    1995-10-01

    A key component of the DIII-D Advanced Tokamak and Radiative Divertor Programs is the development and implementation of methods to actively control a large number of plasma parameters. These parameters include plasma shape and position, total stored energy, density, rf loading resistance, radiated power and more detailed control of the current profile. To support this research goal, a flexible and easily expanded digital control system has been developed and implemented. We have made parallel progress in modeling of the plasma, poloidal coils, vacuum vessel, and power system dynamics and in ensuring the integrity of diagnostic and command circuits used in control. Recent activity has focused on exploiting the mature digital control platform through the implementation of simple feedback controls of more exotic plasma parameters such as enhanced divertor radiation, neutral pressure and Marfe creation and more sophisticated identification and digital feedback control algorithms for plasma shape, vertical position, and safety factor on axis (q 0 ). A summary of recent progress in each of these areas will be presented

  1. Screening of ionic cores in partially ionized plasmas within linear response

    International Nuclear Information System (INIS)

    Gericke, D. O.; Vorberger, J.; Wuensch, K.; Gregori, G.

    2010-01-01

    We employ a pseudopotential approach to investigate the screening of ionic cores in partially ionized plasmas. Here, the effect of the tightly bound electrons is condensed into an effective potential between the (free) valence electrons and the ionic cores. Even for weak electron-ion coupling, the corresponding screening clouds show strong modifications from the Debye result for elements heavier than helium. Modifications of the theoretically predicted x-ray scattering signal and implications on measurements are discussed.

  2. Real-time control of Tokamak plasmas: from control of physics to physics-based control

    International Nuclear Information System (INIS)

    Felici, F. A. A.

    2011-11-01

    Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have

  3. Initial Studies of Core and Edge Transport of NSTX Plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Bell, M.G.; Bell, R.E.; Bush, C.E.; Bourdelle, C.; Darrow, D.; Dorland, W.; Ejiri, A.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.J.; Menard, J.E.; Mueller, D.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Taylor, G.; Johnson, D.W.; Kaita, R.; Ono, M.; Paoletti, F.; Peebles, W.; Peng, Y-K.M.; Roquemore, A.L.; Skinner, C.H.; Soukhanovskii, V.A.

    2001-01-01

    Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high-beta plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by energy confinement times in neutral-beam-heated discharges that are favorable with respect to predictions from the ITER-89P scaling expression. Analysis of heat fluxes based on profile measurements with neutral-beam injection (NBI) suggest that the ion thermal transport may be exceptionally low, and that electron thermal transport is the dominant loss channel. This analysis motivates studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with k(subscript ''theta'') rho(subscript ''i'') ∼ 0.1-1 may be suppressed in these plasmas, while modes with k(subscript ''theta'') rho(subscript ''i'') ∼ 50 may be robust. High-harmonic fast-wave (HHFW) heating efficiently heats electrons on NSTX, and studies have begun using it to assess transport in the electron channel. Regarding edge transport, H-mode [high-confinement mode] transitions occur with either NBI or HHFW heating. The power required for low-confinement mode (L-mode) to H-mode transitions far exceeds that expected from empirical edge-localized-mode-free H-mode scaling laws derived from moderate aspect ratio devices. Finally, initial fluctuation measurements made with two techniques are permitting the first characterizations of edge turbulence

  4. Tokamak plasma current disruption infrared control system

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ulrickson, M.

    1987-01-01

    This patent describes a device for magnetically confining a plasma driven by a plasma current and contained within a toroidal vacuum chamber, the device having an inner toroidal limiter on an inside wall of the vacuum chamber and an arrangement for the rapid prediction and control in real time of a major plasma disruption. The arrangement is described which includes: scanning means sensitive to infrared radiation emanating from within the vacuum chamber, the infrared radiation indicating the temperature along a vertical profile of the inner toroidal limiter. The scanning means is arranged to observe the infrared radiation and to produce in response thereto an electrical scanning output signal representative of a time scan of temperature along the vertical profile; detection means for analyzing the scanning output signal to detect a first peaked temperature excursion occurring along the profile of the inner toroidal limiter, and to produce a detection output signal in repsonse thereto, the detection output signal indicating a real time prediction of a subsequent major plasma disruption; and plasma current reduction means for reducing the plasma current driving the plasma, in response to the detection output signal and in anticipation of a subsequent major plasma disruption

  5. Plasma control system upgrade and increased plasma stability in NSTX

    Energy Technology Data Exchange (ETDEWEB)

    Mastrovito, D., E-mail: dmastrovito@pppl.go [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States); Gates, D.; Gerhard, S.; Lawson, J.; Ludescher-Furth, C.; Marsala, R. [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States)

    2010-07-15

    Plasma control on the National Spherical Torus Experiment (NSTX) was previously accomplished using eight 333 MHz G4 processors built by Sky computers. Several planned improvements and additional control algorithms required significant upgrades to our real-time control computers and real-time data acquisition infrastructure. Several in-house modules have been designed and implemented including: the digital time stamp module (DITS) and for digital/analog front panel data port (FPDP) output, the FPDP output module digital/analog (FOMD/A). Standard Linux based Intel computers perform the real-time control tasks and InfiniBand as been employed for communication between a user-accessible 'host' server and the real-time computer. In addition to several independent real-time processes the General Atomics developed PCS (Bell (2006) ) system infrastructure continues to be used on NSTX. While maintaining previous functionality, improvements in the control system software include: an RWM feedback algorithm, beta feedback NBI control, more comprehensive error logging and trapping, more user-friendly interface, more complete archiving and restoring functionality, and better status reporting and diagnostic tools. Once completed, we succeeded in increasing overall plasma stability and decreasing control system latency by several times.

  6. Novel aspects of plasma control in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Humphreys, D.; Jackson, G.; Walker, M.; Welander, A. [General Atomics P.O. Box 85608, San Diego, California 92186-5608 (United States); Ambrosino, G.; Pironti, A. [CREATE/University of Naples Federico II, Napoli (Italy); Vries, P. de; Kim, S. H.; Snipes, J.; Winter, A.; Zabeo, L. [ITER Organization, St. Paul Lez durance Cedex (France); Felici, F. [Eindhoven University of Technology, Eindhoven (Netherlands); Kallenbach, A.; Raupp, G.; Treutterer, W. [Max-Planck Institut für Plasmaphysik, Garching (Germany); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Lister, J.; Sauter, O. [Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Moreau, D. [CEA, IRFM, 13108 St. Paul-lez Durance (France); Schuster, E. [Lehigh University, Bethlehem, Pennsylvania (United States)

    2015-02-15

    ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

  7. Magnetic Configuration Control of ITER Plasmas

    International Nuclear Information System (INIS)

    Albanese, R.; Artaserse, G.; Mattei, M.; Ambrosino, G.; Crisanti, F.; Tommasi, G. de; Fresa, R.; Portone, A.; Sartori, F.; Villone, F.

    2006-01-01

    The aim of this paper is to review the capability of the ITER Poloidal Field (PF) system of controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. The paper is broadly divided in two main sections devoted, respectively, to open loop (feed-forward) and closed loop (feedback) control. In the first part of the study the PF system is assessed with respect to the initiation, ramp-up, sustained burn, ramp-down phases of the main plasma inductive scenario. The limiter-to-divertor configuration transition phase is considered in detail with the aim of assessing the PF capability to form an X-point at the lowest possible current and, therefore, to relax the thermal load on the limiter surfaces. Moreover, during the sustained burn it is important to control plasmas with a broad range of current density profiles. In the second part of the study the plasma vertical feedback control requirements are assessed in details, in particular for the high elongation configurations achievable during the early limiter-to-X point transition phase. Non-rigid plasma displacement models are used to assess the control system voltage and current requirements of different radial field control circuits obtained, for example, by connecting the outermost PF coils, some CS coils, coils sub-sections etc. At last, the main 3D effects of the vessel ports are modeled and their impact of vertical stabilization evaluated. (author)

  8. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

  9. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    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.

    2004-03-01

    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

  10. EPR: High load variation performances with the 'Tmode' core control

    International Nuclear Information System (INIS)

    Grossetete, A.

    2008-01-01

    The load variation performances on a PWR are directly linked to the core control design. This design is mainly characterized by the definition of the control rod banks and the way to both perform the banks movements and to modify the core boron concentration by injection of boric acid or water. The following paper presents the principles of the T mode, the new fully automatic core control mode for the EPR which provides high performance in terms of maneuverability and optimizes the effluents. First, the paper describes the division of the control rods into two control banks (Pbank for temperature and Hbank for power distribution). Then typical movements of these banks during power changes are shown. Then, the principle of the 3 control loops (Tave, AO, Pmax), used to obtain these desired control rod movements, is given. Finally, a load following transient simulation is presented. (authors)

  11. EPR: high load variation performances with the 'TMODE' core control

    International Nuclear Information System (INIS)

    Pairot, Frederic

    2008-01-01

    The load variation performances on a PWR are directly linked to the core control design. This design is mainly characterized by the definition of the control rod banks and the way to both perform the banks movements and to modify the core boron concentration by injection of boric acid or water. The following paper presents the principles of the T mode, the new fully automatic core control mode for the EPR which provides high performance in terms of maneuverability and optimizes the effluents. First, the paper describes the division of the control rods into two control banks (Pbank for temperature and Hbank for power distribution). Then typical movements of these banks during power changes are shown. Then, the principle of the 3 control loops (Tave, AO, Pmax), used to obtain these desired control rod movements, is given. Finally, a load following transient simulation is presented. (author)

  12. Effects of nitrogen seeding on core ion thermal transport in JET ILW L-mode plasmas

    NARCIS (Netherlands)

    Bonanomi, N.; Mantica, P.; Citrin, J.; Giroud, C.; Lerche, E.; Sozzi, C.; Taylor, D.; Tsalas, M.; Van Eester, D.; JET Contributors,

    2018-01-01

    A set of experiments was carried out in JET ILW (Joint European Torus with ITER-Like Wall) L-mode plasmas in order to study the effects of light impurities on core ion thermal transport. N was puffed into some discharges and its profile was measured by active Charge Exchange diagnostics, while ICRH

  13. Soliton-plasma nonlinear dynamics in mid-IR gas-filled hollow-core fibers

    DEFF Research Database (Denmark)

    Habib, Selim; Markos, Christos; Bang, Ole

    2017-01-01

    We investigate numerically soliton-plasma interaction in a noble-gas-filled silica hollow-core anti-resonant fiber pumped in the mid-IR at 3.0 mu m. We observe multiple soliton self-compression stages due to distinct stages where either the self-focusing or the self-defocusing nonlinearity...

  14. Introduction to plasma physics and controlled fusion

    CERN Document Server

    Chen, Francis F

    2016-01-01

    The third edition of this classic text presents a complete introduction to plasma physics and controlled fusion, written by one of the pioneering scientists in this expanding field.  It offers both a simple and intuitive discussion of the basic concepts of the subject matter and an insight into the challenging problems of current research. This outstanding text offers students a painless introduction to this important field; for teachers, a large collection of problems; and for researchers, a concise review of the fundamentals as well as original treatments of a number of topics never before explained so clearly.  In a wholly lucid manner the second edition covered charged-particle motions, plasmas as fluids, kinetic theory, and nonlinear effects.  For the third edition, two new chapters have been added to incorporate discussion of more recent advances in the field.  The new chapter 9 on Special Plasmas covers non-neutral plasmas, pure electron plasmas, solid and ultra-cold plasmas, pair-ion plasmas, d...

  15. Control of Dispersion in Hollow Core Photonic Crystal Fibers

    DEFF Research Database (Denmark)

    Roberts, John

    2007-01-01

    The dispersion of hollow core photonic crystal fibers can be tailored by modifying a single ring of holes in the cladding. The dispersion can be lowered and flattened, or alternatively greatly increased, in a controlled manner.......The dispersion of hollow core photonic crystal fibers can be tailored by modifying a single ring of holes in the cladding. The dispersion can be lowered and flattened, or alternatively greatly increased, in a controlled manner....

  16. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    1999-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  17. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    2001-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  18. Performance of ARCHITECT HCV core antigen test with specimens from US plasma donors and injecting drug users.

    Science.gov (United States)

    Mixson-Hayden, Tonya; Dawson, George J; Teshale, Eyasu; Le, Thao; Cheng, Kevin; Drobeniuc, Jan; Ward, John; Kamili, Saleem

    2015-05-01

    Hepatitis C virus (HCV) core antigen is a serological marker of current HCV infection. The aim of this study was mainly to evaluate the performance characteristics of the ARCHITECT HCV core antigen assay with specimens from US plasma donors and injecting drug users. A total of 551 serum and plasma samples with known anti-HCV and HCV RNA status were tested for HCV core antigen using the Abbott ARCHITECT HCV core antigen test. HCV core antigen was detectable in 100% of US plasma donor samples collected during the pre-seroconversion phase of infection (anti-HCV negative/HCV RNA positive). Overall sensitivity of the HCV core antigen assay was 88.9-94.3% in samples collected after seroconversion. The correlation between HCV core antigen and HCV RNA titers was 0.959. HCV core antigen testing may be reliably used to identify current HCV infection. Published by Elsevier B.V.

  19. Introduction to plasma physics and controlled fusion

    CERN Document Server

    Chen, Francis F

    1984-01-01

    This complete introduction to plasma physics and controlled fusion by one of the pioneering scientists in this expanding field offers both a simple and intuitive discussion of the basic concepts of this subject and an insight into the challenging problems of current research. In a wholly lucid manner the work covers single-particle motions, fluid equations for plasmas, wave motions, diffusion and resistivity, Landau damping, plasma instabilities and nonlinear problems. For students, this outstanding text offers a painless introduction to this important field; for teachers, a large collection of problems; and for researchers, a concise review of the fundamentals as well as original treatments of a number of topics never before explained so clearly. This revised edition contains new material on kinetic effects, including Bernstein waves and the plasma dispersion function, and on nonlinear wave equations and solitons.

  20. Plasma-wall interaction data needs critical to a Burning Core Experiment (BCX)

    International Nuclear Information System (INIS)

    1985-11-01

    The Division of Development and Technology has sponsored a four day US-Japan workshop ''Plasma-Wall Interaction Data Needs Critical to a Burning Core Experiment (BCX)'', held at Sandia National Laboratories, Livermore, California on June 24 to 27, 1985. The workshop, which brought together fifty scientists and engineers from the United States, Japan, Germany, and Canada, considered the plasma-material interaction and high heat flux (PMI/HHF) issues for the next generation of magnetic fusion energy devices, the Burning Core Experiment (BCX). Materials options were ranked, and a strategy for future PMI/HHF research was formulated. The foundation for international collaboration and coordination of this research was also established. This volume contains the last three of the five technical sessions. The first of the three is on plasma materials interaction issues, the second is on research facilities and the third is from smaller working group meetings on graphite, beryllium, advanced materials and future collaborations

  1. Real-time communication for distributed plasma control systems

    Energy Technology Data Exchange (ETDEWEB)

    Luchetta, A. [Consorzio RFX, Associazione Euratom-ENEA sulla Fusione, Corso Stati Uniti 4, Padova 35127 (Italy)], E-mail: adriano.luchetta@igi.cnr.it; Barbalace, A.; Manduchi, G.; Soppelsa, A.; Taliercio, C. [Consorzio RFX, Associazione Euratom-ENEA sulla Fusione, Corso Stati Uniti 4, Padova 35127 (Italy)

    2008-04-15

    Real-time control applications will benefit in the near future from the enhanced performance provided by multi-core processor architectures. Nevertheless real-time communication will continue to be critical in distributed plasma control systems where the plant under control typically is distributed over a wide area. At RFX-mod real-time communication is crucial for hard real-time plasma control, due to the distributed architecture of the system, which consists of several VMEbus stations. The system runs under VxWorks and uses Gigabit Ethernet for sub-millisecond real-time communication. To optimize communication in the system, a set of detailed measurements has been carried out on the target platforms (Motorola MVME5100 and MVME5500) using either the VxWorks User Datagram Protocol (UDP) stack or raw communication based on the data link layer. Measurements have been carried out also under Linux, using its UDP stack or, in alternative, RTnet, an open source hard real-time network protocol stack. RTnet runs under Xenomai or RTAI, two popular real-time extensions based on the Linux kernel. The paper reports on the measurements carried out and compares the results, showing that the performance obtained by using open source code is suitable for sub-millisecond real-time communication in plasma control.

  2. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  3. Controlled fabrication of multi-core alginate microcapsules.

    Science.gov (United States)

    Eqbal, Md Danish; Gundabala, Venkat

    2017-12-01

    In this work, we present a robust microfluidic platform for controlled and complete on-chip generation of alginate microcapsules with single and double liquid cores. A combined Coflow and T-junction configuration implemented in a hybrid glass-PDMS (Polydimethylsiloxane) device is used for the generation of microcapsules with oil as liquid core. Frequency matching of oil-alginate double emulsion generation with that of aqueous Calcium chloride droplet generation allows for controlled merging of the two, resulting in reliable production of microcapsules. Confocal imaging of microcapsule cross-section reveals presence of intact liquid core. In the case of double core microcapsules, the two cores are well separated by alginate layer ensuring their long term stability. The current approach is expected to have advantages over existing techniques for liquid core microcapsule generation in terms of continuity of the process, control over core stability, and non-damage to cells when used for cell encapsulation applications. Copyright © 2017 Elsevier Inc. All rights reserved.

  4. Core functions of the Web-of-Cells control scheme

    DEFF Research Database (Denmark)

    Evenblij, Berend; Rikos, Evangelos; Heussen, Kai

    In order to maintain frequency (balancing) and voltage control in the future power system, the ELECTRA Web-of-Cells (WoC) control scheme introduces six high-level use cases, which are Balance Restoration Control (BRC), Frequency Containment Control (FCC), Inertia Response Power Control (IRPC), Ba......), Balance Steering Control (BSC), Primary Voltage Control (PVC) and Post Primary Voltage Control (PPVC). This document presents the detailed description of the core functions that are needed and sufficient for controlling the grid in a Web-of-Cells architecture....

  5. Cold plasma: Quality control and regulatory considerations

    Science.gov (United States)

    In recent years, cold plasma has emerged as a promising antimicrobial treatment for fresh and fresh-cut produce, nuts, spices, seeds, and other foods. Research has demonstrated effective control of human pathogens such as Salmonella, Listeria monocytogenes, Escherichia coli O157:H7, norovirus, and o...

  6. Plasma surface interactions in controlled fusion devices

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L.

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak

  7. Plasma surface interactions in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L. [and others

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak.

  8. The JET PCU project: An international plasma control project

    International Nuclear Information System (INIS)

    Sartori, F.; Crisanti, F.; Albanese, R.; Ambrosino, G.; Toigo, V.; Hay, J.; Lomas, P.; Rimini, F.; Shaw, S.R.; Luchetta, A.; Sousa, J.; Portone, A.; Bonicelli, T.; Ariola, M.; Artaserse, G.; Bigi, M.; Card, P.; Cavinato, M.; De Tommasi, G.; Gaio, E.

    2008-01-01

    This paper describes the new JET enhancement project 'Plasma Control Upgrade' (PCU). Initially aimed at an overhaul of JET plasma control capabilities it was eventually focused on improving the vertical stabilisation (VS) system ability to recover from large ELM (edge localised mode) perturbations. The paper describes the results of the first two years where the activity was aimed principally at researching a solution that could be implemented within the timing and budget constraints. A very important task was that of improving the modelling of JET plasma, iron core and passive structures. Using dedicated experiments, the models were progressively refined until it was possible not just to explain the experimental data but predict the VS system behaviour. At the same time the project team studied the best options for power supply (PS) and control system upgrades and evaluated whether a change of turns in the stabilisation coil was desirable and possible. A new fast radial field power supply is now being ordered and the VS control system is being upgraded

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

    International Nuclear Information System (INIS)

    Marcus, F.B.

    1978-01-01

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

  10. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

  11. Analysis of the plasma magnetohydrodynamic equilibrium in iron core transformer Tokamak HL-1M

    International Nuclear Information System (INIS)

    Chen Xiaoguang; Yuan Baoshan

    1992-01-01

    The physical and mathematical model are presented on the problem of MHD equilibrium with the self consistent in iron core transformer HL-1M. Calculation and analysis for the plasma equilibrium of the stable boundary and free boundary are shown respectively, in an axisymmetric equilibrium model of two dimensions. First, a variation formulation of the problem is written and the equations of the poloided flux ψ are solved by a finite element method; the Picard and Newton algorithms are tested for the non-linearities. The plasma boundary and the magnetic surfaces are being simulated, with the currents in the coils, the total plasma current, its current density function and the magnetic permeability of the iron being the data for the problem; a certain number of the characteristic parameter of the equilibrium configuration is calculated. Secondly, a simple method of calculation is adopted in the determination of equilibrium fields and currents in iron core HL-1M tokamak device. In the plasma equilibrium, the magnetic effect of the air gaps in the iron core and the iron magnetic shielded plate are considered in HL-1M device. Reliable data are provided for designing and constructing the poloidal field system of HL-1M device. A good computer code is constructed, which may be useful in operating on analysis for the future device

  12. Tungsten transport and sources control in JET ITER-like wall H-mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Fedorczak, N., E-mail: nicolas.fedorczak@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Monier-Garbet, P. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Pütterich, T. [MPI für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Brezinsek, S. [Institute of Energy and Climate Research, Forschungszentrum Jlich, Assoc EURATOM-FZJ, Jlich (Germany); Devynck, P.; Dumont, R.; Goniche, M.; Joffrin, E. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Lerche, E. [Association EURATOM-Belgian State, LPP-ERM-KMS, TEC partner, Brussels (Belgium); Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lipschultz, B. [York Plasma Institute, University of York, Heslington, York YO10 5DD (United Kingdom); Luna, E. de la [Laboratorio Nacional de Fusin, Asociacin EURATOM/CIEMAT, 28040 Madrid (Spain); Maddison, G. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Maggi, C. [MPI für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Matthews, G. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Nunes, I. [Istituto de plasmas e fusao nuclear, Lisboa (Portugal); Rimini, F. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Solano, E.R. [Laboratorio Nacional de Fusin, Asociacin EURATOM/CIEMAT, 28040 Madrid (Spain); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Tsalas, M. [Association EURATOM-Hellenic Republic, NCSR Demokritos 153 10, Attica (Greece); Vries, P. de [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2015-08-15

    A set of discharges performed with the JET ITER-like wall is investigated with respect to control capabilities on tungsten sources and transport. In attached divertor regimes, increasing fueling by gas puff results in higher divertor recycling ion flux, lower divertor tungsten source, higher ELM frequency and lower core plasma radiation, dominated by tungsten ions. Both pedestal flushing by ELMs and divertor screening (including redeposition) are possibly responsible. For specific scenarios, kicks in plasma vertical position can be employed to increase the ELM frequency, which results in slightly lower core radiation. The application of ion cyclotron radio frequency heating at the very center of the plasma is efficient to increase the core electron temperature gradient and flatten electron density profile, resulting in a significantly lower central tungsten peaking. Beryllium evaporation in the main chamber did not reduce the local divertor tungsten source whereas core radiation was reduced by approximately 50%.

  13. Novel magnetic controlled plasma sputtering method

    International Nuclear Information System (INIS)

    Axelevich, A.; Rabinovich, E.; Golan, G.

    1996-01-01

    A novel method to improve thin film vacuum sputtering is presented. This method is capable of controlling the sputtering plasma via an external set of magnets, in a similar fashion to the tetrode sputtering method. The main advantage of the Magnetic Controlled Plasma Sputtering (MCPS) is its ability to independently control all deposition parameters without any interference or cross-talk. Deposition rate, using the MCPS, is found to be almost twice the rate of triode and tetrode sputtering techniques. Experimental results using the MCPS to deposit Ni layers are described. It was demonstrated that using the MCPS method the ion beam intensity at the target is a result of the interaction of a homogeneous external magnetic field and the controlling magnetic fields. The MCPS method was therefore found to be beneficial for the production of pure stoichiometric thin solid films with high reproducibility. This method could be used for the production of compound thin films as well. (authors)

  14. The COMPASS Tokamak Plasma Control Software Performance

    Czech Academy of Sciences Publication Activity Database

    Valcárcel, D.F.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Janky, F.; Havlíček, Josef; Beňo, R.; Horáček, Jan; Hron, Martin; Pánek, Radomír

    2011-01-01

    Roč. 58, č. 4 (2011), s. 1490-1496 ISSN 0018-9499. [Real Time Conference, RT10/17th./. Lisboa, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-Time * ATCA * Data Acquisition * Plasma Control Software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.447, year: 2011 http://dx.doi.org/10.1109/TNS.2011.2143726

  15. Plasma Edge Control in Tore Supra

    International Nuclear Information System (INIS)

    Evans, T.E.; Mioduszewski, P.K.; Foster, C.; Haste, G.; Horton, L.; Grosman, A.; Ghendrih, P.; Chatelier, M.; Capes, H.; Michelis, C. De; Fall, T.; Geraud, A.; Grisolia, C.; Guilhem, D.; Hutter, T.

    1990-01-01

    TORE SUPRA is a large superconducting tokamak designed for sustaining long inductive pulses (t∼ 30 s). In particular, all the first wall components have been designed for steady-state heat and particle exhaust, particle injection, and additional heating. In addition to these technological assets, a strict control of the plasma-wall interactions is required. This has been done at low power: experiments with ohmic heating have been mainly devoted to the pump limiter, ergodic divertor and pellet injection experiments. Some specific problems arising in large tokamaks are encountered; the pump limiter and the ergodic divertor yield the expected effects on the plasma edge. The effects on the bulk are discussed

  16. Measurement of the complete core plasma flow across the LOC-SOC transition at ASDEX Upgrade

    Science.gov (United States)

    Lebschy, A.; McDermott, R. M.; Angioni, C.; Geiger, B.; Prisiazhniuk, D.; Cavedon, M.; Conway, G. D.; Dux, R.; Dunne, M. G.; Kappatou, A.; Pütterich, T.; Stroth, U.; Viezzer, E.; the ASDEX Upgrade Team

    2018-02-01

    A newly installed core charge exchange recombination spectroscopy (CXRS) diagnostic at ASDEX Upgrade (AUG) enables the evaluation of the core poloidal rotation (upol ) through the inboard-outboard asymmetry of the toroidal rotation with an accuracy of 0.5 to 1 km s-1 . Using this technique, the total plasma flow has been measured in Ohmic L-mode plasmas across the transition from the linear to saturated ohmic confinement (LOC-SOC) regimes. The core poloidal rotation of the plasma around mid-radius is found to be always in the ion diamagnetic direction, in disagreement with neoclassical (NC) predictions. The edge rotation is found to be electron-directed and consistent with NC codes. This measurement provides as well the missing ingredient to evaluate the core E×B velocity (uE×B ) from data only, which can then be compared to measurements of the perpendicular velocity of the turbulent fluctuations (u\\perp ) to gain information on the turbulent phase velocity (vph ). The non neoclassical upol from CXRS leads to good agreement between uE×B and u\\perp indicating that vph is small and at similar values as found with gyrokinetic simulations. Moreover, the data shows a shift of vph in the ion-diamagnetic direction at the edge after the transition from LOC to SOC consistent with a change in the dominant turbulence regime. The upgrade of the core CXRS system provides as well a deeper insight into the intrinsic rotation. This paper shows that the reversal of the core toroidal rotation occurs clearly after the LOC-SOC transition and concomitant with the peaking of the electron density.

  17. Improving work control systems: The core team concept

    International Nuclear Information System (INIS)

    Jorgensen, M.D.; Simpson, W.W.

    1996-01-01

    The improved work control system at the Idaho Chemical Processing Plant minimizes review and approval time, maximizes field work time, and maintains full compliance with applicable requirements. The core team method gives ownership and accountability to knowledgeable individuals, and the teams use sophisticated scheduling techniques to improve information sharing and cost control and to establish accurate roll-up master schedules

  18. Studies on spheromak plasma production by external-flux-core method, (2)

    International Nuclear Information System (INIS)

    Arata, Masanori; Katsurai, Makoto

    1984-01-01

    The spheromak technique, one of magnetic plasma containment techniques, has such arrangement of magnetic fields that the toroidal magnetic field is produced by the poloidal current flowing in plasma, and the poloidal magnetic field is produced by the toroidal current in plasma and the current in external coils. The authors proposed external flux core method as the technique of plasma formation by this spheromak method, in which the toroidal magnetic field is injected by the discharge using electrodes, whereas the poloidal magnetic field is injected by induction discharge without electrode. Its fundamental action was analyzed by computer simulation and confirmed by experiment. In this study, the behavior of the spheromak plasma produced was investigated in detail and summarized. The contents were the measurement of the spheromak configuration produced and the estimation of plasma parameters. The experimental setup, the principle of action, and the experimental results of magnetic field distribution obtained by a magnetic probe, ion current measured by an electrostatic probe, electron temperature by spectroscopic measurement and the behavior of spheromak plasma observed with an image converter camera are reported. (Kako, I.)

  19. Plasma position control device for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Fujita, Jun-ya; Ioki, Kimihiro

    1995-10-03

    The present invention concerns plasma position control coils having a feeder line structure not requiring high strength for the support portion. Namely, the coils are formed by twisting feeder lines extended from plasma position control coils in a vacuum vessel. The twisted feeder lines are supported using an appropriate structural member. Electromagnetic load is generated to the feeder lines being extended from the position control coils and traversing toroidal fields at a current introduction lines and at current delivery lines respectively. However, since the feeder lines have substantially spiral shape consisting of two twisted lines, the electromagnetic load and the moment caused by the generated load which are inversed to each other are off set. Accordingly, only extremely small force is exerted on the fittings which support the feeder lines. Therefore, small strength may suffice for the fittings and the gaps of mounting the fittings may be made longer. (I.S.).

  20. Plasma position control device for thermonuclear device

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Fujita, Jun-ya; Ioki, Kimihiro.

    1995-01-01

    The present invention concerns plasma position control coils having a feeder line structure not requiring high strength for the support portion. Namely, the coils are formed by twisting feeder lines extended from plasma position control coils in a vacuum vessel. The twisted feeder lines are supported using an appropriate structural member. Electromagnetic load is generated to the feeder lines being extended from the position control coils and traversing toroidal fields at a current introduction lines and at current delivery lines respectively. However, since the feeder lines have substantially spiral shape consisting of two twisted lines, the electromagnetic load and the moment caused by the generated load which are inversed to each other are off set. Accordingly, only extremely small force is exerted on the fittings which support the feeder lines. Therefore, small strength may suffice for the fittings and the gaps of mounting the fittings may be made longer. (I.S.)

  1. The CANopen Controller IP Core: Implementation, Synthesis and Test Results

    Science.gov (United States)

    Caramia, Maurizio; Bolognino, Luca; Montagna, Mario; Tosi, Pietro; Errico, Walter; Bigongiari, Franco; Furano, Gianluca

    2011-08-01

    This paper will describe the implementation and test results of the CANopen Controller IP Core (CCIPC) implemented by Thales Alenia Space and SITAEL Aerospace with the support of ESA in the frame of the EXOMARS Project. The CCIPC is a configurable VHDL implementation of the CANOPEN protocol [1]; it is foreseen to be used as CAN bus slave controller within the EXOMARS Entry Descending and Landing Demonstrato Module (EDM) and Rover Module. The CCIPC features, configuration capability, synthesis and test results will be described and the evidence of the state of maturity of this innovative IP core will be demonstrated.

  2. Magnetic configuration control of ITER plasmas

    International Nuclear Information System (INIS)

    Albanese, R.; Mattei, M.; Portone, A.; Ambrosino, G.; Artaserse, G.; Crisanti, F.; De Tommasi, G.; Fresa, R.; Sartori, F.; Villone, F.

    2007-01-01

    The aim of this paper is to present some new tools used to review the capability of the ITER Poloidal Field (PF) system in controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. Some preliminary results obtained during ongoing activities in collaboration between ENEA/CREATE and EFDA are presented. The paper is divided in two main parts devoted, respectively, to the presentation of a procedure for the PF current optimisation during the scenario, and of a software environment for the study of the PF system capabilities using the plasma linearized response. The proposed PF current optimisation procedure is then used to assess Scenario 2 design, also taking into account the presence of axisymmetric eddy currents and possible variations of poloidal beta and internal inductance. The numerical linear model based tool derived from the JET oriented eXtreme Shape Controller (XSC) tools is finally used to obtain results on the strike point sweeping in ITER

  3. Core/shell silicon/polyaniline particles via in-flight plasma-induced polymerization

    International Nuclear Information System (INIS)

    Yasar-Inceoglu, Ozgul; Mangolini, Lorenzo; Zhong, Lanlan

    2015-01-01

    Although silicon nanoparticles have potential applications in many relevant fields, there is often the need for post-processing steps to tune the property of the nanomaterial and to optimize it for targeted applications. In particular surface modification is generally necessary to both tune dispersibility of the particles in desired solvents to achieve optimal coating conditions, and to interface the particles with other materials to realize functional heterostructures. In this contribution we discuss the realization of core/shell silicon/polymer nanoparticles realized using a plasma-initiated in-flight polymerization process. Silicon particles are produced in a non-thermal plasma reactor using silane as a precursor. After synthesis they are aerodynamically injected into a second plasma reactor into which aniline vapor is introduced. The second plasma initiates the polymerization reactor leading to the formation of a 3–4 nm thick polymer shell surrounding the silicon core. The role of processing conditions on the properties of the polymeric shell is discussed. Preliminary results on the testing of this material as an anode for lithium ion batteries are presented. (paper)

  4. Core/shell silicon/polyaniline particles via in-flight plasma-induced polymerization

    Science.gov (United States)

    Yasar-Inceoglu, Ozgul; Zhong, Lanlan; Mangolini, Lorenzo

    2015-08-01

    Although silicon nanoparticles have potential applications in many relevant fields, there is often the need for post-processing steps to tune the property of the nanomaterial and to optimize it for targeted applications. In particular surface modification is generally necessary to both tune dispersibility of the particles in desired solvents to achieve optimal coating conditions, and to interface the particles with other materials to realize functional heterostructures. In this contribution we discuss the realization of core/shell silicon/polymer nanoparticles realized using a plasma-initiated in-flight polymerization process. Silicon particles are produced in a non-thermal plasma reactor using silane as a precursor. After synthesis they are aerodynamically injected into a second plasma reactor into which aniline vapor is introduced. The second plasma initiates the polymerization reactor leading to the formation of a 3-4 nm thick polymer shell surrounding the silicon core. The role of processing conditions on the properties of the polymeric shell is discussed. Preliminary results on the testing of this material as an anode for lithium ion batteries are presented.

  5. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

    International Nuclear Information System (INIS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

  6. ac power control in the Core Flow Test Loop

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1980-01-01

    This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

  7. The genetic network controlling plasma cell differentiation.

    Science.gov (United States)

    Nutt, Stephen L; Taubenheim, Nadine; Hasbold, Jhagvaral; Corcoran, Lynn M; Hodgkin, Philip D

    2011-10-01

    Upon activation by antigen, mature B cells undergo immunoglobulin class switch recombination and differentiate into antibody-secreting plasma cells, the endpoint of the B cell developmental lineage. Careful quantitation of these processes, which are stochastic, independent and strongly linked to the division history of the cell, has revealed that populations of B cells behave in a highly predictable manner. Considerable progress has also been made in the last few years in understanding the gene regulatory network that controls the B cell to plasma cell transition. The mutually exclusive transcriptomes of B cells and plasma cells are maintained by the antagonistic influences of two groups of transcription factors, those that maintain the B cell program, including Pax5, Bach2 and Bcl6, and those that promote and facilitate plasma cell differentiation, notably Irf4, Blimp1 and Xbp1. In this review, we discuss progress in the definition of both the transcriptional and cellular events occurring during late B cell differentiation, as integrating these two approaches is crucial to defining a regulatory network that faithfully reflects the stochastic features and complexity of the humoral immune response. 2011 Elsevier Ltd. All rights reserved.

  8. MATURING ECRF TECHNOLOGY FOR PLASMA CONTROL

    International Nuclear Information System (INIS)

    CALLIS, R.W.; CARY, W.P.; CHU, S.; LOANE, J.L.; ELLIS, R.A.; FELCH, K.; GORELOV, Y.A.; GRUNLOH, H.J.; HOSEA, J.; KAJIWARA, K.; LOHR, J.; LUCE, T.C.; PEAVY, J.J.; PINSKER, R.I.; PONCE, D.; PRATER, R.; SHAPIRO, M.; TEMKIN, R.J.; TOOKER, J.F.

    2002-01-01

    OAK A271 MUTURING ECRF TECHNOLOGY FOR PLASMA CONTROL. Understanding of the physics of internal transport barriers (ITBs) is being furthered by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and 2-D profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E x B flow shear rate is close to the linear growth rate of the ITG modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the 2-D profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E x B shear rate is comparable to the linear growth rates at the location of the ITB

  9. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.

    2001-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  10. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.; Voitsekhovitch, I.

    1999-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  11. Physics of plasma-wall interactions in controlled fusion

    International Nuclear Information System (INIS)

    Post, D.E.; Behrisch, R.

    1984-01-01

    In the areas of plasma physics, atomic physics, surface physics, bulk material properties and fusion experiments and theory, the following topics are presented: the plasma sheath; plasma flow in the sheath and presheath of a scrape-off layer; probes for plasma edge diagnostics in magnetic confinement fusion devices; atomic and molecular collisions in the plasma boundary; physical sputtering of solids at ion bombardment; chemical sputtering and radiation enhanced sublimation of carbon; ion backscattering from solid surfaces; implantation, retention and release of hydrogen isotopes; surface erosion by electrical arcs; electron emission from solid surfaces;l properties of materials; plasma transport near material boundaries; plasma models for impurity control experiments; neutral particle transport; particle confinement and control in existing tokamaks; limiters and divertor plates; advanced limiters; divertor tokamak experiments; plasma wall interactions in heated plasmas; plasma-wall interactions in tandem mirror machines; and impurity control systems for reactor experiments

  12. Ignition and burn control in tokamak plasmas

    International Nuclear Information System (INIS)

    Borrass, K.; Gruber, O.; Lackner, K.; Minardi, E.; Neuhauser, J.; Wilhelm, R.; Wunderlich, R.; Bromberg, L.; Cohn, D.R.

    1981-01-01

    Different schemes for the control of the thermal instability in an ignited fusion reactor are analysed by zero- and one-dimensional models. Passive stabilization methods considered are ripple-enhanced ion heat conduction, the effect of the major-radius variation of the plasma column in a time-independent vertical field, and the combination of both effects, including the spatial variation of the toroidal-ripple amplitude. Active control methods analysed are high-Q-driven operation and feedback-controlled major-radius variation following different scenarios. One-dimensional analyses taking into account only conductive losses show the existence of a single unstable mode in the energy balance, justifying, under these assumptions, the study of only global control. (author)

  13. Core-control assembly with a fixed fuel support

    International Nuclear Information System (INIS)

    Challberg, R.C.

    1993-01-01

    A core-control assembly is described comprising: a control rod having a plurality of blades; a control-rod guide tube for guiding vertical motion of said control rod; a fuel support for supporting fuel bundles separated by said blades, said fuel support having an aperture conforming to a cross section of said control rod through said blades for preventing rotational movement of said control rod to a decoupling orientation when said control rod is between a maximum power position and a minimum power position, said minimum power position being above said maximum power position, said fuel support being supported by said control-rod guide tube; control-rod drive means for controlling vertical motion of said control rod, said control-rod drive means providing for vertical motion between said maximum power position and said minimum power position, said control-rod drive means providing for vertical movement to a decoupling position, said decoupling position being no lower than said minimum power position, said decoupling position being at a level sufficient to permit said control rod to rotate to a decoupling orientation relative to said fuel support; and coupling means for coupling said control rod to said control rod drive means, said coupling means being releasable by rotational movement of said control rod to said decoupling orientation relative to said control-rod drive means

  14. Tungsten Transport in the Core of JET H-mode Plasmas, Experiments and Modelling

    Science.gov (United States)

    Angioni, Clemente

    2014-10-01

    The physics of heavy impurity transport in tokamak plasmas plays an essential role towards the achievement of practical fusion energy. Reliable predictions of the behavior of these impurities require the development of realistic theoretical models and a complete understanding of present experiments, against which models can be validated. Recent experimental campaigns at JET with the ITER-like wall, with a W divertor, provide an extremely interesting and relevant opportunity to perform this combined experimental and theoretical research. Theoretical models of both neoclassical and turbulent transport must consistently include the impact of any poloidal asymmetry of the W density to enable quantitative predictions of the 2D W density distribution over the poloidal cross section. The agreement between theoretical predictions and experimentally reconstructed 2D W densities allows the identification of the main mechanisms which govern W transport in the core of JET H-mode plasmas. Neoclassical transport is largely enhanced by centrifugal effects and the neoclassical convection dominates, leading to central accumulation in the presence of central peaking of the density profiles and insufficiently peaked ion temperature profiles. The strength of the neoclassical temperature screening is affected by poloidal asymmetries. Only around mid-radius, turbulent diffusion offsets neoclassical transport. Consistently with observations in other devices, ion cyclotron resonance heating in the plasma center can flatten the electron density profile and peak the ion temperature profile and provide a means to reverse the neoclassical convection. MHD activity may hamper or speed up the accumulation process depending on mode number and plasma conditions. Finally, the relationship of JET results to a parallel modelling activity of the W behavior in the core of ASDEX Upgrade plasmas is presented. This project has received funding from the European Union's Horizon 2020 research and innovation

  15. Architecture of WEST plasma control system

    International Nuclear Information System (INIS)

    Ravenel, N.; Nouailletas, R.; Barana, O.; Brémond, S.; Moreau, P.; Guillerminet, B.; Balme, S.; Allegretti, L.; Mannori, S.

    2014-01-01

    To operate advanced plasma scenario (long pulse with high stored energy) in present and future tokamak devices under safe operation conditions, the control requirements of the plasma control system (PCS) leads to the development of advanced feedback control and real time handling exceptions. To develop these controllers and these exceptions handling strategies, a project aiming at setting up a flight simulator has started at CEA in 2009. Now, the new WEST (W Environment in Steady-state Tokamak) project deals with modifying Tore Supra into an ITER-like divertor tokamak. This upgrade impacts a lot of systems including Tore Supra PCS and is the opportunity to improve the current PCS architecture to implement the previous works and to fulfill the needs of modern tokamak operation. This paper is dealing with the description of the architecture of WEST PCS. Firstly, the requirements will be presented including the needs of new concepts (segments configuration, alternative (or backup) scenario, …). Then, the conceptual design of the PCS will be described including the main components and their functions. The third part will be dedicated to the proposal RT framework and to the technologies that we have to implement to reach the requirements

  16. Plasma actuators for bluff body flow control

    Science.gov (United States)

    Kozlov, Alexey V.

    The aerodynamic plasma actuators have shown to be efficient flow control devices in various applications. In this study the results of flow control experiments utilizing single dielectric barrier discharge plasma actuators to control flow separation and unsteady vortex shedding from a circular cylinder in cross-flow are reported. This work is motivated by the need to reduce landing gear noise for commercial transport aircraft via an effective streamlining created by the actuators. The experiments are performed at Re D = 20,000...164,000. Circular cylinders in cross-flow are chosen for study since they represent a generic flow geometry that is similar in all essential aspects to a landing gear oleo or strut. The minimization of the unsteady flow separation from the models and associated large-scale wake vorticity by using actuators reduces the radiated aerodynamic noise. Using either steady or unsteady actuation at ReD = 25,000, Karman shedding is totally eliminated, turbulence levels in the wake decrease significantly and near-field sound pressure levels are reduced by 13.3 dB. Unsteady actuation at an excitation frequency of St D = 1 is found to be most effective. The unsteady actuation also has the advantage that total suppression of shedding is achieved for a duty cycle of only 25%. However, since unsteady actuation is associated with an unsteady body force and produces a tone at the actuation frequency, steady actuation is more suitable for noise control applications. Two actuation strategies are used at ReD = 82,000: spanwise and streamwise oriented actuators. Near field microphone measurements in an anechoic wind tunnel and detailed study of the near wake using LDA are presented in the study. Both spanwise and streamwise actuators give nearly the same noise reduction level of 11.2 dB and 14.2 dB, respectively, and similar changes in the wake velocity profiles. The contribution of the actuator induced noise is found to be small compared to the natural shedding

  17. An advanced plasma control system for Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Wijnands, T.; Martin, G.

    1996-01-01

    First results on plasma control with the new plasma control system of Tore Supra are presented. The system has been especially designed for long pulse operation: plasmas are controlled on reference signals, which can be varied in real time by using diagnostic measurements. On line determination of the global plasma equilibrium has enabled new operation scenarios in which both the power from the poloidal field generators and the total Lower Hybrid (LH) power are used to control the plasma. Experiments with feedback control of the safety factor on the plasma boundary, control of the LH driven current, control of the flux on the plasma boundary and control of the internal inductance are discussed. (author). 12 refs.

  18. An advanced plasma control system for Tore Supra

    International Nuclear Information System (INIS)

    Wijnands, T.; Martin, G.

    1996-01-01

    First results on plasma control with the new plasma control system of Tore Supra are presented. The system has been especially designed for long pulse operation: plasmas are controlled on reference signals, which can be varied in real time by using diagnostic measurements. On line determination of the global plasma equilibrium has enabled new operation scenarios in which both the power from the poloidal field generators and the total Lower Hybrid (LH) power are used to control the plasma. Experiments with feedback control of the safety factor on the plasma boundary, control of the LH driven current, control of the flux on the plasma boundary and control of the internal inductance are discussed. (author)

  19. Plasma-wall interaction data needs critical to a Burning Core Experiment (BCX)

    International Nuclear Information System (INIS)

    1985-11-01

    The Division of Development and Technology has sponsored a four day US-Japan workshop ''Plasma-Wall Interaction Data Needs Critical to a Burning Core Experiment (BCX)'', held at Sandia National Laboratories, Livermore, California on June 24 to 27, 1985. The workshop, which brought together fifty scientists and engineers from the United States, Japan, Germany, and Canada, considered the plasma-material interaction and high heat flux (PMI/HHF) issues for the next generation of magnetic fusion energy devices, the Burning Core Experiment (BCX). Materials options were ranked, and a strategy for future PMI/HHF research was formulated. The foundation for international collaboration and coordination of this research was also established. This volume contains the first two of the five technical sessions. The first one being the BCX overview, the second on the BCX candidate materials. The remaining three sessions in volume two are on the plasma materials interaction issues, research facilities and small working group meeting on graphite, beryllium, advanced materials and future collaborations

  20. Results from core-edge experiments in high Power, high performance plasmas on DIII-D

    Directory of Open Access Journals (Sweden)

    T.W. Petrie

    2017-08-01

    Full Text Available Significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q⊥P ∝ [PSOL x IP] 0.92 for PSOL= 8−19MW and IP= 1.0–1.4MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-D plasmas may be problematical at high power and H98 (≥ 1.5 due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q⊥P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot” but also that heating near the slot opening is a significant source for impurity contamination of the core.

  1. Plasma-wall interaction data needs critical to a Burning Core Experiment (BCX)

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    The Division of Development and Technology has sponsored a four day US-Japan workshop ''Plasma-Wall Interaction Data Needs Critical to a Burning Core Experiment (BCX)'', held at Sandia National Laboratories, Livermore, California on June 24 to 27, 1985. The workshop, which brought together fifty scientists and engineers from the United States, Japan, Germany, and Canada, considered the plasma-material interaction and high heat flux (PMI/HHF) issues for the next generation of magnetic fusion energy devices, the Burning Core Experiment (BCX). Materials options were ranked, and a strategy for future PMI/HHF research was formulated. The foundation for international collaboration and coordination of this research was also established. This volume contains the last three of the five technical sessions. The first of the three is on plasma materials interaction issues, the second is on research facilities and the third is from smaller working group meetings on graphite, beryllium, advanced materials and future collaborations.

  2. Numerical simulation and optimal control in plasma physics

    International Nuclear Information System (INIS)

    Blum, J.

    1989-01-01

    The topics covered in this book are: A free boundary problem: the axisymmetric equilibrium of the plasma in a Tokamak; Static control of the plasma boundary by external currents; Existence and control of a solution to the equilibrium problem in a simple case; Study of equilibrium solution branches and application to the stability of horizontal displacements; Identification of the plasma boundary and plasma current density from magnetic measurements; Evolution of the equilibrium at the diffusion time scale; Evolution of the equilibrium of a high aspect-ratio circular plasma; Stability and control of the horizontal displacement of the plasma

  3. The ITER Plasma Control System Simulation Platform

    International Nuclear Information System (INIS)

    Walker, M.L.; Ambrosino, G.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W.; Welander, A.S.; Winter, A.

    2015-01-01

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  4. The ITER Plasma Control System Simulation Platform

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.L., E-mail: walker@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Ambrosino, G.; De Tommasi, G. [CREATE/Università di Napoli Federico II, Napoli (Italy); Humphreys, D.A. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Napoli (Italy); Neu, G.; Rapson, C.J.; Raupp, G.; Treutterer, W. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Welander, A.S. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Winter, A. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2015-10-15

    Highlights: • A development and test environment called PCSSP has been constructed for the ITER PCS. • A description of requirements and use cases, a final design and software architecture design, users guide, and a prototype implementation have been delivered. • The prototype implementation was demonstrated at IO in December of 2013. • PCSSP will be deployed for alpha testing to the IO, the development group, and selected other ITER partners upon completion of the next development phase. - Abstract: The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an “alpha” version of PCSSP and preparation for the next development phase. High

  5. The plasma-wall interaction region: a key low temperature plasma for controlled fusion

    International Nuclear Information System (INIS)

    Counsell, G F

    2002-01-01

    The plasma-wall interaction region of a fusion device provides the interface between the hot core plasma and the material surfaces. To obtain acceptably low levels of erosion from these surfaces requires most of the power leaving the core to be radiated. This is accomplished in existing devices by encouraging plasma detachment, in which the hot plasma arriving in the region is cooled by volume recombination and ion-neutral momentum transfer with a dense population of neutrals recycled from the surface. The result is a low temperature (1 eV e e >10 19 m -3 ) but weakly ionized (n 0 >10 20 m -3 , n e /n 0 <0.1) plasma found nowhere else in the fusion environment. This plasma provides many of the conditions found in industrial plasmas exploiting plasma chemistry and the presence of carbon in the region (in the form of carbon-fibre composite used in the plasma facing materials) can result in the formation of deposited hydrocarbon films. The plasma-wall interaction region is therefore among the most difficult in fusion to model, requiring an understanding of atomic, molecular and surface physics issues

  6. Transforming the ASDEX Upgrade discharge control system to a general-purpose plasma control platform

    International Nuclear Information System (INIS)

    Treutterer, Wolfgang; Cole, Richard; Gräter, Alexander; Lüddecke, Klaus; Neu, Gregor; Rapson, Christopher; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas

    2015-01-01

    Highlights: • Control framework split in core and custom part. • Core framework deployable in other fusion device environments. • Adaptible through customizable modules, plug-in support and generic interfaces. - Abstract: The ASDEX Upgrade Discharge Control System DCS is a modern and mature product, originally designed to regulate and supervise ASDEX Upgrade Tokamak plasma operation. In its core DCS is based on a generic, versatile real-time software framework with a plugin architecture that allows to easily combine, modify and extend control function modules in order to tailor the system to required features and let it continuously evolve with the progress of an experimental fusion device. Due to these properties other fusion experiments like the WEST project have expressed interest in adopting DCS. For this purpose, essential parts of DCS must be unpinned from the ASDEX Upgrade environment by exposure or introduction of generalised interfaces. Re-organisation of DCS modules allows distinguishing between intrinsic framework core functions and device-specific applications. In particular, DCS must be prepared for deployment in different system environments with their own realisations for user interface, pulse schedule preparation, parameter server, time and event distribution, diagnostic and actuator systems, network communication and data archiving. The article explains the principles of the revised DCS structure, derives the necessary interface definitions and describes major steps to achieve the separation between general-purpose framework and fusion device specific components.

  7. Transforming the ASDEX Upgrade discharge control system to a general-purpose plasma control platform

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, Wolfgang, E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Cole, Richard [Unlimited Computer Systems, Seeshaupter Str. 15, 82393 Iffeldorf (Germany); Gräter, Alexander [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Lüddecke, Klaus [Unlimited Computer Systems, Seeshaupter Str. 15, 82393 Iffeldorf (Germany); Neu, Gregor; Rapson, Christopher; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Highlights: • Control framework split in core and custom part. • Core framework deployable in other fusion device environments. • Adaptible through customizable modules, plug-in support and generic interfaces. - Abstract: The ASDEX Upgrade Discharge Control System DCS is a modern and mature product, originally designed to regulate and supervise ASDEX Upgrade Tokamak plasma operation. In its core DCS is based on a generic, versatile real-time software framework with a plugin architecture that allows to easily combine, modify and extend control function modules in order to tailor the system to required features and let it continuously evolve with the progress of an experimental fusion device. Due to these properties other fusion experiments like the WEST project have expressed interest in adopting DCS. For this purpose, essential parts of DCS must be unpinned from the ASDEX Upgrade environment by exposure or introduction of generalised interfaces. Re-organisation of DCS modules allows distinguishing between intrinsic framework core functions and device-specific applications. In particular, DCS must be prepared for deployment in different system environments with their own realisations for user interface, pulse schedule preparation, parameter server, time and event distribution, diagnostic and actuator systems, network communication and data archiving. The article explains the principles of the revised DCS structure, derives the necessary interface definitions and describes major steps to achieve the separation between general-purpose framework and fusion device specific components.

  8. Plasma physics for controlled fusion. 2. ed.

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Kenro

    2016-08-01

    This new edition presents the essential theoretical and analytical methods needed to understand the recent fusion research of tokamak and alternate approaches. The author describes magnetohydrodynamic and kinetic theories of cold and hot plasmas in detail. The book covers new important topics for fusion studies such as plasma transport by drift turbulence, which depend on the magnetic configuration and zonal flows. These are universal phenomena of microturbulence. They can modify the onset criterion for turbulent transport, instabilities driven by energetic particles as well as alpha particle generation and typical plasma models for computer simulation. The fusion research of tokamaks with various new versions of H modes are explained. The design concept of ITER, the international tokamak experimental reactor, is described for inductively driven operations as well as steady-state operations using non-inductive drives. Alternative approaches of reversed-field pinch and its relaxation process, stellator including quasi-symmetric system, open-end system of tandem mirror and inertial confinement are also explained. Newly added and updated topics in this second edition include zonal flows, various versions of H modes, and steady-state operations of tokamak, the design concept of ITER, the relaxation process of RFP, quasi-symmetric stellator, and tandem mirror. The book addresses graduate students and researchers in the field of controlled fusion.

  9. Plasma physics for controlled fusion. 2. ed.

    International Nuclear Information System (INIS)

    Miyamoto, Kenro

    2016-01-01

    This new edition presents the essential theoretical and analytical methods needed to understand the recent fusion research of tokamak and alternate approaches. The author describes magnetohydrodynamic and kinetic theories of cold and hot plasmas in detail. The book covers new important topics for fusion studies such as plasma transport by drift turbulence, which depend on the magnetic configuration and zonal flows. These are universal phenomena of microturbulence. They can modify the onset criterion for turbulent transport, instabilities driven by energetic particles as well as alpha particle generation and typical plasma models for computer simulation. The fusion research of tokamaks with various new versions of H modes are explained. The design concept of ITER, the international tokamak experimental reactor, is described for inductively driven operations as well as steady-state operations using non-inductive drives. Alternative approaches of reversed-field pinch and its relaxation process, stellator including quasi-symmetric system, open-end system of tandem mirror and inertial confinement are also explained. Newly added and updated topics in this second edition include zonal flows, various versions of H modes, and steady-state operations of tokamak, the design concept of ITER, the relaxation process of RFP, quasi-symmetric stellator, and tandem mirror. The book addresses graduate students and researchers in the field of controlled fusion.

  10. Special processor for in-core control systems

    International Nuclear Information System (INIS)

    Golovanov, M.N.; Duma, V.R.; Levin, G.L.; Mel'nikov, A.V.; Polikanin, A.V.; Filatov, V.P.

    1978-01-01

    The BUTs-20 special processor is discussed, designed to control the units of the in-core control equipment which are incorporated into the VECTOR communication channel, and to provide preliminary data processing prior to computer calculations. A set of instructions and flowsheet of the processor, organization of its communication with memories and other units of the system are given. The processor components: a control unit and an arithmetic logical unit are discussed. It is noted that the special processor permits more effective utilization of the computer time

  11. LIDAR TS for ITER core plasma. Part II: simultaneous two wavelength LIDAR TS

    Science.gov (United States)

    Gowers, C.; Nielsen, P.; Salzmann, H.

    2017-12-01

    We have shown recently, and in more detail at this conference (Salzmann et al) that the LIDAR approach to ITER core TS measurements requires only two mirrors in the inaccessible port plug area of the machine. This leads to simplified and robust alignment, lower risk of mirror damage by plasma contamination and much simpler calibration, compared with the awkward and vulnerable optical geometry of the conventional imaging TS approach, currently under development by ITER. In the present work we have extended the simulation code used previously to include the case of launching two laser pulses, of different wavelengths, simultaneously in LIDAR geometry. The aim of this approach is to broaden the choice of lasers available for the diagnostic. In the simulation code it is assumed that two short duration (300 ps) laser pulses of different wavelengths, from an Nd:YAG laser are launched through the plasma simultaneously. The temperature and density profiles are deduced in the usual way but from the resulting combined scattered signals in the different spectral channels of the single spectrometer. The spectral response and quantum efficiencies of the detectors used in the simulation are taken from catalogue data for commercially available Hamamatsu MCP-PMTs. The response times, gateability and tolerance to stray light levels of this type of photomultiplier have already been demonstrated in the JET LIDAR system and give sufficient spatial resolution to meet the ITER specification. Here we present the new simulation results from the code. They demonstrate that when the detectors are combined with this two laser, LIDAR approach, the full range of the specified ITER core plasma Te and ne can be measured with sufficient accuracy. So, with commercially available detectors and a simple modification of a Nd:YAG laser similar to that currently being used in the design of the conventional ITER core TS design mentioned above, the ITER requirements can be met.

  12. Impact of wall materials and seeding gases on the pedestal and on core plasma performance

    Directory of Open Access Journals (Sweden)

    E. Wolfrum

    2017-08-01

    Full Text Available Plasmas in machines with all metal plasma facing components have a lower Zeff, less radiation cooling in the scrape-off layer and divertor regions and are prone to impurity accumulation in the core. Higher gas puff and the seeding of low-Z impurities are applied to prevent impurity accumulation, to increase the frequency of edge localised modes and to cool the divertor. A lower power threshold for the transition from low-confinement mode to high confinement mode has been found in all metal wall machines when compared to carbon wall machines. The application of lithium before or during discharges can lead to ELM free H-modes. The seeding of high-Z impurities increases core radiation, reduces the power flux across the separatrix and, if applied in the right amount, does not lead to deterioration of the confinement. All these effects have in common that they can often be explained by the shape or position of the density profile. Not only the peakedness of the density profile in the core but also the position of the edge pressure gradient influences global confinement. It is shown how (i ionisation in the pedestal region due to higher reflection of deuterium from high-Z walls, (ii reduced recycling in consequence of lithium wall conditioning, (iii the fostering of edge modes with lithium dropping, (iv increased gas puff and (v the cooling of the scrape-off layer by medium-Z impurities such as nitrogen affect the edge density profile. The consequence is a shift in the pressure profile relative to the separatrix, leading to improved pedestal stability of H-mode plasmas when the direction is inwards.

  13. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

    2005-01-01

    Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

  14. XFEL resonant photo-pumping of dense plasmas and dynamic evolution of autoionizing core hole states

    Science.gov (United States)

    Rosmej, F. B.; Moinard, A.; Renner, O.; Galtier, E.; Lee, J. J.; Nagler, B.; Heimann, P. A.; Schlotter, W.; Turner, J. J.; Lee, R. W.; Makita, M.; Riley, D.; Seely, J.

    2016-03-01

    Similarly to the case of LIF (Laser-Induced Fluorescence), an equally revolutionary impact to science is expected from resonant X-ray photo-pumping. It will particularly contribute to a progress in high energy density science: pumped core hole states create X-ray transitions that can escape dense matter on a 10 fs-time scale without essential photoabsorption, thus providing a unique possibility to study matter under extreme conditions. In the first proof of principle experiment at the X-ray Free Electron Laser LCLS at SCLAC [Seely, J., Rosmej, F.B., Shepherd, R., Riley, D., Lee, R.W. Proposal to Perform the 1st High Energy Density Plasma Spectroscopic Pump/Probe Experiment”, approved LCLS proposal L332 (2010)] we have successfully pumped inner-shell X-ray transitions in dense plasmas. The plasma was generated with a YAG laser irradiating solid Al and Mg targets attached to a rotating cylinder. In parallel to the optical laser beam, the XFEL was focused into the plasma plume at different delay times and pump energies. Pumped X-ray transitions have been observed with a spherically bent crystal spectrometer coupled to a Princeton CCD. By using this experimental configuration, we have simultaneously achieved extremely high spectral (λ/δλ ≈ 5000) and spatial resolution (δx≈70 μm) while maintaining high luminosity and a large spectral range covered (6.90 - 8.35 Å). By precisely measuring the variations in spectra emitted from plasma under action of XFEL radiation, we have successfully demonstrated transient X- ray pumping in a dense plasma.

  15. Radiation pattern of open ended waveguide in air core surrounded by annular plasma column

    International Nuclear Information System (INIS)

    Sharma, D.R.; Verma, J.S.

    1977-01-01

    Radiation pattern of open ended waveguide excited in circular symmetric mode (TM 01 ) in an air core having central conductor and surrounded by an annular plasma column is studied. The field distribution at the open end of the waveguide is considered to be equivalent to the vector sum of magnetic current rings of various radii, ranging from the outer radius of the inner conductor to the inner radius of the outer conductor of the waveguide at the open end. The radiation field is obtained as a vector sum of field components due to individual rings of current. Such a configuration gives rise to multiple narrow radiation beams away from the critical angle. (author)

  16. Evidence of stochastic region near a rational surface in core plasmas of LHD

    International Nuclear Information System (INIS)

    Ida, K.; Tamura, N.; Tuchiya, H.

    2010-11-01

    Clear evidence of stochastization of the magnetic surfaces near a rational surface is observed in the core plasma with weak magnetic shear in the Large Helical Device (LHD) by applying heat pulses driven by modulated electron cyclotron heating (MECH). The stochastization of the magnetic surfaces is confirmed by the observation of flattening of electron temperature (T e ) profiles and very fast propagation of the heat pulse, which is in contrast to the slow heat pulse propagation observed in the T e flat region of the nested magnetic island. (author)

  17. Large enhanced dielectric permittivity in polyaniline passivated core-shell nano magnetic iron oxide by plasma polymerization

    Energy Technology Data Exchange (ETDEWEB)

    Joy, Lija K.; Sooraj, V.; Sethulakshmi, N.; Anantharaman, M. R., E-mail: mraiyer@yahoo.com [Department of Physics, Cochin University of Science and Technology, Cochin-682022, Kerala (India); Sajeev, U. S. [Department of Physics, Government College, Kottayam-686613, Kerala (India); Nair, Swapna S. [Department of Physics, School of Mathematical and Physical Sciences, Central University of Kerala, Kasargode-671123, Kerala (India); Narayanan, T. N. [CSIR-Central Electrochemical Research Institute, Karaikkudi-630006, Tamil Nadu (India); Ajayan, P. M. [Department of Material Science and Nano Engineering, Rice University, 6100 Main Street, Houston, Texas 7700 (United States)

    2014-03-24

    Commercial samples of Magnetite with size ranging from 25–30 nm were coated with polyaniline by using radio frequency plasma polymerization to achieve a core shell structure of magnetic nanoparticle (core)–Polyaniline (shell). High resolution transmission electron microscopy images confirm the core shell architecture of polyaniline coated iron oxide. The dielectric properties of the material were studied before and after plasma treatment. The polymer coated magnetite particles exhibited a large dielectric permittivity with respect to uncoated samples. The dielectric behavior was modeled using a Maxwell–Wagner capacitor model. A plausible mechanism for the enhancement of dielectric permittivity is proposed.

  18. Brazilian programme for plasma physics and controlled thermonuclear fusion

    International Nuclear Information System (INIS)

    Chian, A.C.L.; Reusch, M.F.; Nascimento, I.C.; Pantuso-Sudano, J.

    1992-01-01

    A proposal for a National Programme of Plasma Physics and Controlled Thermonuclear Fusion in Brazil is presented, aimimg the dissemination of the researchers thought in plasma physics for the national authorities and the scientific community. (E.O.)

  19. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

    Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)

    2013-10-15

    Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].

  20. Effect of density control and impurity transport on internal transport barrier formation in tokamak plasma

    International Nuclear Information System (INIS)

    Yamakami, Tomoyuki; Fujita, Takaaki; Arimoto, Hideki; Yamazaki, Kozo

    2014-01-01

    In future fusion reactors, density control, such as fueling by pellet injection, is an effective method to control the formation of the internal transport barrier (ITB) in reversed magnetic shear plasma, which can improve plasma performance. On the other hand, an operation with ITB can cause accumulation of impurities inside the core ITB region. We studied the relation between pellet injection and ITB formation and the effect of impurity transport on the core of ITB for tokamak plasmas by using the toroidal transport analysis linkage. For ITB formation, we showed that the pellet has to be injected beyond the position where the safety factor q takes the minimum value. We confirmed that the accumulation of impurities causes the attenuation of ITB owing to radiation loss inside the ITB region. Moreover, in terms of the divertor heat flux reduction by impurity gas, the line radiation loss is high for high-Z noble gas impurities, such as Kr, whereas factor Q decreases slightly. (author)

  1. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  2. Control rod repositioning considerations in core design analysis

    International Nuclear Information System (INIS)

    Armstrong, B.C.; Buechel, R.J.

    1990-01-01

    Control rod repositioning is a method for minimizing rod cluster control assembly (RCCA) wear in the upper internals area where the guide cards interface with the rodlets of the RCCAs. A number of utilities have implemented strategies for rod repositioning, which often has no impact on the nuclear analysis for cases where the control rods are never repositioned into the active fuel. Other strategies involve repositioning the control rods several steps into the active fuel. The impact of this type of repositioning on the axial power shape and consequently the total peaking factor F Q T varies, depending on the method in which the repositioning is implemented at the plant. Operating for long periods with all the control and shutdown rods inserted several steps in the active fuel followed by withdrawing them fully from the core results in a shifting of the power distribution toward the top of the core and must be accounted for in the design analysis. On the other hand, an optional plan for control rod repositioning that considers margins available in related design parameters can be devised that minimizes the effects of the repositioning for the reload. This paper summarizes a rod repositioning strategy implemented for a recent reload and some calculated power shape results for this strategy and other scenarios

  3. Plasma physics and controlled nuclear fusion research

    International Nuclear Information System (INIS)

    1980-01-01

    Full text: During the last decade, growing efforts have been devoted to studying the possible forms an electricity-producing thermonuclear reactor might take and the various technical problems that will have to be overcome. Previous IAEA Conferences took place in Salzburg (1961), Culham (1965), Novosibirsk (1968), Madison (1971), Tokyo (1974), Berchtesgaden (1976) and Innsbruck (1978) The exchange of information that has characterized this series of meetings is an important example of international co-operation and has contributed substantially to progress in controlled fusion research. The results of experiments in major research establishments, as well as the growing scientific insights in the field of plasma physics, give hope that the realization of nuclear fusion will be made possible on a larger scale and beyond the laboratory stage by the end of this century. The increase of the duration of existing tokamak discharges requires solution of the impurity control problem. First results from the new big machines equipped with the poloidal divertor recently came into operation. PDX (USA) and ASDEX (F.R. of Germany) show that various divertor configurations can be established and maintained and that the divertors function in the predicted manner. The reduction of high-Z impurities on these machines by a factor 10 was achieved. As a result of extensive research on radio-frequency (RF) plasma heating on tokamaks: PLT (USA), TFR (France), JFT-2 (Japan), the efficiency of this attractive method of plasma heating comparable to neutral beam heating was demonstrated. It was shown that the density of the input power of about 5-10 kW/cm 2 is achievable and this limit is high enough for application to reactor-like machines. One of the inspiring results reported at the conference was the achievement of value (the ratio of plasma pressure to magnetic field pressure) of ∼ 3% on tokamaks T-11 (USSR) and ISX-B (USA). It is important to note that this value exceeds the

  4. Plasma equilibrium control during slow plasma current quench with avoidance of plasma-wall interaction in JT-60U

    Science.gov (United States)

    Yoshino, R.; Nakamura, Y.; Neyatani, Y.

    1997-08-01

    In JT-60U a vertical displacement event (VDE) is observed during slow plasma current quench (Ip quench) for a vertically elongated divertor plasma with a single null. The VDE is generated by an error in the feedback control of the vertical position of the plasma current centre (ZJ). It has been perfectly avoided by improving the accuracy of the ZJ measurement in real time. Furthermore, plasma-wall interaction has been avoided successfully during slow Ip quench owing to the good performance of the plasma equilibrium control system

  5. Plasma control for efficient extreme ultra-violet source

    International Nuclear Information System (INIS)

    Takahashi, Kensaku; Nakajima, Mitsuo; Kawamura, Tohru; Shiho, Makoto; Hotta, Eiki; Horioka, Kazuhiko

    2008-01-01

    To generate a high efficiency extreme-ultraviolet (EUV) source, effects of plasma shape for controlling radiative plasmas based on xenon capillary discharge are experimentally investigated. The radiation characteristics observed via tapered capillary discharge are compared with those of straight one. From the comparison, the long emission period and different plasma behaviors of tapered capillary discharge are confirmed. This means that control of the plasma geometry is effective for prolonging the EUV emission period. This result also indicates that the plasma shape control seems to have a potential for enhancing the conversion efficiency. (author)

  6. Sensitivity of control times in function of core parameters and oscillations control in thermal nuclear systems

    International Nuclear Information System (INIS)

    Amorim, E.S. do; D'Oliveira, A.B.; Galvao, O.B.; Oyama, K.

    1981-03-01

    Sensitivity of control times to variation of a thermal reactor core parameters is defined by suitable changes in the power coefficient, core size and fuel enrichment. A control strategy is developed based on control theory concepts and on considerations of the physics of the problem. Digital diffusion theory simulation is described which tends to verify the control concepts considered, face dumped oscillations introduced in one thermal nuclear power system. The effectivity of the control actions, in terms of eliminating oscillations, provided guidelines for the working-group engaged in the analysis of the control rods and its optimal performance. (Author) [pt

  7. Developments in plasma physics and controlled fusion

    International Nuclear Information System (INIS)

    Thompson, W.B.

    1980-01-01

    Some developments in plasma physics over the past twenty years are considered from the theoretical physics standpoint under the headings; oscillations, waves and instabilities, plasma turbulence, basic kinetic theory, and developments in fusion. (UK)

  8. Core radial electric field and transport in Wendelstein 7-X plasmas

    Science.gov (United States)

    Pablant, N. A.; Langenberg, A.; Alonso, A.; Beidler, C. D.; Bitter, M.; Bozhenkov, S.; Burhenn, R.; Beurskens, M.; Delgado-Aparicio, L.; Dinklage, A.; Fuchert, G.; Gates, D.; Geiger, J.; Hill, K. W.; Höfel, U.; Hirsch, M.; Knauer, J.; Krämer-Flecken, A.; Landreman, M.; Lazerson, S.; Maaßberg, H.; Marchuk, O.; Massidda, S.; Neilson, G. H.; Pasch, E.; Satake, S.; Svennson, J.; Traverso, P.; Turkin, Y.; Valson, P.; Velasco, J. L.; Weir, G.; Windisch, T.; Wolf, R. C.; Yokoyama, M.; Zhang, D.; W7-X Team

    2018-02-01

    The results from the investigation of neoclassical core transport and the role of the radial electric field profile (Er) in the first operational phase of the Wendelstein 7-X (W7-X) stellarator are presented. In stellarator plasmas, the details of the Er profile are expected to have a strong effect on both the particle and heat fluxes. Investigation of the radial electric field is important in understanding neoclassical transport and in validation of neoclassical calculations. The radial electric field is closely related to the perpendicular plasma flow (u⊥) through the force balance equation. This allows the radial electric field to be inferred from measurements of the perpendicular flow velocity, which can be measured using the x-ray imaging crystal spectrometer and correlation reflectometry diagnostics. Large changes in the perpendicular rotation, on the order of Δu⊥˜ 5 km/s (ΔEr ˜ 12 kV/m), have been observed within a set of experiments where the heating power was stepped down from 2 MW to 0.6 MW. These experiments are examined in detail to explore the relationship between heating power temperature, and density profiles and the radial electric field. Finally, the inferred Er profiles are compared to initial neoclassical calculations based on measured plasma profiles. The results from several neoclassical codes, sfincs, fortec-3d, and dkes, are compared both with each other and the measurements. These comparisons show good agreement, giving confidence in the applicability of the neoclassical calculations to the W7-X configuration.

  9. Evaluation of compatibility of flowing liquid lithium curtain for blanket with core plasma in fusion reactors

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua; Peng Lilin; Yan Jiancheng

    2003-01-01

    A global model analysis of the compatibility of flowing liquid lithium curtain for blanket with core plasma has been performed. The relationships between the surface temperature of lithium curtain and mean effective plasma charges, fuel dilution and produced fusion power have been obtained. Results show that under normal circumstances, the evaporation of liquid lithium does not affect Z eff seriously, but affects fuel dilution and fusion power sensitively. The authors have investigated the relationships between the flow velocity of liquid lithium and its surface temperature rise based on the conditions of the option II of the fusion experimental breeder (FEB-E) design with reversed shear configuration and fairly high power density. The authors concluded that the effects of evaporation from liquid lithium curtain for FEB-E on plasma are negligible even if the flow velocity of liquid lithium is as low as 0.5 m·s -1 . Finally, the sputtering yield of liquid lithium saturated by hydrogen isotopes is briefly discussed

  10. Advances in integrated plasma control on DIII-D

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Hahn, S.H.; Humphreys, D.A.; In, Y.; Johnson, R.D.; Kim, J.S.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Welander, A.S.; Xiao, B.

    2007-01-01

    The DIII-D advanced tokamak physics program requires extremely high performance from the DIII-D plasma control system, including simultaneous accurate regulation of plasma shape, stored energy, density and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of new model-based plasma controllers on DIII-D. We also describe the development of the control software, hardware, and model-based control algorithms for the superconducting EAST and KSTAR tokamaks

  11. A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

    International Nuclear Information System (INIS)

    Takase, Haruhiko; Senda, Ikuo

    1999-01-01

    A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

  12. Plasma physics and controlled nuclear fusion

    International Nuclear Information System (INIS)

    Sato, Tetsuya

    1993-05-01

    The report contains the proceedings of a conference on plasma physics. A fraction of topics included MHD instabilities, magnetic confinement and plasma heating in the field of fusion plasmas, in 8 papers falling in the INIS scope have been abstracted and indexed for the INIS database. (K.A.)

  13. Control of open end plasma flow utilizing orbital stochasticity

    International Nuclear Information System (INIS)

    Hojo, Hitoshi

    1995-01-01

    It has been known that the control of plasma outside the confinement region of diverter plasma and others in a magnetic field confinement device is very important for improveing the confinement of bulk plasma. The control of plasma outside a confinement region bears two roles, one is the reduction of the thermal load on a diverter plate and others due to the plasma particles lost from the confinement region, and another is the restriction of the back flow of cold plasma and impurities generated outside the confinement region to a bulk plasma region. In this study, the new method of controlling plasma outside a confinement region called magnetic diverter is considered. To the plasma particles advancing along magnetic force lines, the reflection and capture of the plasma particles occur in the region of orbital stochasticity, and the thermal load on an end plate and the reverse flow to a bulk plasma region are restricted. The numerical computation model used regarding the particle control utilizing the orbital stochasticity and the results of calculating the orbit of plasma particles in a magnetic field are reported. (K.I.)

  14. Manufacturing of central control system of 'JT-60' a plasma feasibility experiment device

    International Nuclear Information System (INIS)

    Kondo, Ikuo; Kimura, Toyoaki; Murai, Katsuji; Iba, Daizo; Takemaru, Koichi.

    1984-01-01

    For constructing a critical-plasma-experiment apparatus JT-60, it was necessary to develop a new control system which enables to operate safely and smoothly a large scale nuclear fusion apparatus and to carry out efficient experiment. For the purpose, the total system control facility composed of such controllers as CAMAC system, timing system and protective interlock panel with multi-computer system as the core was developed. This system generalizes, keeps watch on and controls the total facilities as the key point of the control system of JT-60, and allows flexible operation control corresponding to the diversified experimental projects. At the same time, it carries out the fast real-time control of high temperature, high density plasma. In this paper, the system constitution, function and the main contents of development of the total system control facility are reported. JT-60 is constructed to attain the critical plasma condition as the premise of nuclear fusion reactors and to scientifically verify controlled nuclear fusion. Plasma expe riment will be started in April, 1985. The real-time control of plasma for carrying out high beta operation is planned, intending to develop future economical practical reactors. (Kako, I.)

  15. Control of plasma position in the CASTOR tokamak

    International Nuclear Information System (INIS)

    Valovic, M.

    1988-11-01

    A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs

  16. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    International Nuclear Information System (INIS)

    1993-12-01

    The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support

  17. Current status and prospect of plasma control system for steady-state operation on QUEST

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Zushi, Hideki; Hanada, Kazuaki; Fujisawa, Akihide; Tokunaga, Kazutoshi; Idei, Hiroshi; Nagashima, Yoshihiko; Kawasaki, Shoji; Nakashima, Hisatoshi; Higashijima, Aki

    2016-01-01

    Highlights: • Overall configuration of plasma control system on QUEST are presented. • Multi core system and reflective memories are used for the real-time control. • Hall sensors are used for the identification of plasma current and its position. • Repetitive gas fueling with the feed-back control of Hα signal is implemented. - Abstract: The plasma control system (PCS) of QUEST is developed according to the progress of QUEST project. Since one of the critical goals of the project is to achieve the steady-state operation with high temperature vacuum vessel wall, the PCS is also required to have the capability to control the plasma for a long period. For the increase of the loads to processing power of the PCS, the PCS is decentralized with the use of reflective memories (RFMs). The PCS controls the plasma edge position with the real-time identification of plasma current and its position. This identification is done with not only flux loops but also hall sensors. The gas fueling method by piezo valve with monitoring the Hα signal filtered by a digital low-pass filter are proposed and suitable for the steady-state operation on QUEST. The present status and prospect of the PCS are presented with recent topics.

  18. Current status and prospect of plasma control system for steady-state operation on QUEST

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Makoto, E-mail: hasegawa@triam.kyushu-u.ac.jp; Nakamura, Kazuo; Zushi, Hideki; Hanada, Kazuaki; Fujisawa, Akihide; Tokunaga, Kazutoshi; Idei, Hiroshi; Nagashima, Yoshihiko; Kawasaki, Shoji; Nakashima, Hisatoshi; Higashijima, Aki

    2016-11-15

    Highlights: • Overall configuration of plasma control system on QUEST are presented. • Multi core system and reflective memories are used for the real-time control. • Hall sensors are used for the identification of plasma current and its position. • Repetitive gas fueling with the feed-back control of Hα signal is implemented. - Abstract: The plasma control system (PCS) of QUEST is developed according to the progress of QUEST project. Since one of the critical goals of the project is to achieve the steady-state operation with high temperature vacuum vessel wall, the PCS is also required to have the capability to control the plasma for a long period. For the increase of the loads to processing power of the PCS, the PCS is decentralized with the use of reflective memories (RFMs). The PCS controls the plasma edge position with the real-time identification of plasma current and its position. This identification is done with not only flux loops but also hall sensors. The gas fueling method by piezo valve with monitoring the Hα signal filtered by a digital low-pass filter are proposed and suitable for the steady-state operation on QUEST. The present status and prospect of the PCS are presented with recent topics.

  19. Intelligent plasma operation and control system for HL-2M

    International Nuclear Information System (INIS)

    Xia, F.; Chen, L.Y.; Wang, C.; Zhang, G.; Song, X.M.; Song, X.; Pan, L.; Zhao, L.; Pan, W.; Lan, J.T.

    2015-01-01

    Full text of publication follows. The Intelligent Plasma Operation and Control System (IPOCS) for HL-2M is under construction based on the current actual situation of HL-2A control system in SWIP. The purpose of IPOCS is to replace the routine processes before, during and post discharge with the automatic process algorithms and give various information, for example, status, warning, reason, instruction, suggestion, physics phenomenon etc. to staff who can access the intranet. In this case, the core research objectives can be focused on and the related operations can be carried out referring to the information. The ultimate goal of IPOCS is to improve the efficiency of plant operation and control and physics research for HL-2M and for the fusion reactor in the future. There are three layers in IPOCS, which are named the information collection layer, the data processing layer and the presentation layer respectively. Information collection layer consists of data acquisition system, EPICS system, Audio and Video system, discharge scheduling system, data storage system etc. All raw data are collected and stored in this layer. The data processing layer is the core of IPOCS. Most algorithms are executed here. The basic computation platform is based on cluster (56 cores at present) and Parallel Computing Toolbox provided by Matlab. The physics database for SWIP based on HDF5 and offline EFIT also locate at this layer. All the operations can be finished during the time interval between two shots. The last layer is to present the information from the former two layers. The typical hardware including the large display screen system in the control room, the voice broadcasting system, personal monitors and smart phones, etc.. Several applications has being developed such as Control System Studio (CSS) in SWIP version, WebOPI, Automatic Alarm Display system and so on. IPOCS for HL-2M is at the very beginning phase at present. With more and more systems and algorithms got

  20. Response of a core coherent density oscillation on electron cyclotron resonance heating in Heliotron J plasma

    Science.gov (United States)

    Kobayashi, T.; Kobayashi, S.; Lu, X. X.; Kenmochi, N.; Ida, K.; Ohshima, S.; Yamamoto, S.; Kado, S.; Kokubu, D.; Nagasaki, K.; Okada, H.; Minami, T.; Otani, Y.; Mizuuchi, T.

    2018-01-01

    We report properties of a coherent density oscillation observed in the core region and its response to electron cyclotron resonance heating (ECH) in Heliotron J plasma. The measurement was performed using a multi-channel beam emission spectroscopy system. The density oscillation is observed in a radial region between the core and the half radius. The poloidal mode number is found to be 1 (or 2). By modulating the ECH power with 100 Hz, repetition of formation and deformation of a strong electron temperature gradient, which is likely ascribed to be an electron internal transport barrier, is realized. Amplitude and rotation frequency of the coherent density oscillation sitting at the strong electron temperature gradient location are modulated by the ECH, while the poloidal mode structure remains almost unchanged. The change in the rotation velocity in the laboratory frame is derived. Assuming that the change of the rotation velocity is given by the background E × B velocity, a possible time evolution of the radial electric field was deduced.

  1. Dynamic Stall Control Using Plasma Actuators

    Science.gov (United States)

    Webb, Nathan; Singhal, Achal; Castaneda, David; Samimy, Mo

    2017-11-01

    Dynamic stall occurs in many applications, including sharp maneuvers of fixed wing aircraft, wind turbines, and rotorcraft and produces large unsteady aerodynamic loads that can lead to flutter and mechanical failure. This work uses flow control to reduce the unsteady loads by excitation of instabilities in the shear layer over the separated region using nanosecond pulse driven dielectric barrier discharge (NS-DBD) plasma actuators. These actuators have been shown to effectively delay or mitigate static stall. A wide range of flow parameters were explored in the current work: Reynolds number (Re = 167,000 to 500,000), reduced frequency (k = 0.025 to 0.075), and excitation Strouhal number (Ste = 0 to 10). Based on the results, three major conclusions were drawn: (a) Low Strouhal number excitation (Ste <0.5) results in oscillatory aerodynamic loads in the stalled stage of dynamic stall; (b) All excitation resulted in earlier flow reattachment; and (c) Excitation at progressively higher Ste weakened and eventually eliminated the dynamic stall vortex (DSV), thereby dramatically reducing the unsteady loading. The decrease in the strength of the DSV is achieved by the formation of shear layer coherent structures that bleed the leading-edge vorticity prior to the ejection of the DSV.

  2. Feedback control of plasma configuration in JT-60

    International Nuclear Information System (INIS)

    Ninomiya, Hiromasa; Kikuchi, Mitsuru; Yoshino, Ryuji; Hosogane, Nobuyuki; Kimura, Toyoaki; Kurihara, Kenichi; Takahashi, Minoru; Hayashi, Kazuo.

    1986-08-01

    Plasma current, plasma position (center of the outermost magnetic surface), decay index n index and width of the divertor throat are feedback controlled by using 5 kinds of poloidal field coils in JT-60. 5 control commands are calculated in a feedback control computer in each 1 msec. These feedback control functions are checked in ohmically heated plasma. The control characteristics of the plasma are well understood by the simplified control analysis and are consistent with the precise matrix transfer function analysis in the frequency domain and the simulation analysis which include the effects of eddy currents, delay time elements and mutual interactions between controllers. The usefulness of these analyses is experimentally confirmed. Each controlled variable is well feedback controlled to the command and the experimentally realized equilibrium configuration is checked by the well calibrated magnetic probes. Fast boundary identification code is used for the identification of the equilibrium and results are consistent with the precalculated plasma equilibria. By using this feedback control system of the plasma configuration and the equilibrium identification method, we have obtained the stable limiter and divertor configuration. The maximum parameters obtained during OH(I) experimental period are plasma current I p = 1.8 MA, the effective safety factor q eff e = 5.7 x 10 19 m -3 (Murakami parameter of 4.5) and the pulse length of 5 ∼ 10 sec. (author)

  3. Variable solar control using thermotropic core/shell particles

    Energy Technology Data Exchange (ETDEWEB)

    Muehling, Olaf; Seeboth, Arno; Ruhmann, Ralf; Potechius, Elvira; Vetter, Renate [Fraunhofer Institute for Applied Polymer Research (IAP), Department of Chromogenic Polymers, Volmerstr. 7B, 12489 Berlin (Germany); Haeusler, Tobias [Brandenburg University of Technology (BTU Cottbus), Chair of Applied Physics/Thermophysics, Konrad-Zuse-Str. 1, 03046 Cottbus (Germany)

    2009-09-15

    Subject of our recent investigations is the utilization of a reversible thermotropic material for a self-regulating sun protection glazing that controls the solar energy input in order to avoid overheating. Based on the well-established UV curing technology for laminated glass a superior thermotropic material with tunable switching characteristics and of low material costs was developed. The polymer layer contains core/shell particles homogeneously dispersed in a UV-cured resin. The particle core in turn consists of an n-alkane mixture that is responsible for the temperature-induced clear/opaque switching. To obtain particles of well-defined size and with a narrow size distribution, the miniemulsion polymerization technique was used. The visible and solar optical properties (normal-normal, normal-hemispherical, and normal-diffuse transmittance) in the off (clear) and in the on state (opaque) were determined by UV/Vis/NIR spectroscopy. Samples containing particles of high median diameter (>800 nm) primarily scatter in the forward direction. However, with smaller particles (300-600 nm) a higher backscattering (reflection) efficiency was achieved. The largest difference in the normal-hemispherical transmittance could be found with a particle amount of 6% and a median scattering domain diameter of {proportional_to}380 nm. (author)

  4. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  5. Chaos control and taming of turbulence in plasma devices

    DEFF Research Database (Denmark)

    Klinger, T.; Schröder, C.; Block, D.

    2001-01-01

    Chaos and turbulence are often considered as troublesome features of plasma devices. In the general framework of nonlinear dynamical systems, a number of strategies have been developed to achieve active control over complex temporal or spatio-temporal behavior. Many of these techniques apply...... to plasma instabilities. In the present paper we discuss recent progress in chaos control and taming of turbulence in three different plasma "model" experiments: (1) Chaotic oscillations in simple plasma diodes, (2) ionization wave turbulence in the positive column of glow discharges, and (3) drift wave...

  6. Current control necessary for toroidal plasma equilibrium

    International Nuclear Information System (INIS)

    Nagao, S.

    1987-01-01

    It is shown that a significant amount of dipole current is necessary for the plasma equilibrium of toroidal configurations in general. Through the vector product with the poloidal field, this dipole current force has to balance with the hoop force of plasma pressure itself of the annular shape. The measurement of such a current of dipole type may be interesting for the confirmation of the plasma equilibrium in the toroidal system. Moreover it is certained that there is a new mode of a tokamak operation with such a dipole current component and with smaller vertical field than that based on the classical tokamak theory. (author) [pt

  7. Digital control of plasma position in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

    2002-03-01

    Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

  8. Atto-second control of collective electron motion in plasmas

    International Nuclear Information System (INIS)

    Borot, Antonin; Malvache, Arnaud; Chen, Xiaowei; Jullien, Aurelie; Lopez-Martens, Rodrigo; Geindre, Jean-Paul; Audebert, Patrick; Mourou, Gerard; Quere, Fabien

    2012-01-01

    Today, light fields of controlled and measured waveform can be used to guide electron motion in atoms and molecules with atto-second precision. Here, we demonstrate atto-second control of collective electron motion in plasmas driven by extreme intensity (approximate to 10 18 W cm -2 ) light fields. Controlled few-cycle near-infrared waves are tightly focused at the interface between vacuum and a solid-density plasma, where they launch and guide sub-cycle motion of electrons from the plasma with characteristic energies in the multi-kilo-electron-volt range-two orders of magnitude more than has been achieved so far in atoms and molecules. The basic spectroscopy of the coherent extreme ultraviolet radiation emerging from the light-plasma interaction allows us to probe this collective motion of charge with sub-200 as resolution. This is an important step towards atto-second control of charge dynamics in laser-driven plasma experiments. (authors)

  9. plasmatis Center for Innovation Competence: Controlling reactive component output of atmospheric pressure plasmas in plasma medicine

    Science.gov (United States)

    Reuter, Stephan

    2012-10-01

    The novel approach of using plasmas in order to alter the local chemistry of cells and cell environment presents a significant development in biomedical applications. The plasmatis center for innovation competence at the INP Greifswald e.V. performs fundamental research in plasma medicine in two interdisciplinary research groups. The aim of our plasma physics research group ``Extracellular Effects'' is (a) quantitative space and time resolved diagnostics and modelling of plasmas and liquids to determine distribution and composition of reactive species (b) to control the plasma and apply differing plasma source concepts in order to produce a tailored output of reactive components and design the chemical composition of the liquids/cellular environment and (c) to identify and understand the interaction mechanisms of plasmas with liquids and biological systems. Methods to characterize the plasma generated reactive species from plasma-, gas- and liquid phase and their biological effects will be presented. The diagnostic spectrum ranges from absorption/emission/laser spectroscopy and molecular beam mass spectrometry to electron paramagnetic resonance spectroscopy and cell biological diagnostic techniques. Concluding, a presentation will be given of the comprehensive approach to plasma medicine in Greifswald where the applied and clinical research of the Campus PlasmaMed association is combined with the fundamental research at plasmatis center.

  10. Fundamentals of plasma physics and controlled fusion

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Kenro

    2000-10-01

    The present lecture note was written to fill a gap between text books for undergraduates and specific review articles written by specialists for their young colleagues. The note may be divided in three parts. The first part is on basic characteristics of a plasma in a magnetic field. The second part describes plasma confinement and heating with an emphasis on magnetohydrodynamic instabilities. In addition, propagation of plasma waves, plasma heating by electromagnetic waves are given. The third part is devoted to various specific concepts of nuclear fusion. Emphases are placed on toroidal devices, especially on tokamak devices and stellarators. One might feel heavy mathematics glimpsing the present note, especially in the part treating magnetohydrodynamic instabilities. (author)

  11. Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

    Science.gov (United States)

    Fukuda, Takeshi

    The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

  12. Control of plasma density distribution via wireless power transfer in an inductively coupled plasma

    International Nuclear Information System (INIS)

    Lee, Hee-Jin; Lee, Hyo-Chang; Kim, Young-Cheol; Chung, Chin-Wook

    2013-01-01

    With an enlargement of the wafer size, development of large-area plasma sources and control of plasma density distribution are required. To control the spatial distribution of the plasma density, wireless power transfer is applied to an inductively coupled plasma for the first time. An inner powered antenna and an outer resonant coil connected to a variable capacitor are placed on the top of the chamber. As the self-resonance frequency ω r of the resonant coil is adjusted, the power transfer rate from the inner powered coil to the outer resonant coil is changed and the dramatic evolution of the plasma density profile is measured. As ω r of the outer resonant coil changes from the non-resonant condition (where ω r is not the driving angular frequency ω rf ) to the resonant condition (where ω r = ω rf ), the plasma density profile evolves from a convex shape with maximal plasma density at the radial center into a concave shape with maximal plasma density in the vicinity of the resonant antenna coil. This result shows that the plasma density distribution can be successfully controlled via wireless resonance power transfer. (fast track communication)

  13. Real-time digital control of plasma position and shape on the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Mitri, Mikhael

    2009-01-01

    Beside the objective of contributing to the controlled thermonuclear fusion research and ultimately the development of a fusion based power plant, the main objectives of the thesis are a substantial improvement of plasma vertical position control and plasma shape control as well as a better understanding of formerly unexplained effects, e.g. disturbance fields. As for the vertical position control, a deep analysis has to be undertaken to identify the problem sources. Accurate control of the plasma position is very difficult to achieve. This is mainly due to the complexity of the tokamak and the difficulty in measuring or modelling all relevant discharge variables. Any models would be highly nonlinear and time varying. Thus, for simulation and controller design, a simplified, but nevertheless accurate model has to be developed, based on physics and measured data of the process. Furthermore, the quality of the measured position has to be improved by using new inductive sensors, integrators, and hardware. The integration drift problem has to be analysed and resolved by developing a drift-free integration method. Concerning the shape control, a better understanding of the relation between the stray fields and the iron core saturation is required. Furthermore, the influence on the plasma elongation has to be determined. Upon this, a shape compensation algorithm has to be developed accordingly. The accuracy of the shape control has to be better than 1%. (orig.)

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

    Science.gov (United States)

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

    2018-02-01

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

  15. Alushta-2012. International Conference-School on Plasma Physics and Controlled Fusion and the Adjoint Workshop 'Nano-and micro-sized structures in plasmas'. Book of Abstracts

    International Nuclear Information System (INIS)

    Makhlaj, V.A.

    2012-01-01

    The Conference was devoted to a new valuable information about the present status of plasma physics and controlled fusion research. The main topics was : magnetic confinement systems; plasma heating and current drive; ITER and fusion reactor aspects; basic plasma physics; space plasma; plasma dynamics and plasma-wall interaction; plasma electronics; low temperature plasma and plasma technologies; plasma diagnostics; formation of nano-and micro-sized structures in plasmas; properties of plasmas with nano- and micro- objects

  16. Plasma physics and controlled nuclear fusion research 1988. V.3

    International Nuclear Information System (INIS)

    1989-01-01

    Volume 3 of the proceedings of the twelfth international conference on plasma physics and controlled nuclear fusion, held in Nice, France, 12-19 October, 1988, contains papers presented on inertial fusion. Direct and indirect laser implosion experiments, programs of laser construction, computer modelling of implosions and resulting plasmas, and light ion beam fusion experiments are discussed. Refs, figs and tabs

  17. Optimal control theory applied to fusion plasma thermal stabilization

    International Nuclear Information System (INIS)

    Sager, G.; Miley, G.; Maya, I.

    1985-01-01

    Many authors have investigated stability characteristics and performance of various burn control schemes. The work presented here represents the first application of optimal control theory to the problem of fusion plasma thermal stabilization. The objectives of this initial investigation were to develop analysis methods, demonstrate tractability, and present some preliminary results of optimal control theory in burn control research

  18. Implementation of GPU parallel equilibrium reconstruction for plasma control in EAST

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Yao, E-mail: yaohuang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); School of Nuclear Science & Technology, University of Science & Technology of China (China); Luo, Z.P.; Yuan, Q.P.; Pei, X.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yue, X.N. [School of Nuclear Science & Technology, University of Science & Technology of China (China)

    2016-11-15

    Highlights: • We described parallel equilibrium reconstruction code P-EFIT running on GPU was integrated with EAST plasma control system. • Compared with RT-EFIT used in EAST, P-EFIT has better spatial resolution and full algorithm of EFIT per iteration. • With the data interface through RFM, 65 × 65 spatial grids P-EFIT can satisfy the accuracy and time feasibility requirements for plasma control. • Successful control using ISOFLUX/P-EFIT was established in the dedicated experiment during the EAST 2014 campaign. • This work is a stepping-stone towards versatile ISOFLUX/P-EFIT control, such as real-time equilibrium reconstruction with more diagnostics. - Abstract: Implementation of P-EFIT code for plasma control in EAST is described. P-EFIT is based on the EFIT framework, but built with the CUDA™ architecture to take advantage of massively parallel Graphical Processing Unit (GPU) cores to significantly accelerate the computation. 65 × 65 grid size P-EFIT can complete one reconstruction iteration in 300 μs, with one iteration strategy, it can satisfy the needs of real-time plasma shape control. Data interface between P-EFIT and PCS is realized and developed by transferring data through RFM. First application of P-EFIT to discharge control in EAST is described.

  19. Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    Goutych, I F; Gresillon, D; Sitenko, A G

    1997-12-31

    This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas.

  20. Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    Goutych, I.F.; Gresillon, D.; Sitenko, A.G.

    1996-01-01

    This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas

  1. A structured architecture for advanced plasma control experiments

    International Nuclear Information System (INIS)

    Penaflor, B.G.; Ferron, J.R.; Walker, M.L.

    1996-10-01

    Recent new and improved plasma control regimes have evolved from enhancements to the systems responsible for managing the plasma configuration on the DIII-D tokamak. The collection of hardware and software components designed for this purpose is known at DIII-D as the Plasma Control System or PCS. Several new user requirements have contributed to the rapid growth of the PCS. Experiments involving digital control of the plasma vertical position have resulted in the addition of new high performance processors to operate in real-time. Recent studies in plasma disruptions involving the use of neural network based software have resulted in an increase in the number of input diagnostic signals sampled. Better methods for estimating the plasma shape and position have brought about numerous software changes and the addition of several new code modules. Furthermore, requests for performing multivariable control and feedback on the current profile are continuing to add to the demands being placed on the PCS. To support all of these demands has required a structured yet flexible hardware and software architecture for maintaining existing capabilities and easily adding new ones. This architecture along with a general overview of the DIII-D Plasma Control System is described. In addition, the latest improvements to the PCS are presented

  2. BOOK REVIEW: Controlled Fusion and Plasma Physics

    Science.gov (United States)

    Engelmann, F.

    2007-07-01

    This new book by Kenro Miyamoto provides an up-to-date overview of the status of fusion research and the important parts of the underlying plasma physics at a moment where, due to the start of ITER construction, an important step in fusion research has been made and many new research workers will enter the field. For them, and also for interested graduate students and physicists in other fields, the book provides a good introduction into fusion physics as, on the whole, the presentation of the material is quite appropriate for getting acquainted with the field on the basis of just general knowledge in physics. There is overlap with Miyamoto's earlier book Plasma Physics for Nuclear Fusion (MIT Press, Cambridge, USA, 1989) but only in a few sections on subjects which have not evolved since. The presentation is subdivided into two parts of about equal length. The first part, following a concise survey of the physics basis of thermonuclear fusion and of plasmas in general, covers the various magnetic configurations studied for plasma confinement (tokamak; reversed field pinch; stellarator; mirror-type geometries) and introduces the specific properties of plasmas in these devices. Plasma confinement in tokamaks is treated in particular detail, in compliance with the importance of this field in fusion research. This includes a review of the ITER concept and of the rationale for the choice of ITER's parameters. In the second part, selected topics in fusion plasma physics (macroscopic instabilities; propagation of waves; kinetic effects such as energy transfer between waves and particles including microscopic instabilities as well as plasma heating and current drive; transport phenomena induced by turbulence) are presented systematically. While the emphasis is on displaying the essential physics, deeper theoretical analysis is also provided here. Every chapter is complemented by a few related problems, but only partial hints for their solution are given. A selection of

  3. Control of plasma column horizontal position in TBR-1

    International Nuclear Information System (INIS)

    Tuszel, A.G.; Rincoski, C.R.M.

    1990-01-01

    The TBR-1 is a small tokamak built at the Physics Institute of the University of Sao Paulo. It was originally designed with a simple vertical field power supply made of one fast capacitor bank for vertical current build-up and one slow capacitor bank for flat-top phase, without any control but the adjustable initial voltages of the capacitors. With such an elementary system, the plasma cannot be held in the center of the vacuum vessel for the whole duration of the plasma. This led to a suboptimal performance with easy disruptions. A control system was designed to hold the plasma centered in the radial coordinate. (Author)

  4. Electron internal transport barrier formation and dynamics in the plasma core of the TJ-II stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Estrada, T [Laboratorio Nacional de Fusion por Confinamiento Magnetico, Asociacion Euratom-CIEMAT, 28040 Madrid (Spain); Krupnik, L [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine); Dreval, N [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine); Melnikov, A [Institute of Nuclear Fusion, RRC ' Kurchatov Institute' , Moscow, Russia (Russian Federation); Khrebtov, S M [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine); Hidalgo, C [Laboratorio Nacional de Fusion por Confinamiento Magnetico, Asociacion Euratom-CIEMAT, 28040 Madrid (Spain); Milligen, B van [Laboratorio Nacional de Fusion por Confinamiento Magnetico, Asociacion Euratom-CIEMAT, 28040 Madrid (Spain); Castejon, F [Laboratorio Nacional de Fusion por Confinamiento Magnetico, Asociacion Euratom-CIEMAT, 28040 Madrid (Spain); AscasIbar, E [Laboratorio Nacional de Fusion por Confinamiento Magnetico, Asociacion Euratom-CIEMAT, 28040 Madrid (Spain); Eliseev, L [Institute of Nuclear Fusion, RRC ' Kurchatov Institute' , Moscow, Russia (Russian Federation); Chmyga, A A [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine); Komarov, A D [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine); Kozachok, A S [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine); Tereshin, V [Institute of Plasma Physics, NSC ' KIPT' , Kharkov (Ukraine)

    2004-01-01

    The influence of magnetic topology on the formation of electron internal transport barriers (e-ITBs) has been studied experimentally in electron cyclotron heated plasmas in the stellarator TJ-II. e-ITB formation is characterized by an increase in core electron temperature and plasma potential. The positive radial electric field increases by a factor of 3 in the central plasma region when an e-ITB forms. The experiments reported demonstrate that the formation of an e-ITB depends on the magnetic configuration. Calculations of the modification of the rotational transform due to plasma current lead to the interpretation that the formation of an e-ITB can be triggered by positioning a low order rational surface close to the plasma core region. In configurations without any central low order rational, no barrier is formed for any accessible value of heating power. Different mechanisms associated with neoclassical/turbulent bifurcations and kinetic effects are put forward to explain the impact of magnetic topology on radial electric fields and confinement.

  5. Plasma under control: Advanced solutions and perspectives for plasma flux management in material treatment and nanosynthesis

    Science.gov (United States)

    Baranov, O.; Bazaka, K.; Kersten, H.; Keidar, M.; Cvelbar, U.; Xu, S.; Levchenko, I.

    2017-12-01

    Given the vast number of strategies used to control the behavior of laboratory and industrially relevant plasmas for material processing and other state-of-the-art applications, a potential user may find themselves overwhelmed with the diversity of physical configurations used to generate and control plasmas. Apparently, a need for clearly defined, physics-based classification of the presently available spectrum of plasma technologies is pressing, and the critically summary of the individual advantages, unique benefits, and challenges against key application criteria is a vital prerequisite for the further progress. To facilitate selection of the technological solutions that provide the best match to the needs of the end user, this work systematically explores plasma setups, focusing on the most significant family of the processes—control of plasma fluxes—which determine the distribution and delivery of mass and energy to the surfaces of materials being processed and synthesized. A novel classification based on the incorporation of substrates into plasma-generating circuitry is also proposed and illustrated by its application to a wide variety of plasma reactors, where the effect of substrate incorporation on the plasma fluxes is emphasized. With the key process and material parameters, such as growth and modification rates, phase transitions, crystallinity, density of lattice defects, and others being linked to plasma and energy fluxes, this review offers direction to physicists, engineers, and materials scientists engaged in the design and development of instrumentation for plasma processing and diagnostics, where the selection of the correct tools is critical for the advancement of emerging and high-performance applications.

  6. Development of the 'JFT-2' tokamak plasma position control system

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

    1980-01-01

    Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

  7. 10th International Conference and School on Plasma Physics and Controlled Fusion. Book of Abstracts

    International Nuclear Information System (INIS)

    Anon

    2004-01-01

    About 240 abstracts by Ukrainian and foreign authors submitted to 10-th International Conference and School on Plasma Physics and Controlled fusion have been considered by Conference Program Committee members. All the abstracts have been divided into 8 groups: magnetic confinement systems: stellarators, tokamaks, alternative conceptions; ITER and Fusion reactor aspects; basic plasma physics; space plasma; plasma dynamics and plasma-wall interaction; plasma electronics; low temperature plasma and plasma technologies; plasma diagnostics

  8. Repetitive 'snakes' and their damping effect on core toroidal rotation in EAST plasmas with multiple H-L-H transitions

    International Nuclear Information System (INIS)

    Xu Liqing; Hu Liqun

    2015-01-01

    Repetitive impurity snake-modes have been observed after H-L mode transitions (high to low confinement modes) in EAST plasmas exhibiting multiple H-L-H transitions. Such snake-modes have been observed to lower the core plasma toroidal rotation. A critical impurity strength factor associated with snake-mode formation has been estimated to be as high as α_Z_,_c =n_Z_,_cZ"2 / n_e ∼0.75. These observations have implications for ITER H-mode sustainability when the heating power is only slightly above the H-mode power threshold. (author)

  9. Estimation and Control for Autonomous Coring from a Rover Manipulator

    Science.gov (United States)

    Hudson, Nicolas; Backes, Paul; DiCicco, Matt; Bajracharya, Max

    2010-01-01

    A system consisting of a set of estimators and autonomous behaviors has been developed which allows robust coring from a low-mass rover platform, while accommodating for moderate rover slip. A redundant set of sensors, including a force-torque sensor, visual odometry, and accelerometers are used to monitor discrete critical and operational modes, as well as to estimate continuous drill parameters during the coring process. A set of critical failure modes pertinent to shallow coring from a mobile platform is defined, and autonomous behaviors associated with each critical mode are used to maintain nominal coring conditions. Autonomous shallow coring is demonstrated from a low-mass rover using a rotary-percussive coring tool mounted on a 5 degree-of-freedom (DOF) arm. A new architecture of using an arm-stabilized, rotary percussive tool with the robotic arm used to provide the drill z-axis linear feed is validated. Particular attention to hole start using this architecture is addressed. An end-to-end coring sequence is demonstrated, where the rover autonomously detects and then recovers from a series of slip events that exceeded 9 cm total displacement.

  10. Method of controlling plasma discharge in a thermonuclear device

    International Nuclear Information System (INIS)

    Kawasaki, Kozo; Ishida, Takayuki; Takemaru, Koichi; Kawasaki, Takahide.

    1982-01-01

    Purpose: To prolong the plasma discharging period by previously increasing the temperature at the thick portion of a vacuum container prior to the plasma discharge to thereby decrease the temperature difference caused by the plasma discharge between the thick portion and the bellows. Method: Temperature values at the outer surface of the thick portion and the bellows of a vacuum container detected by temperature sensors are applied to the input processing section of a temperature control device, and baking control is carried out by way of the output processing section so that each of the portions of the vacuum container may be maintained at the temperature set by the temperature setting section based on the calculation performed in the control processing section. By previously increasing the temperature β at the thick portion higher by about 100 0 C than the temperature α for the bellows in the baking treatment prior to the plasma discharge, the plasma discharge period during which the temperature levels at both of the portions are reversed after the plasma discharge and the temperature difference arrives at a predetermined level i.g., of 100 0 C can significantly be prolonged as compared with the case where the plasma discharge is started at the same temperature for both of the portions. (Yoshino, Y.)

  11. Control of horizontal plasma position by feedforward-feedback system with digital computer in the JIPP T-II tokamak

    International Nuclear Information System (INIS)

    Toi, Kazuo; Sakurai, Keiichi; Itoh, Satoshi; Matsuura, Kiyokata; Tanashi, Shugo

    1980-01-01

    In the resistive shell tokamak, JIPP T-II, the control of horizontal plasma position is successfully carried out by calculating the equilibrium equation of a large-aspect-ratio tokamak plasma surrounded by a thin resistive shell of a skin time of 5.2 ms, every 1.39 ms with a digital computer. The iron core effect is also taken into account by a simple form in the equation. The required strenght of vertical field is determined by the control demand composed of two groups; one is a ''feedback'' term expressed by the deviation of plasma position from the desired one and proportion-integration-differentiation correction (PID-controller), and the other is a ''feedforward'' term which is in proportion to the plasma current. The experimental results in a quasi-constant phase of plasma current are in good agreement with the stability analysis of the control system by using the so-called Bode-diagram which is calculated on the assumption that the plasma current is independent of time. By this control system, the horizontal plasma displacement has been suppressed within 1 cm of the initiation of discharge to the termination in the high-density and low-q(a) plasma of 15 cm radius which is obtained by both strong gas puffing and second current rise. (author)

  12. Core failure accident pathways and ways to control it

    International Nuclear Information System (INIS)

    Mayinger, F.

    1982-01-01

    In the German Risk Study accidents are assumed to result in core meltdown whenever the criteria spelt out in the guidelines of the Advisory Committee on Reactor Safeguards are no longer met. This assumption must be seen in the light of an earlier state of the art in which no detailed information could be obtained about intermediate stages in emergency core cooling systems working according to permit up to the complete failure of all heat removal systems. However, experimental studies and theoretical analyses conducted over the past few years have advanced the state of the art such that it is now possible to predict with considerably more physical reality the behavior of a core in a loss-of-coolant accident. These findings are not only based on calculations, but also on the results of experiments in large facilities allowing direct comparisons to be made with conditions in nuclear power plants. Studies of the effects of systems failures both in major leakages and in the small leakages regarded to be much more dangerous show much more favorable conditions with respect to core coolability than had to be anticipated on the basis of earlier assumptions. This also implies that it would neither be necessary nor meaningful to reinforce emergency core cooling systems. Instead, it is much more important, besides having technically highly qualified and thoroughly trained operating crews, to inform those crews reliably of the hydrodynamic and thermodynamic state of the primary system, especially the core. (orig.) [de

  13. Reactive gas control of non-stable plasma conditions

    International Nuclear Information System (INIS)

    Bellido-Gonzalez, V.; Daniel, B.; Counsell, J.; Monaghan, D.

    2006-01-01

    Most industrial plasma processes are dependant upon the control of plasma properties for repeatable and reliable production. The speed of production and range of properties achieved depend on the degree of control. Process control involves all the aspects of the vacuum equipment, substrate preparation, plasma source condition, power supplies, process drift, valves (inputs/outputs), signal and data processing and the user's understanding and ability. In many cases, some of the processes which involve the manufacturing of interesting coating structures, require a precise control of the process in a reactive environment [S.J. Nadel, P. Greene, 'High rate sputtering technology for throughput and quality', International Glass Review, Issue 3, 2001, p. 45. ]. Commonly in these circumstances the plasma is not stable if all the inputs and outputs of the system were to remain constant. The ideal situation is to move a process from set-point A to B in zero time and maintain the monitored signal with a fluctuation equal to zero. In a 'real' process that's not possible but improvements in the time response and energy delivery could be achieved with an appropriate algorithm structure. In this paper an advanced multichannel reactive plasma gas control system is presented. The new controller offers both high-speed gas control combined with a very flexible control structure. The controller uses plasma emission monitoring, target voltage or any process sensor monitoring as the input into a high-speed control algorithm for gas input. The control algorithm and parameters can be tuned to different process requirements in order to optimize response times

  14. Electron energy distribution function control in gas discharge plasmas

    International Nuclear Information System (INIS)

    Godyak, V. A.

    2013-01-01

    The formation of the electron energy distribution function (EEDF) and electron temperature in low temperature gas discharge plasmas is analyzed in frames of local and non-local electron kinetics. It is shown, that contrary to the local case, typical for plasma in uniform electric field, there is the possibility for EEDF modification, at the condition of non-local electron kinetics in strongly non-uniform electric fields. Such conditions “naturally” occur in some self-organized steady state dc and rf discharge plasmas, and they suggest the variety of artificial methods for EEDF modification. EEDF modification and electron temperature control in non-equilibrium conditions occurring naturally and those stimulated by different kinds of plasma disturbances are illustrated with numerous experiments. The necessary conditions for EEDF modification in gas discharge plasmas are formulated

  15. Coupled core criticality calculations with control rods located in the central reflector region

    Energy Technology Data Exchange (ETDEWEB)

    Sobhy, M [Reactor depatrment, nuclear research center, Inshaas (Egypt)

    1995-10-01

    The reactivity of a coupled core is controlled by a set of control rods distributed in the central reflector region. The reactor contains two compact cores cooled and moderated by light water. Control rods are designed to have reactivity worths sufficient to start, control and shutdown the coupled system. Each core in a coupled system is in subcritical conditions without any absorber then each core needs to the other core to fulfill nuclear chain reaction and to approach the criticality. In this case, each core is considered clean which is suitable for research reactor with low flux disturbance and better neutron economy, in addition to the advantage of disappearing the cut corner fuel baskets. This facilitate the in core fuel management with identical fuel baskets. Hot spots will disappear. This leads to a good heat transfer process. the excess reactivity and the shutdown margin are calculated for some of reflector as coupling region gives sufficient area for coupled core are calculated cost. The fluctuations of reactivity for coupled core are calculated by noise analysis technique and compared with that for rode core. The results show low reactivity perturbation associated with coupled core.

  16. Advances in Integrated Plasma Control on DIII-D

    International Nuclear Information System (INIS)

    Walker, M.L.; Ferron, J.R.; Humphreys, D.A.

    2006-01-01

    The DIII-D experimental program in advanced tokamak (AT) physics requires extremely high performance from the DIII-D plasma control system (PCS) [B.G.Penaflor, et al., 4 th IAEA Tech. Mtg on Control and Data Acq., San Diego, CA (2003)], including simultaneous and highly accurate regulation of plasma shape, stored energy, density, and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that 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 optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of several new model-based plasma controllers on DIII-D. We discuss experimental use of advanced shape control algorithms containing nonlinear techniques for improving control of steady state plasmas, model-based controllers for optimal rejection of edge localized mode disturbances during resistive wall mode stabilization, model-based controllers for neoclassical tearing mode stabilization, including methods for maximizing stabilization effectiveness with substantial constraints on available power, model-based integrated control of plasma rotation and beta, and initial experience in development of model-based controllers for advanced tokamak current profile modification. The experience gained from DIII-D has been applied to the development of control systems for the EAST and KSTAR tokamaks. We describe the development of the control software, hardware, and model-based control algorithms for these superconducting tokamaks, with emphasis on relevance of

  17. Plasma impurity-control studies in CTX

    International Nuclear Information System (INIS)

    Barnes, C.W.; Henins, I.; Hoida, H.W.; Jarboe, T.R.; Linford, R.K.; Marshall, J.; Sherwood, A.R.; Tuszewski, M.

    1981-01-01

    In the past, magnetized coaxial gun generated Compact Toroids (CTs) have exhibited magnetic field and density lifetimes of about 250 to 350 μs and electron temperatures of about 10 eV. In recent experiments, after hydrogen discharge cleaning the gun and flux conserver surfaces, the lifetimes have been extended to 550 μs. This improvement in lifetime, together with spectroscopic and bolometric measurements, are consistent with the interpretation that the CT plasma losses are impurity dominated and that discharge cleaning is reducing the impurities. Details of these measurements are described as well as successful experiments which led to a more open flux conserver

  18. Plasma impurity-control studies in CTX

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.W.; Henins, I.; Hoida, H.W.; Jarboe, T.R.; Linford, R.K.; Marshall, J.; Sherwood, A.R.; Tuszewski, M.

    1981-01-01

    In the past, magnetized coaxial gun generated Compact Toroids (CTs) have exhibited magnetic field and density lifetimes of about 250 to 350 ..mu..s and electron temperatures of about 10 eV. In recent experiments, after hydrogen discharge cleaning the gun and flux conserver surfaces, the lifetimes have been extended to 550 ..mu..s. This improvement in lifetime, together with spectroscopic and bolometric measurements, are consistent with the interpretation that the CT plasma losses are impurity dominated and that discharge cleaning is reducing the impurities. Details of these measurements are described as well as successful experiments which led to a more open flux conserver.

  19. Controlling Laser Plasma Instabilities Using Temporal Bandwidth

    Science.gov (United States)

    Tsung, Frank; Weaver, J.; Lehmberg, R.

    2016-10-01

    We are performing particle-in-cell simulations using the code OSIRIS to study the effects of laser plasma interactions in the presence of temporal bandwidth under conditions relevant to current and future experiments on the NIKE laser. Our simulations show that, for sufficiently large bandwidth (where the inverse bandwidth is comparable with the linear growth time), the saturation level, and the distribution of hot electrons, can be effected by the addition of temporal bandwidths (which can be accomplished in experiments using beam smoothing techniques such as ISI). We will quantify these effects and investigate higher dimensional effects such as laser speckles. This work is supported by DOE and NRL.

  20. Strategies for the plasma position and shape control in IGNITOR

    International Nuclear Information System (INIS)

    Ramogida, G.; Alladio, F.; Albanese, R.

    2006-01-01

    The control of the plasma position and shape is a crucial issue in IGNITOR as in every compact, high field, elongated tokamak. The capability of the Poloidal Field Coil system, as presently designed, to provide an effective vertical stabilization of the plasma has been investigated using the CREATE L response model [R. Albanese, F. Villone, '' The Linearized CREATE L Plasma Response Model for the Control of Current, Position and Shape in Tokamaks '', Nucl. Fus., vol. 38, p. 723 (1998)]. This linearized MHD model assumes an axisymmetric deformable plasma described by few global parameters. An optimization of the vertical position control strategy has been carried out and the most effective coil combination has been selected to stabilize the plasma while fulfilling engineering constraints on the coils and minimizing the required power and voltage. The two pairs of coils selected for the vertical control will be fed up with up-down anti-symmetric currents provided by a dedicated supply and overlapped to the scenario currents. The growth rate of the vertical instability and the power required by the active stabilization system have been estimated with this model, indicating that it is possible to design a control system able to guarantee a stability region that includes the most interesting operation conditions. An assessment of the requirements for the plasma cross section shape control has been carried out considering independent perturbations of the plasma global parameters as disturbances and showing that the undesired shape modification rejection is possible with the present PFC and power supply system. The PF coils have been ranked with respect to their capability to restore the shape modifications due to different plasma disturbances and the most effective coil combination, that minimizes recovery time and voltage required, has been selected. In order to have additional means to monitor and control the centre of the plasma column, under demanding conditions

  1. Strategies for the plasma position and shape control in IGNITOR

    International Nuclear Information System (INIS)

    Villone, F.; Albanese, R.; Ambrosino, G.; Pironti, A.; Rubinacci, G.; Ramogida, G.; Alladio, F.; Bombarda, F.; Coletti, A.; Cucchiaro, A.; Maddaluno, G.; Pizzicaroli, G.; Pizzuto, A.; Roccella, M.; Santinelli, M.; Coppi, B.

    2007-01-01

    The capability of the poloidal field coil system, as presently designed, to provide an effective vertical stabilization of the plasma in the IGNITOR machine has been investigated using the CREATE L response model. An optimization of the vertical position control strategy has been carried out and the most effective coil combination has been selected to stabilize the plasma while fulfilling engineering constraints on the coils and minimizing the required power and voltage. The growth rate of the vertical instability and the power required by the active stabilization system has been estimated with this model. The possible failure of the relevant electromagnetic diagnostics has been taken into account, evaluating the robustness of the plasma position reconstruction strategy. A realistic description of the power supply system has permitted to carry out the optimization of the proportional-integrative-derivative (PID) controller, both with a voltage and a current loop control scheme. An assessment of the requirements for the plasma cross section shape control has been carried out considering perturbations of the plasma global parameters independent of each other and showing that the undesired shape modification rejection is possible with the present PFC and power supply system. The PF coils have been rated relative to their capability to restore shape modifications due to different plasma disturbances. The most effective coil combination, that minimizes recovery time and voltage required, has been identified

  2. Automation of Aditya tokamak plasma position control DC power supply

    Energy Technology Data Exchange (ETDEWEB)

    Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.

    2016-11-15

    Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

  3. Integrated predictive modeling of high-mode tokamak plasmas using a combination of core and pedestal models

    International Nuclear Information System (INIS)

    Bateman, Glenn; Bandres, Miguel A.; Onjun, Thawatchai; Kritz, Arnold H.; Pankin, Alexei

    2003-01-01

    A new integrated modeling protocol is developed using a model for the temperature and density pedestal at the edge of high-mode (H-mode) plasmas [Onjun et al., Phys. Plasmas 9, 5018 (2002)] together with the Multi-Mode core transport model (MMM95) [Bateman et al., Phys. Plasmas 5, 1793 (1998)] in the BALDUR integrated modeling code to predict the temperature and density profiles of 33 H-mode discharges. The pedestal model is used to provide the boundary conditions in the simulations, once the heating power rises above the H-mode power threshold. Simulations are carried out for 20 discharges in the Joint European Torus and 13 discharges in the DIII-D tokamak. These discharges include systematic scans in normalized gyroradius, plasma pressure, collisionality, isotope mass, elongation, heating power, and plasma density. The average rms deviation between experimental data and the predicted profiles of temperature and density, normalized by central values, is found to be about 10%. It is found that the simulations tend to overpredict the temperature profiles in discharges with low heating power per plasma particle and to underpredict the temperature profiles in discharges with high heating power per particle. Variations of the pedestal model are used to test the sensitivity of the simulation results

  4. Progress and plan of KSTAR plasma control system upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Sang-hee, E-mail: hahn76@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Y.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Penaflor, B.G. [General Atomics, San Diego, CA (United States); Bak, J.G.; Han, H.; Hong, J.S.; Jeon, Y.M.; Jeong, J.H.; Joung, M.; Juhn, J.W.; Kim, J.S.; Kim, H.S.; Lee, W.R.; Woo, M.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Eidietis, N.W.; Ferron, J.R.; Humphreys, D.A.; Hyatt, A.; Johnson, R.D.; Piglowski, D.A. [General Atomics, San Diego, CA (United States); and others

    2016-11-15

    Highlights: • Recent achievements of the KSTAR plasma control system are described. • Requirements and results of the testbed system for the future upgrade of the KSTAR plasma control system are presented. • An overview of the upgrade layout based is given. - Abstract: The plasma control system (PCS) has been one of essential systems in annual KSTAR plasma campaigns: starting from a single-process version in 2008, extensive upgrades are done through the previous 7 years in order to achieve major goals of KSTAR performance enhancement. Major implementations are explained in this paper. In consequences of successive upgrades, the present KSTAR PCS is able to achieve ∼48 s of 500 kA plasma pulses with full real-time shaping controls and real-time NB power controls. It has become a huge system capable of dealing with 8 separate categories of algorithms, 26 actuators directly controllable during the shot, and real-time data communication units consisting of +180 analog channels and +600 digital input/outputs through the reflective memory (RFM) network. The next upgrade of the KSTAR PCS is planned in 2015 before the campaign. An overview of the upgrade layout will be given for this paper. The real-time system box is planned to use the CERN MRG-Realtime OS, an ITER-compatible standard operating system. New hardware is developed for faster real-time streaming system for future installations of actuators/diagnostics.

  5. The design of remote participation platform for EAST plasma control

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Q.P., E-mail: qpyuan@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science & Technology of China, Hefei (China); Zhang, R.R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Chai, W.T.; Liu, J.; Xiao, R.; Zhou, Z.C.; Pei, X.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science & Technology of China, Hefei (China)

    2016-11-15

    Highlights: • The remote participation platform for EAST plasma control is composed of real time control service and scenario management. • The web based interface has been developed for supporting remote participation. • The functionality module has been designed and assistant tools have been developed. - Abstract: EAST has become a physics experimental platform for high parameter and steady-state long-pulse plasma operation. A new remote participation platform for EAST plasma control is designed, which is composed of gatekeeper system, web-based user interface system, discharge scenario management system, online simulation system and data interface with on-site plasma control system (PCS). The identification and access privilege of remote participator is validated by the gatekeeper system. Only authorized users can set control parameters for next shot plasma control or making discharge scenario for future shot through WebPCS which is a web-based user interface and designed based on B/S structure. The systematic architecture design and preliminary deployment of such remote platform will be presented in this paper.

  6. JSPS-CAS Core University Program seminar on summary of 10-year collaborations in plasma and nuclear fusion research area

    International Nuclear Information System (INIS)

    Toi, Kazuo; Wang Kongjia

    2011-07-01

    The JSPS-CAS Core University Program (CUP) seminar on “Summary of 10-year Collaborations in Plasma and Nuclear Fusion Research Area” was held from March 9 to March 11, 2011 in the Okinawa Prefectural Art Museum, Naha city, Okinawa, Japan. The collaboration program on plasma and nuclear fusion started from 2001 under the auspices of Japanese Society of Promotion of Science (JSPS) and Chinese Academy of Sciences (CAS). This year is the last year of the CUP. This seminar was organized in the framework of the CUP. In the seminar, 29 oral talks were presented, having 14 Chinese and 30 Japanese participants. These presentations covered key topics related to the collaboration categories: (1) improvement of core plasma properties, (2) basic research on fusion reactor technologies, and (3) theory and numerical simulation. This seminar aims at summarizing the results obtained through the collaborations for 10 years, and discussing future prospects of China-Japan collaboration in plasma and nuclear fusion research areas. (author)

  7. Soft X-ray Tangential Imaging of the NSTX Core Plasma by Means of a MPGD Pin-hole Camera

    International Nuclear Information System (INIS)

    Pacella, D.; Leigheb, M.; Bellazzini, R.; Brez, A.; Finkenthal, M.; Stutman, D.; Kaita, R.; Sabbagh, S.A.

    2003-01-01

    A fast X-ray system based on a Micro Pattern Gas Detector has been used, for the first time, to investigate emission from the plasma core of the National Spherical Tokamak eXperiment (NSTX) at the Princeton Plasma Physics Laboratory. The results presented in this work demonstrate the capability of such a device to measure with a time resolution of the order of 1 ms the curvature and the elongation of the X-ray iso-emissivity contours, under various plasma conditions. Also, comparisons with the magnetic surface structure calculated by the EFIT code show good agreement between reconstructed flux surface and the soft X-ray emissions (SXR) for poloidal beta values up to 0.6. For greater values of beta, X-ray iso-emissivity contours become circular, while magnetic flux surface reconstructions yield elongation 1.5 < k < 2.2. The X-ray images have been acquired with a (statistical) signal to noise ratio (SNR) per pixel of about 30. Thanks to the direct and efficient X-ray conversion and its operation in a photon counting mode, this new diagnostic tool allows the routine investigation of the plasma core with a sampling rate of 1 kHz and extremely high SNR under all experimental conditions in NSTX

  8. Physics and chemistry of plasma pollution control technology

    International Nuclear Information System (INIS)

    Chang, J S

    2008-01-01

    Gaseous pollution control technologies for acid gases (NO x , SO x , etc), volatile organic compounds, greenhouse gases, ozone layer depleting substances, etc have been commercialized based on catalysis, incineration and adsorption methods. However, non-thermal plasma techniques based on electron beams and corona discharges are becoming significant due to advantages such as lower costs, higher removal efficiency and smaller space volume. In order to commercialize this new technology, the pollution gas removal rate, energy efficiency of removal, pressure drop of reactors and useable by-product production rates must be improved and identification of major fundamental processes and optimizations of reactor and power supply for an integrated system must be investigated. In this work, the chemistry and physics of plasma pollution control are discussed and the limitation of this type of plasma is outlined based on the plasma parameters.

  9. Impact of gas puffing location on density control and plasma parameters in TJ-II

    International Nuclear Information System (INIS)

    Tabares, F.L.; Garcia-Cortes, I.; Estrada, T.; Tafalla, D.; Hidalgo, A.; Ferreira, J.A.; Pastor, I.; Herranz, J.; Ascasibar, E.

    2005-01-01

    Under pure Electron Cyclotron Resonance Heating (ECRH) conditions in TJ-II plasmas (P<300 kW, 53.2 GHz, 2nd harmonic X-mode, power density < 25 W/m''3), plasma start-up and good density control are obtained only by the proper combination of wall conditions and gas puffing characteristics. Such a control is particularly critical for the optimisation of the NBI power transfer to the target plasmas. The relatively low cut-off limit is easily reached due not only to the unfavourable wall/puffing-fuelling ratio but also due to the steep density profiles developed during the Enhanced Particle Confinement (EPC) modes. These modes are triggered by the gas puffing waveform, and they cannot be achieved for high iota magnetic configurations in TJ-II. Comparative experiments under metallic and boronised wall conditions have shown that the sensitivity of the EPC modes to the puffing rate is at least partially related to the energy balance at the plasma periphery under central heating scenarios. In this work, the impact of gas-fuelling location on the plasma parameters and density control is described. For that purpose, three different fuelling locations have been investigated; broad distribution from a side ports, localized injection from long tubes at different poloidal positions and highly localized injection through a movable limiter. Edge density and temperature profiles from a broad set of diagnostics (atomic beams, reflectometry, Thompson Scattering ECE, etc...) are analysed and compared. It has been found that preventing from transition to the EPC mode is achieved by using slow puffing rates, while neutral penetration into the plasma core can be enhanced for highly localized gas sources. Wall inventory, however, has been found to pl ay a dominant role in the fuelling of the plasma under most conditions. (author)

  10. Plasma physics and controlled nuclear fusion research 1990. V. 1

    International Nuclear Information System (INIS)

    1991-01-01

    Volume 1 of the Proceedings of the Thirteenth International Conference on Plasma Physics and Controlled Nuclear Fusion Research contains papers given in two of the sessions: A and E. Session A contains the Artsimovich Memorial Lecture and papers on tokamaks; session E papers on plasma heating and current drive. The titles and authors of each paper are listed in the Contents. Abstracts accompany each paper. Refs, figs and tabs

  11. Controlling the emission current from a plasma cathode

    International Nuclear Information System (INIS)

    Bagaev, S.P.; Gushenets, V.I.; Schanin, P.M.

    1993-01-01

    The processes determining the time and amplitude characteristics of the grid-controlled electron emission from the plasma of an arc discharge have been analyzed. It has been shown that by applying to the grid confining the plasma emission boundary of a modulated voltage it is possible to form current pulse of up to 1 kA with nanosecond risetimes and falltimes and a pulse repetitive rate of 100 kHz

  12. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.; Bell, R.E.

    2001-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  13. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.; Bell, R.E.

    1999-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  14. Implementation strategy for the ITER plasma control system

    International Nuclear Information System (INIS)

    Winter, A.; Ambrosino, G.; Bauvir, B.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neto, A.; Raupp, G.; Snipes, J.A.; Stephen, A.V.; Treutterer, W.; Walker, M.L.; Zabeo, L.

    2015-01-01

    This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

  15. Implementation strategy for the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Ambrosino, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Bauvir, B. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); De Tommasi, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Humphreys, D.A. [General Atomics, San Diego, CA (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Dip. Ingegneria Industriale e dell’Informazione (Italy); Neto, A. [Fusion for Energy, Barcelona (Spain); Raupp, G. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Snipes, J.A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Stephen, A.V. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon (United Kingdom); Treutterer, W. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Walker, M.L. [General Atomics, San Diego, CA (United States); Zabeo, L. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

    2015-10-15

    This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

  16. An experimental study of icing control using DBD plasma actuator

    Science.gov (United States)

    Cai, Jinsheng; Tian, Yongqiang; Meng, Xuanshi; Han, Xuzhao; Zhang, Duo; Hu, Haiyang

    2017-08-01

    Ice accretion on aircraft or wind turbine has been widely recognized as a big safety threat in the past decades. This study aims to develop a new approach for icing control using an AC-DBD plasma actuator. The experiments of icing control (i.e., anti-/de-icing) on a cylinder model were conducted in an icing wind tunnel with controlled wind speed (i.e., 15 m/s) and temperature (i.e., -10°C). A digital camera was used to record the dynamic processes of plasma anti-icing and de-icing whilst an infrared imaging system was utilized to map the surface temperature variations during the anti-/de-icing processes. It was found that the AC-DBD plasma actuator is very effective in both anti-icing and de-icing operations. While no ice formation was observed when the plasma actuator served as an anti-icing device, a complete removal of the ice layer with a thickness of 5 mm was achieved by activating the plasma actuator for ˜150 s. Such information demonstrated the feasibility of plasma anti-/de-icing, which could potentially provide more effective and safer icing mitigation strategies.

  17. Real time control of plasmas and ECRH systems on TCV

    NARCIS (Netherlands)

    Paley, J.I.; Felici, F.; Berrino, J.; Coda, S.; Cruz, N.; Duval, B.P.; Goodman, T.P.; Martin, Y.; Moret, J.-M.; Piras, F.; Rodrigues, A.P.; Santos, B.; Varandas, C.A.F.

    2008-01-01

    Developments in the real time control hardware on TCV paired with the flexibility of plasma shaping and ECRH actuators are opening many opportunities to perform real time experiments and develop algorithms and methods for fusion applications. The ability to control MHD instabilities is particularly

  18. Control of radial electric field in torus plasma

    International Nuclear Information System (INIS)

    Ida, K.; Idei, H.; Sanuki, H.

    1994-09-01

    The radial electric fields is controlled by changing the direction of neutral beam from co to counter to plasma current in tokamak, while it is controlled by the 2nd harmonic ECH and NBI and pellet injection in heliotron/torsatron. (author)

  19. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

    Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)

  20. Observation of inward and outward particle convection in the core of electron cyclotron heated and current driven plasmas in the Tokamak a Configuration Variable

    International Nuclear Information System (INIS)

    Furno, I.; Weisen, H.

    2003-01-01

    In the Tokamak a Configuration Variable [F. Hofmann, J.B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 (1994)], inward or outward convection in the core of electron cyclotron heated and current driven plasmas is observed, depending on discharge conditions. In sawtoothing discharges with central electron cyclotron heating, outward convection is observed when a quasicontinuous m=1 kink mode is present, resulting in inverted sawteeth on the central electron density, while in the absence thereof, inward convection between successive sawtooth crashes leads to 'normal' sawteeth. The occurrence of a kink mode depends sensitively on plasma triangularity. When sawteeth are stabilized with central co- or counterelectron cyclotron current drive, stationary hollow electron density profiles are observed in the presence of m=1 modes, while peaked or flat profiles are observed in magnetohydrodynamic quiescent discharges. The observation of peaked density profiles in fully electron cyclotron driven plasmas demonstrates that pinch processes other than the Ware pinch must be responsible for these phenomena

  1. Controlling interface characteristics by adjusting core-shell structure

    International Nuclear Information System (INIS)

    Chang, H.Y.; Cheng, S.Y.; Sheu, C.I.

    2004-01-01

    Most grain boundary layer ceramics comprise semiconductive/conductive grains and insulated grain boundaries. Such a structure can be theoretically regarded as a shell (grain boundary layer) surrounds a core (conductive or semiconductive grain). The core-shell structure of titanium (Ti)-strontium titanate (ST) is composed of three zones - ST, non-stoichiometric strontium-titanium oxide and Ti, in order from shell to core. It was successfully prepared using a hydrothermal method. The Ti-ST core-shell structure was sintered in a reducing atmosphere and then annealed in air to achieve the metal-insulator-metal structure (MIM structure). The resulting MIM structure, annealed in air, changes with the oxygen stoichiometry of the ST shell (insulator layer) at various temperatures, which is thus used to tune its electrical characteristics. The characteristics exhibit nonlinear behavior. Accordingly, the thickness of the insulator layer can be adjusted in various annealing atmospheres and at various temperatures to develop various interfacial devices, such as varistors, capacitors and thermistors, without the use of complex donor/acceptor doping technology

  2. Jet flow and premixed jet flame control by plasma swirler

    Energy Technology Data Exchange (ETDEWEB)

    Li, Gang, E-mail: ligang@iet.cn [Key laboratory of light duty gas turbine, Institute of Engineering Thermophysics, Chinese Academy of Sciences, Beijing 100190 (China); Jiang, Xi [School of Engineering and Materials Science, Queen Mary University of London, Mile End Road, London E1 4NS (United Kingdom); Zhao, Yujun [School of Mechanism, Electronic and Control Engineering, Beijing Jiaotong University, Beijing 100044 (China); Liu, Cunxi [Key laboratory of light duty gas turbine, Institute of Engineering Thermophysics, Chinese Academy of Sciences, Beijing 100190 (China); Chen, Qi [School of Mechanism, Electronic and Control Engineering, Beijing Jiaotong University, Beijing 100044 (China); Xu, Gang; Liu, Fuqiang [Key laboratory of light duty gas turbine, Institute of Engineering Thermophysics, Chinese Academy of Sciences, Beijing 100190 (China)

    2017-04-04

    A swirler based on dielectric barrier discharge plasma actuators is designed and its effectiveness in both jet flow and premixed jet flame control is demonstrated. In contrast to traditional spanwise-oriented actuators, plasma actuators are placed along the axial direction of the injector to induce a circumferential velocity to the main flow and create a swirl flow without any insertion or moving part. In the DBD plasma swirl injector, the discharge does not ignite the mixture nor does it induce flashback. Flame visualization is obtained by cameras while velocity profiles are obtained by Laser Doppler Anemometry measurements. The results obtained indicate the effectiveness of the new design. - Highlights: • The discharge does not ignite the mixture nor does it induce flashback. • The prominent advantage of this novel plasma swirler is its swirl number adjustable without any mechanical movement. • The frequency of the plasma swirler is adjustable. • The plasma swirler can be used as an oscillator to the reactants. • The plasma swirler can be used alone or combine with other traditional swirlers.

  3. ACVP-02: Plasma SIV/SHIV RNA Viral Load Measurements through the AIDS and Cancer Virus Program Quantitative Molecular Diagnostics Core | Frederick National Laboratory for Cancer Research

    Science.gov (United States)

    The SIV plasma viral load assay performed by the Quantitative Molecular Diagnostics Core (QMDC) utilizes reagents specifically designed to detect and accurately quantify the full range of SIV/SHIV viral variants and clones in common usage in the rese

  4. Control of horizontal plasma position by feedforward-feedback system with digital computer in JIPP T-II tokamak

    International Nuclear Information System (INIS)

    Toi, K.; Sakurai, K.; Itoh, S.; Matsuura, K.; Tanahashi, S.

    1980-01-01

    In the resistive shell tokamak, JIPP T-II, the control of horizontal plasma position is successfully carried out by calculating the equilibrium equation in a thin resistive shell from a large-aspect-ratio approximation every 1.39 msec with a digital computer. The iron core effect also is taken account by a simple form in the equation. The required strength of vertical field is determined by the control-demand composed of a ''feedback'' term with Proportion-Integration-Differentiation correction (PID-controller) and ''feedforward'' one in proportion to plasma current. The experimental results have a satisfactory agreement with the analysis of control system. By this control system, the horizontal displacement has been suppressed within 1 cm throughout a discharge for the plasma of 15 cm-radius with high density and low q(a)-value obtained by the second current rise and strong gas puffing. (author)

  5. Design advances of the Core Plasma Thomson Scattering diagnostic for ITER.

    Czech Academy of Sciences Publication Activity Database

    Scannell, R.; Maslov, M.; Naylor, G.; O’Gorman, T.; Kempenaars, M.; Carr, M.; Bílková, Petra; Böhm, Petr; Giudicotti, L.; Pasqualotto, R.; Bassan, M.; Vayakis, G.; Walsh, M.; Huxford, R.

    2017-01-01

    Roč. 12, November (2017), č. článku C11010. ISSN 1748-0221. [International Symposium on Laser-Aided Plasma Diagnostics (LAPD2017) /18./. Prague, 24.09.2017-28.09.2017] Institutional support: RVO:61389021 Keywords : Nuclear instruments and methods for hot plasma diagnostics * Plasma diagnostics - interferometry * spectroscopy and imaging Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.220, year: 2016 http://iopscience.iop.org/article/10.1088/1748-0221/12/11/C11010/pdf

  6. Bluff Body Flow Control Using Dielectric Barrier Discharge Plasma Actuators

    Science.gov (United States)

    Thomas, Flint; Kozlov, Alexey

    2008-11-01

    The results of an experimental investigation involving the use of dielectric barrier discharge plasma actuators to control bluff body flow is presented. The motivation for the work is plasma landing gear noise control for commercial transport aircraft. For these flow control experiments, the cylinder in cross-flow is chosen for study since it represents a generic flow geometry that is similar in all essential aspects to a landing gear strut. The current work is aimed both at extending the plasma flow control concept to Reynolds numbers typical of landing approach and take-off and on the development of optimum plasma actuation strategies. The cylinder wake flow with and without actuation are documented in detail using particle image velocimetry (PIV) and constant temperature hot-wire anemometry. The experiments are performed over a Reynolds number range extending to ReD=10^5. Using either steady or unsteady plasma actuation, it is demonstrated that even at the highest Reynolds number Karman shedding is totally eliminated and turbulence levels in the wake decrease by more than 50%. By minimizing the unsteady flow separation from the cylinder and associated large-scale wake vorticity, the radiated aerodynamic noise is also reduced.

  7. Architectural concept for the ITER Plasma Control System

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Humphreys, D., E-mail: humphreys@fusion.gat.com [General Atomics, San Diego, CA (United States); Raupp, G., E-mail: Gerhard.Raupp@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Schuster, E., E-mail: schuster@lehigh.edu [Lehigh University, Bethlehem, PA (United States); Snipes, J., E-mail: Joseph.Snipes@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France); De Tommasi, G., E-mail: detommas@unina.it [CREATE/Università di Napoli Federico II, Napoli (Italy); Walker, M., E-mail: walker@fusion.gat.com [General Atomics, San Diego, CA (United States); Winter, A., E-mail: Axel.Winter@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France)

    2014-05-15

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  8. Architectural concept for the ITER Plasma Control System

    International Nuclear Information System (INIS)

    Treutterer, W.; Humphreys, D.; Raupp, G.; Schuster, E.; Snipes, J.; De Tommasi, G.; Walker, M.; Winter, A.

    2014-01-01

    The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

  9. Lithium technologies for edge plasma control

    International Nuclear Information System (INIS)

    Sergeev, Vladimir Yu.; Kuteev, Boris V.; Bykov, Aleksey S.; Krylov, Sergey V.; Skokov, Viacheslav G.; Timokhin, Vladimir M.

    2012-01-01

    Highlights: ► We have investigated two new modes of operation been in T-10 limiter tokamak experiments with a novel rotary feeder of lithium dust. ► The observed decreases of bolometer and D β signals, with increase of the electron density during the lithium dust injection, reveal the effects of the first wall conditioning. ► The lithium technology may provide inherent safety mission for major disruption mitigation in a tokamak reactor, which requires demonstration in contemporary tokamak experiments. - Abstract: We have investigated two new modes of operation been in T-10 limiter tokamak experiments with a novel rotary feeder of lithium dust. Quasi steady-state mode I and pulse mode II of dust delivery were realized in both OH and OH + ECRH disruption free plasmas at the lithium flow rate up to 2 × 10 21 atoms/s. A higher flow rate in mode II with injection rate of ∼5 × 10 21 atoms/s caused a series of minor disruptions, which was completed by discharge termination after the major disruption. The observed decreases of bolometer and D β signals, with increase of the electron density during the lithium dust injection, reveal the effects of the first wall conditioning. The lithium technology may provide inherent safety pathway for major disruption mitigation in a tokamak reactor, which requires demonstration in contemporary tokamak experiments.

  10. New DIII-D tokamak plasma control system

    International Nuclear Information System (INIS)

    Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter

  11. James Webb Space Telescope Core 2 Test - Cryogenic Thermal Balance Test of the Observatorys Core Area Thermal Control Hardware

    Science.gov (United States)

    Cleveland, Paul; Parrish, Keith; Thomson, Shaun; Marsh, James; Comber, Brian

    2016-01-01

    The James Webb Space Telescope (JWST), successor to the Hubble Space Telescope, will be the largest astronomical telescope ever sent into space. To observe the very first light of the early universe, JWST requires a large deployed 6.5-meter primary mirror cryogenically cooled to less than 50 Kelvin. Three scientific instruments are further cooled via a large radiator system to less than 40 Kelvin. A fourth scientific instrument is cooled to less than 7 Kelvin using a combination pulse-tube Joule-Thomson mechanical cooler. Passive cryogenic cooling enables the large scale of the telescope which must be highly folded for launch on an Ariane 5 launch vehicle and deployed once on orbit during its journey to the second Earth-Sun Lagrange point. Passive cooling of the observatory is enabled by the deployment of a large tennis court sized five layer Sunshield combined with the use of a network of high efficiency radiators. A high purity aluminum heat strap system connects the three instrument's detector systems to the radiator systems to dissipate less than a single watt of parasitic and instrument dissipated heat. JWST's large scale features, while enabling passive cooling, also prevent the typical flight configuration fully-deployed thermal balance test that is the keystone of most space missions' thermal verification plans. This paper describes the JWST Core 2 Test, which is a cryogenic thermal balance test of a full size, high fidelity engineering model of the Observatory's 'Core' area thermal control hardware. The 'Core' area is the key mechanical and cryogenic interface area between all Observatory elements. The 'Core' area thermal control hardware allows for temperature transition of 300K to approximately 50 K by attenuating heat from the room temperature IEC (instrument electronics) and the Spacecraft Bus. Since the flight hardware is not available for test, the Core 2 test uses high fidelity and flight-like reproductions.

  12. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    Energy Technology Data Exchange (ETDEWEB)

    Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)

    2016-10-15

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

  13. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    International Nuclear Information System (INIS)

    Mitrishkin, Yuri V.; Pavlova, Evgeniia A.; Kuznetsov, Evgenii A.; Gaydamaka, Kirill I.

    2016-01-01

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

  14. Real time plasma control experiments using the JET auxiliary plasma heating systems as the actuator

    International Nuclear Information System (INIS)

    Zornig, N.H.

    1999-01-01

    The role of the Real Time Power Control system (RTPC) in the Joint European Torus (JET) is described in depth. The modes of operation are discussed in detail and a number of successful experiments are described. These experiments prove that RTPC can be used for a wide range of experiments, including: (1) Feedback control of plasma parameters in real time using Ion Cyclotron Resonance Heating (ICRH) or Neutral Beam Heating (NBH) as the actuator in various JET operating regimes. It is demonstrated that in a multi-parameter space it is not sufficient to control one global plasma parameter in order to avoid performance limiting events. (2) Restricting neutron production and subsequent machine activation resulting from high performance pulses. (3) The simulation of α-particle heating effects in a DT-plasma in a D-only plasma. The heating properties of α-particles are simulated using ICRH-power, which is adjusted in real time. The simulation of α-particle heating in JET allows the effects of a change in isotopic mass to be separated from α-particle heating. However, the change in isotopic mass of the plasma ions appears to affect not only the global energy confinement time (τ E ) but also other parameters such as the electron temperature at the plasma edge. This also affects τ E , making it difficult to make a conclusive statement about any isotopic effect. (4) For future JET experiments a scheme has been designed which simulates the behaviour of a fusion reactor experimentally. The design parameters of the International Thermonuclear Experimental Reactor (ITER) are used. In the proposed scheme the most relevant dimensionless plasma parameters are similar in JET and ITER. It is also shown how the amount of heating may be simulated in real time by RTPC using the electron temperature and density as input parameters. The results of two demonstration experiments are presented. (author)

  15. Structure of the automatic system for plasma equilibrium position control

    International Nuclear Information System (INIS)

    Gubarev, V.F.; Krivonos, Yu.G.; Samojlenko, Yu.I.; Snegur, A.A.

    1978-01-01

    Considered are the principles of construction of the automatic system for plasma filament equilibrium position control inside the discharge chamber for the installation of a tokamak type. The combined current control system in control winding is suggested. The most powerful subsystem creates current in the control winding according to the program calculated beforehand. This system provides plasma rough equilibrium along the ''big radius''. The subsystem performing the current change in small limits according to the principle of feed-back coupling is provided simultaneously. The stabilization of plasma position is achieved in the discharge chamber. The advantage of construction of such system is in decreasing of the automatic requlator power without lowering the requirements to the accuracy of equilibrium preservation. The subsystem of automatic control of plasma position over the vertical is put into the system. Such an approach to the construction of the automatic control system proves to be correct; it is based on the experience of application of similar devices for some existing thermonuclear plants

  16. Combined plasma gas-phase synthesis and colloidal processing of InP/ZnS core/shell nanocrystals

    OpenAIRE

    Hue Ryan; Gladfelter Wayne; Gresback Ryan; Kortshagen Uwe

    2011-01-01

    Abstract Indium phosphide nanocrystals (InP NCs) with diameters ranging from 2 to 5 nm were synthesized with a scalable, flow-through, nonthermal plasma process at a rate ranging from 10 to 40 mg/h. The NC size is controlled through the plasma operating parameters, with the residence time of the gas in the plasma region strongly influencing the NC size. The NC size distribution is narrow with the standard deviation being less than 20% of the mean NC size. Zinc sulfide (ZnS) shells were grown ...

  17. Modelling of reactor control and protection systems in the core simulator program GARLIC

    International Nuclear Information System (INIS)

    Beraha, D.; Lupas, O.; Ploegert, K.

    1984-01-01

    For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

  18. Implications of wall recycling and carbon source locations on core plasma fueling and impurity content in DIII-D

    International Nuclear Information System (INIS)

    Groth, M.; Porter, G.D.; Fenstermacher, M.E.; Lasnier, C.J.; Meyer, W.M.; Rensink, M.E.; Wolf, N.S.; Boedo, J.A.; Moyer, R.A.; Rudakov, D.L.; Brooks, N.H.; Groebner, R.J.; Petrie, T.W.; Owen, L.W.; Wang, G.; Zeng, L.; Watkins, J.G.

    2005-01-01

    Measurement and modeling of the 2-D poloidal D α intensity distribution in DIII-D low and medium density L-mode and ELMy H-mode plasmas indicate that hydrogen neutrals predominantly fuel the core from the divertor X-point region. The 2-D distribution of neutral deuterium and low-charge-state carbon were measured in the divertor and the high-field side midplane scrape-off layer (SOL) using tangentially viewing cameras. The emission in the high-field SOL at the equatorial plane was found to be three to four orders of magnitude lower than at the strike points in the divertor, suggesting a strong divertor particle source. Modeling using the UEDGE/DEGAS codes predicted the poloidal fueling distribution to be dependent on the direction of the ion Bx∇B drift. In plasmas with the Bx∇B drift into the divertor stronger fueling from the inner divertor than from the outer is predicted, due to a lower-temperature and higher-density plasma in the inner leg. UEDGE simulations with carbon produced by both physical and chemical sputtering at the divertor plates and walls only are in agreement with a large set of diagnostic data. The simulations indicate flow reversal in the inner divertor that augments the leakage of carbon ions from the divertor into the core. (author)

  19. FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

    International Nuclear Information System (INIS)

    Suratia, Pooja; Patel, Jigneshkumar; Rajpal, Rachana; Kotia, Sorum; Govindarajan, J.

    2012-01-01

    Highlights: ► Evaluation and comparison of the working performance of FLC is done with that of PID Controller. ► FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. ► FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. ► Developed FLC controller is able to maintain the plasma column within required range of ±0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional–Integral–Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

  20. Plasma control using neural network and optical emission spectroscopy

    International Nuclear Information System (INIS)

    Kim, Byungwhan; Bae, Jung Ki; Hong, Wan-Shick

    2005-01-01

    Due to high sensitivity to process parameters, plasma processes should be tightly controlled. For plasma control, a predictive model was constructed using a neural network and optical emission spectroscopy (OES). Principal component analysis (PCA) was used to reduce OES dimensionality. This approach was applied to an oxide plasma etching conducted in a CHF 3 /CF 4 magnetically enhanced reactive ion plasma. The etch process was systematically characterized by means of a statistical experimental design. Three etch outputs (etch rate, profile angle, and etch rate nonuniformity) were modeled using three different approaches, including conventional, OES, and PCA-OES models. For all etch outputs, OES models demonstrated improved predictions over the conventional or PCA-OES models. Compared to conventional models, OES models yielded an improvement of more than 25% in modeling profile angle and etch rate nonuniformtiy. More than 40% improvement over PCA-OES model was achieved in modeling etch rate and profile angle. These results demonstrate that nonreduced in situ data are more beneficial than reduced one in constructing plasma control model

  1. Development in Diagnostics Application to Control Advanced Tokamak Plasma

    International Nuclear Information System (INIS)

    Koide, Y.

    2008-01-01

    For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

  2. Maturing ECRF technology for plasma control

    International Nuclear Information System (INIS)

    Callis, R.W.; .; Cary, W.P.; Chu, S.

    2003-01-01

    The availability of high power, (∼1 MW) long pulse length (effectively cw), high frequency, (>100 GHz) gyrotrons has opened the opportunity for enhanced scientific results on magnetic confinement devices for fusion research worldwide. This has led to successful experiments on electron cyclotron heating, electron cyclotron current drive, non-inductive tokamak operation, tokamak energy transport, suppression of instabilities and advanced profile control leading to enhanced performance. The key development in the gyrotron community that has led to the realization of high power long pulse gyrotrons is the availability of edge cooled synthetic diamond gyrotron output windows, which have low loss and excellent thermal and mechanical properties. In addition to the emergence of reliable high power gyrotrons, ancillary equipment for efficient microwave transmission over distances of hundreds of meters, polarization control, diagnostics, and flexible launch geometry have all been developed and proven in regular service. (author)

  3. Current control for magnetized plasma in direct-current plasma-immersion ion implantation

    International Nuclear Information System (INIS)

    Tang Deli; Chu, Paul K.

    2003-01-01

    A method to control the ion current in direct-current plasma-immersion ion implantation (PIII) is reported for low-pressure magnetized inductively coupled plasma. The ion current can be conveniently adjusted by applying bias voltage to the conducting grid that separates plasma formation and implantation (ion acceleration) zones without the need to alter the rf input power, gas flux, or other operating conditions. The ion current that diminishes with an increase in grid bias in magnetized plasmas can be varied from 48 to 1 mA by increasing the grid voltage from 0 to 70 V at -50 kV sample bias and 0.5 mTorr hydrogen pressure. High implantation voltage and monoenergetic immersion implantation can now be achieved by controlling the ion current without varying the macroscopic plasma parameters. The experimental results and interpretation of the effects are presented in this letter. This technique is very attractive for PIII of planar samples that require on-the-fly adjustment of the implantation current at high implantation voltage but low substrate temperature. In some applications such as hydrogen PIII-ion cut, it may obviate the need for complicated sample cooling devices that must work at high voltage

  4. State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

    2017-02-15

    A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  5. Linear quadratic Gaussian controller design for plasma current, position and shape control system in ITER

    International Nuclear Information System (INIS)

    Belyakov, V.; Kavin, A.; Rumyantsev, E.; Kharitonov, V.; Misenov, B.; Ovsyannikov, A.; Ovsyannikov, D.; Veremei, E.; Zhabko, A.; Mitrishkin, Y.

    1999-01-01

    This paper is focused on the linear quadratic Gaussian (LQG) controller synthesis methodology for the ITER plasma current, position and shape control system as well as power derivative management system. It has been shown that some poloidal field (PF) coils have less influence on reference plasma-wall gaps control during plasma disturbances and hence they have been used to reduce total control power derivative by means of the additional non-linear feedback. The design has been done on the basis of linear models. Simulation was provided for non-linear model and results are presented and discussed. (orig.)

  6. Control Yourself, or at Least Your Core Self

    Science.gov (United States)

    Austin, Lisa M.

    2010-01-01

    Contemporary privacy debates regarding new technologies often define privacy in terms of control over personal information such that the privacy "problem" is a lack of control and the privacy "solution" is increased control. This article questions the control-paradigm by pointing to its parallels with earlier debates in the…

  7. Plasma-assisted synthesis and study of structural and magnetic properties of Fe/C core shell

    Science.gov (United States)

    Shinde, K. P.; Ranot, M.; Choi, C. J.; Kim, H. S.; Chung, K. C.

    2017-07-01

    Pure and carbon-encapsulated iron nanoparticles with an average diameter of 25 nm were synthesized by using the DC plasma arc discharge method. Fe core nanoparticles were encapsulated with carbon layer, which is acting as protection layer against both oxidation and chemical reaction. The morphology and the Fe/C core/shell structure of the nanoparticles were studied by using field emission scanning electron microscopy and transmission electron microscopy. The x-ray diffraction study showed that the α-Fe phase exists with γ-Fe as an impurity. The studied samples have been interrelated with the variation of saturation magnetization, remanent magnetization and coercive field with the amount of carbon coating. The pure α-Fe sample shows saturation magnetization = 172 emu/g, and coercive field = 150 Oe, on the other hand few layer carbon coated α-Fe sample shows saturation magnetization =169 emu/g with higher coercive field 398 Oe.

  8. Plasma-assisted synthesis and study of structural and magnetic properties of Fe/C core shell

    Directory of Open Access Journals (Sweden)

    K. P. Shinde

    2017-07-01

    Full Text Available Pure and carbon-encapsulated iron nanoparticles with an average diameter of 25 nm were synthesized by using the DC plasma arc discharge method. Fe core nanoparticles were encapsulated with carbon layer, which is acting as protection layer against both oxidation and chemical reaction. The morphology and the Fe/C core/shell structure of the nanoparticles were studied by using field emission scanning electron microscopy and transmission electron microscopy. The x-ray diffraction study showed that the α-Fe phase exists with γ-Fe as an impurity. The studied samples have been interrelated with the variation of saturation magnetization, remanent magnetization and coercive field with the amount of carbon coating. The pure α-Fe sample shows saturation magnetization = 172 emu/g, and coercive field = 150 Oe, on the other hand few layer carbon coated α-Fe sample shows saturation magnetization =169 emu/g with higher coercive field 398 Oe.

  9. Nonlinear fluid equations for fully toroidal electromagnetic waves for the core tokamak plasma

    Science.gov (United States)

    Weiland, J.; Liu, C. S.; Liu

    2013-12-01

    The rather general set of fluid equations with full curvature effects (Shukla and Weiland, Phys. Rev. A 40, 341 (1989)) has been modified to apply to the core and generalized to include also microtearing modes.

  10. Mathematical modelling of powder material motion and transportation in high-temperature flow core during plasma coatings application

    Science.gov (United States)

    Bogdanovich, V. I.; Giorbelidze, M. G.

    2018-03-01

    A problem of mathematical modelling of powder material motion and transportation in gas thermal flow core has been addressed. Undertaken studies indicate significant impact on dynamics of motion of sprayed particles of phenomenological law for drag coefficient and accounting momentum loss of a plasma jet upon acceleration of these particles and their diameter. It is determined that at great dispersion of spraying particles, they reach detail surface at different velocity and significant particles separation takes place at spraying spot. According to the results of mathematical modelling, requirements for admissible dispersion of diameters of particles used for spraying have been formulated. Research has also allowed reducing separation of particles at the spraying spot due to the selection of the method of powder feed to the anode channel of the plasma torch.

  11. Ignition and burn control characteristics of thermonuclear plasmas

    International Nuclear Information System (INIS)

    Chaniotakis, E.A.

    1990-01-01

    Achieving the long sought goal of fusion energy requires the attainment of an ignited and controlled thermonuclear plasma. Obtaining an ignited plasma in a tokamak device requires consideration of both the physics of the plasma and the engineering of the machine. With the aide of completely analytical procedure optimized and ignited tokamaks are obtained under various physics assumptions. These designs show the possible advantage of tokamaks characterized by high (∼4.5) aspect ratio, and high (∼15 T) toroidal magnetic field. The control of an ignited plasma is investigated by using auxiliary power modulation. With auxiliary power stable operating points can be created with Q ∼50. Recognizing the need for a fast 1 1/2-D transport model for studying profile effects the plasma transport equations are solved using variational methods. A computer model based on the variational method has been developed. This model solves the 1 1/2-D transport equation very fast with little loss of accuracy. 74 refs., 70 figs., 8 tabs

  12. Real time control of plasmas and ECRH systems on TCV

    NARCIS (Netherlands)

    Paley, J.I.; Berrino, J.; Coda, S.; Cruz, N.; Duval, B.P.; Felici, F.; Goodman, T.P.; Martin, Y.; Moret, J.-M.; Piras, F.; Rodriques, A.P.; Santos, B.; Varandas, C.A.F.

    2009-01-01

    Developments in the real time control hardware on Tokamak Configuration Variable (TCV) coupled with the flexibility of plasma shaping and electron cyclotron (EC) heating and current drive actuators are opening many opportunities to perform real time experiments and develop algorithms and methods for

  13. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-01-01

    The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)

  14. Control and metrology of high harmonic generation on plasma mirrors

    International Nuclear Information System (INIS)

    Monchoce, Sylvain

    2014-01-01

    When an ultra intense femtosecond laser with high contrast is focused on a solid target, the laser field at focus is sufficient enough to completely ionize the target surface during the rising edge of the laser pulse and form a plasma. This dense plasma entirely reflects the incident beam in the specular direction: this is a so-called plasma mirror. As the interaction between the laser and the plasma mirror is highly non-linear, it thus leads to the high harmonic generation (HHG) in the reflected beam. In the temporal domain, this harmonic spectrum is associated to a train of atto-second pulses. The aim of my PhD were to experimentally control this HHG and to measure the properties of the harmonics. We first studied the optimization of the harmonic signal, and then the spatial characterization of the harmonic beam in the far-field (harmonic divergence). These characterizations are not only important to develop an intense XUV/atto-second light source, but also to get a better understanding of the laser-matter interaction at very high intensity. We have thus been able to get crucial information of the electrons and ions dynamics of the plasma, showing that the harmonics can also be used as a diagnostic of the laser-plasma interaction. We then developed a new general approach for optically-controlled spatial structuring of overdense plasmas generated at the surface of initially plain solid targets. We demonstrate it experimentally by creating sinusoidal plasma gratings of adjustable spatial periodicity and depth, and study the interaction of these transient structures with an ultra-intense laser pulse to establish their usability at relativistically high intensities. We then show how these gratings can be used as a 'spatial ruler' to determine the source size of the high-order harmonic beams produced at the surface of an overdense plasma. These results open new directions both for the metrology of laser-plasma interactions and the emerging field of ultrahigh

  15. Multitude of Core-Localized Shear Alfvén Waves in a High-Temperature Fusion Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Nazikian, R. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Berk, H. L. [Univ. of Texas, Austin, TX (United States); Budny, R. V. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Burrell, K. H. [General Atomics, San Diego, CA (United States); Doyle, E. J. [Univ. of California, Los Angeles, CA (United States); Fonck, R. J. [Univ. of Wisconsin, Madison, WI (United States); Gorelenkov, N. N. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Holcomb, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kramer, G. J. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Jayakumar, R. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); La Haye, R. J. [General Atomics, San Diego, CA (United States); McKee, G. R. [Univ. of Wisconsin, Madison, WI (United States); Makowski, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Peebles, W. A. [Univ. of California, Los Angeles, CA (United States); Rhodes, T. L. [Univ. of California, Los Angeles, CA (United States); Solomon, W. M. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Strait, E. J. [General Atomics, San Diego, CA (United States); VanZeeland, M. A. [Oak Ridge Institute for Science and Education (ORISE), Oak Ridge, TN (United States); Zeng, L. [Univ. of California, Los Angeles, CA (United States)

    2006-03-01

    Evidence is provided for a multitude of discrete frequency Alfvén waves in the core of magnetically confined high-temperature fusion plasmas. Multiple diagnostic instruments verify wave excitation over a wide spatial range from the device size at the longest wavelengths down to the thermal ion Larmor radius. At the shortest scales, the poloidal wavelengths are like the scale length of electrostatic drift wave turbulence. Theoretical analysis verifies a dominant interaction of the modes with particles in the thermal ion distribution traveling well below the Alfvén velocity.

  16. Study of neutral hydrogen transport in LHD core plasmas based on high dynamic-range Balmer-α spectroscopy

    International Nuclear Information System (INIS)

    Fujii, K.; Goto, M.; Morita, S.

    2014-10-01

    The radial distributions of the neutral hydrogen atom density and pressure in the Large Helical Device (LHD) were studied. The atom density distribution was determined from a detailed analysis of the intensity-calibrated Balmer-α line profile while the pressure distribution was obtained with a simple one-dimensional analytical model. We determined for the first time the atom density at the center of a fusion-oriented plasma, which is approximately three orders smaller than that at the edge. On the contrary, the atom pressure changes only a factor of 10 from the edge to core regions. (author)

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

    International Nuclear Information System (INIS)

    Rice, J.E.; Reinke, M.L.; Podpaly, Y.A.; Churchill, R.M.; Cziegler, I.; Dominguez, A.; Ennever, P.C.; Fiore, C.L.; Granetz, R.S.; Greenwald, M.J.; Hubbard, A.E.; Hughes, J.W.; Irby, J.H.; Ma, Y.; Marmar, E.S.; McDermott, R.M.; Porkolab, M.; Duval, B.P.; Bortolon, A.; Diamond, P.H.

    2011-01-01

    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 k R ≥ 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.

  18. Experimental validation of a Lyapunov-based controller for the plasma safety factor and plasma pressure in the TCV tokamak

    Science.gov (United States)

    Mavkov, B.; Witrant, E.; Prieur, C.; Maljaars, E.; Felici, F.; Sauter, O.; the TCV-Team

    2018-05-01

    In this paper, model-based closed-loop algorithms are derived for distributed control of the inverse of the safety factor profile and the plasma pressure parameter β of the TCV tokamak. The simultaneous control of the two plasma quantities is performed by combining two different control methods. The control design of the plasma safety factor is based on an infinite-dimensional setting using Lyapunov analysis for partial differential equations, while the control of the plasma pressure parameter is designed using control techniques for single-input and single-output systems. The performance and robustness of the proposed controller is analyzed in simulations using the fast plasma transport simulator RAPTOR. The control is then implemented and tested in experiments in TCV L-mode discharges using the RAPTOR model predicted estimates for the q-profile. The distributed control in TCV is performed using one co-current and one counter-current electron cyclotron heating actuation.

  19. Real time control of plasmas and ECRH systems on TCV

    International Nuclear Information System (INIS)

    Paley, J.I.; Berrino, J.; Coda, S.; Duval, B.P.; Felici, F.; Goodman, T.P.; Martin, Y.; Moret, J.M.; Piras, F.; Cruz, N.; Rodriques, A.P.; Santos, B.; Varandas, C.A.F.

    2009-01-01

    Developments in the real time control hardware on Tokamak a Configuration Variable (TCV) coupled with the flexibility of plasma shaping and electron cyclotron (EC) heating and current drive actuators are opening many opportunities to perform real time experiments and develop algorithms and methods for fusion applications. The ability to control magnetohydrodynamic instabilities is particularly important for achieving high performance fusion plasmas and EC is envisaged as a key actuator in maintaining high performance. We have successfully demonstrated control of the sawtooth instability using the EC launcher injection angle to modify the current profile around the q =1 surface. This paper presents an overview of recent real time control experiments on TCV, developments in the hardware and algorithms together with plans for the future.

  20. Adaptive control of a PWR core power using neural networks

    International Nuclear Information System (INIS)

    Arab-Alibeik, H.; Setayeshi, S.

    2005-01-01

    Reactor power control is important because of safety concerns and the call for regular and appropriate operation of nuclear power plants. It seems that the load-follow operation of these plants will be unavoidable in the future. Discrepancies between the real plant and the model used in controller design for load-follow operation encourage one to use auto-tuning and (or) adaptive techniques. Neural network technology shows great promise for addressing many problems in non-model-based adaptive control methods. Also, there has been a great attention to inverse control especially in the neural and fuzzy control context. Fortunately, online adaptation eliminates some limitations of inverse control and its shortcomings for real world applications. We use a neural adaptive inverse controller to control the power of a PWR reactor. The stability of the system and convergence of the controller parameters are guaranteed during online adaptation phase provided the controller is near the plant's real inverse after offline training period. The performance of the controller is verified using nonlinear simulations in diverse operating conditions

  1. New achievements in the EAST plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Q.P., E-mail: qpyuan@ipp.ac.c [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Penaflor, B.G.; Piglowski, D.A. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States); Liu, L.Z. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Johnson, R.D.; Walker, M.L.; Humphreys, D.A. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States)

    2010-07-15

    In order to realize the low latency and distortion-free signal transmission between the plasma control system (PCS) and servo systems, the digital output structure configured with reflective memory board (RFM) was adopted in EAST PCS. And the enhanced performances are reported. Another achievement made in the latest EAST PCS was the implementation of density control algorithm, which controlled the line average density in either voltage or width modulation mode. The new integrated algorithm improved the precision of density calculation and control performance greatly. The details and experiment results are presented in this paper.

  2. Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

    Science.gov (United States)

    Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

    2018-05-01

    ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on

  3. Software development for the PBX-M plasma control system

    International Nuclear Information System (INIS)

    Lagin, L.; Bell, R.; Chu, J.; Hatcher, R.; Hirsch, J.; Okabayashi, M.; Sichta, P.

    1995-01-01

    This paper describes the software development effort for the PBX-M plasma control system. The algorithms being developed for the system will serve to test advanced control concepts for TPX and ITER. This will include real-time algorithms for shaping control, vertical position control, current and density profile control and MHD avoidance. The control system consists of an interactive Host Processor (SPARC-10) interfaced through VME with four real-time Computer Processors (i860) which run at a maximum computational speed of 320 MFLOPs. Plasma shaping programs are being tested to duplicate the present PBX-M analog control system. Advanced algorithms for vertical control and x-point control will then be developed. Interactive graphical user interface programs running on the Host Processor will allow operators to control and monitor shot parameters. A waveform edit program will be used to download pre-programmed waveforms into the Compute Processor memory. Post-shot display programs will be used to interactively display data after the shot. Automatic pre-shot arming and data acquisition programs will run on the Host Processor. Event system programs will process interrupts and activate programs on the Host and Compute Processors. These programs are being written in C and Fortran and use system service routines to communicate with the Compute Processors and its memory. IDL and IDL widgets are being used to build the graphical user interfaces

  4. Plasma shape control by pulsed solenoid on laser ion source

    International Nuclear Information System (INIS)

    Sekine, M.; Ikeda, S.; Romanelli, M.; Kumaki, M.; Fuwa, Y.; Kanesue, T.; Hayashizaki, N.; Lambiase, R.; Okamura, M.

    2015-01-01

    A Laser ion source (LIS) provides high current heavy ion beams with a very simple mechanical structure. Plasma is produced by a pulsed laser ablation of a solid state target and ions are extracted by an electric field. However, it was difficult to manipulate the beam parameters of a LIS, since the plasma condition could only be adjusted by the laser irradiation condition. To enhance flexibility of LIS operation, we employed a pulsed solenoid in the plasma drift section and investigated the effect of the solenoid field on singly charged iron beams. The experimentally obtained current profile was satisfactorily controlled by the pulsed magnetic field. This approach may also be useful to reduce beam emittance of a LIS

  5. Plasma shape control by pulsed solenoid on laser ion source

    Science.gov (United States)

    Sekine, M.; Ikeda, S.; Romanelli, M.; Kumaki, M.; Fuwa, Y.; Kanesue, T.; Hayashizaki, N.; Lambiase, R.; Okamura, M.

    2015-09-01

    A Laser ion source (LIS) provides high current heavy ion beams with a very simple mechanical structure. Plasma is produced by a pulsed laser ablation of a solid state target and ions are extracted by an electric field. However, it was difficult to manipulate the beam parameters of a LIS, since the plasma condition could only be adjusted by the laser irradiation condition. To enhance flexibility of LIS operation, we employed a pulsed solenoid in the plasma drift section and investigated the effect of the solenoid field on singly charged iron beams. The experimentally obtained current profile was satisfactorily controlled by the pulsed magnetic field. This approach may also be useful to reduce beam emittance of a LIS.

  6. Plasma shape control by pulsed solenoid on laser ion source

    Energy Technology Data Exchange (ETDEWEB)

    Sekine, M. [Tokyo Institute of Technology, Meguro-ku, Tokyo 2-12-1 (Japan); RIKEN, Wako, Saitama 351-0198 (Japan); Ikeda, S. [Tokyo Institute of Technology, Yokohama, Kanagawa 226-8502 (Japan); RIKEN, Wako, Saitama 351-0198 (Japan); Romanelli, M. [Cornell University, Ithaca, NY 14850 (United States); Kumaki, M. [RIKEN, Wako, Saitama 351-0198 (Japan); Waseda University, Shinjuku, Tokyo 169-0072 (Japan); Fuwa, Y. [RIKEN, Wako, Saitama 351-0198 (Japan); Kyoto University, Uji, Kyoto 611-0011 (Japan); Kanesue, T. [Brookhaven National Laboratory, Upton, NY 11973 (United States); Hayashizaki, N. [Tokyo Institute of Technology, Meguro-ku, Tokyo 2-12-1 (Japan); Lambiase, R. [Brookhaven National Laboratory, Upton, NY 11973 (United States); Okamura, M. [RIKEN, Wako, Saitama 351-0198 (Japan); Brookhaven National Laboratory, Upton, NY 11973 (United States)

    2015-09-21

    A Laser ion source (LIS) provides high current heavy ion beams with a very simple mechanical structure. Plasma is produced by a pulsed laser ablation of a solid state target and ions are extracted by an electric field. However, it was difficult to manipulate the beam parameters of a LIS, since the plasma condition could only be adjusted by the laser irradiation condition. To enhance flexibility of LIS operation, we employed a pulsed solenoid in the plasma drift section and investigated the effect of the solenoid field on singly charged iron beams. The experimentally obtained current profile was satisfactorily controlled by the pulsed magnetic field. This approach may also be useful to reduce beam emittance of a LIS.

  7. Automation of Aditya vacuum control system based on CODAC Core System

    Energy Technology Data Exchange (ETDEWEB)

    Raulji, Vismaysinh D., E-mail: vismay@ipr.res.in; Pujara, Harshad; Arambhadiya, Bharat; Jadeja, Kumarpalsinh; Bhatt, Shailesh; Rajpal, Rachana

    2016-11-15

    Highlights: • Monitor and control of vacuum control system based on CODAC Core System. • Communication between SIEMENS PLC and open source software EPICS. • With CODAC Core easy to configure and programming of slow controller. - Abstract: The main objective of vacuum control system is to provide ultrahigh vacuum for Aditya Tokamak operations. Aditya Vacuum vessel is having four vacuum pumping lines. To demonstrate implementation of automation; a study case is under taken by automating single Pumping Line of the Aditya vacuum system using CODAC Core System (CCS). Currently, vacuum system is operated manually. The CCS based control system allows remote control, monitoring, alarm handling of vacuum parameters. The CODAC Core System is the Linux based software package that is distributed by ITER Organization for the development of Plant System I&C software. CODAC Core System includes EPICS, CSS (Control System Studio) etc. CSS is used for HMI (Human Machine Interface), alarms and archives. SDD (Self Description Data) tool is used to configure plant system I&C. SDD Editor is an Eclipse based application to define the plant system, interface, I&C component, interfaced signals, configure variable. SCADA (Supervisory Control and Data Acquisition) system is developed in CSS. Data is transferred between PLC and CSS through EPICS. The complete system is tested with Aditya Vacuum Control System with process interlocks. Operator interface is also developed using Lab VIEW as a choice of the user. This paper will describe the salient features of the developed control system in detail.

  8. Radiological Control Technician: Phase 2, Core/site practical training

    International Nuclear Information System (INIS)

    1992-10-01

    This volume provides a description of the guidelines, qualification in radiological instrumentation, qualification in radiological protection, and qualification in emergency preparedness for radiological control technicians

  9. Active Control of Power Exhaust in Strongly Heated ASDEX Upgrade Plasmas

    Science.gov (United States)

    Dux, Ralph; Kallenbach, Arne; Bernert, Matthias; Eich, Thomas; Fuchs, Christoph; Giannone, Louis; Herrmann, Albrecht; Schweinzer, Josef; Treutterer, Wolfgang

    2012-10-01

    Due to the absence of carbon as an intrinsic low-Z radiator, and tight limits for the acceptable power load on the divertor target, ITER will rely on impurity seeding for radiative power dissipation and for generation of partial detachment. The injection of more than one radiating species is required to optimise the power removal in the main plasma and in the divertor region, i.e. a low-Z species for radiation in the divertor and a medium-Z species for radiation in the outer core plasma. In ASDEX Upgrade, a set of robust sensors, which is suitable to feedback control the radiated power in the main chamber and the divertor as well as the electron temperature at the target, has been developed. Different feedback schemes were applied in H-mode discharges with a maximum heating power of up to 23,W, i.e. at ITER values of P/R (power per major radius) to control all combinations of power flux into the divertor region, power flux onto the target or electron temperature at the target through injection of nitrogen as the divertor radiator and argon as the main chamber radiator. Even at the highest heating powers the peak heat flux density at the target is kept at benign values. The control schemes and the plasma behaviour in these discharges will be discussed.

  10. Remote network control plasma diagnostic system for Tokamak T-10

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Optimal control of tokamak and stellarator plasma behaviour

    International Nuclear Information System (INIS)

    Rastovic, Danilo

    2007-01-01

    The control of plasma transport, laminar and turbulent, is investigated, using the methods of scaling, optimal control and adaptive Monte Carlo simulations. For this purpose, the asymptotic behaviour of kinetic equation is considered in order to obtain finite-dimensional invariant manifolds, and in this way the finite-dimensional theory of control can be applied. We imagine the labyrinth of open doors and after applying self-similarity, the motion moved through all the desired doors in repeatable ways as Brownian motions. We take local actions for each piece of contractive ergodic motion, and, after self-organization in adaptive invariant measures, the optimum movement of particles is obtained according to the principle of maximum entropy. This is true for deterministic and stochastic cases that serve as models for plasma dynamics

  12. GYRO Simulations of Core Momentum Transport in DIII-D and JET Plasmas

    International Nuclear Information System (INIS)

    Budny, R.V.; Candy, J.; Waltz, R.E.

    2005-01-01

    Momentum, energy, and particle transport in DIII-D and JET ELMy H-mode plasmas is simulated with GYRO and compared with measurements analyzed using TRANSP. The simulated transport depends sensitively on the nabla(T(sub)i) turbulence drive and the nabla(E(sub)r) turbulence suppression inputs. With their nominal values indicated by measurements, the simulations over-predict the momentum and energy transport in the DIII-D plasmas, and under-predict in the JET plasmas. Reducing |nabla(T(sub)i)| and increasing |nabla(E(sub)r)| by up to 15% leads to approximate agreement (within a factor of two) for the DIII-D cases. For the JET cases, increasing |nabla(T(sub)i)| or reducing |nabla(E(sub)r)| results in approximate agreement for the energy flow, but the ratio of the simulated energy and momentum flows remains higher than measurements by a factor of 2-4

  13. Augmentation and Control of Burn Rates in Plasma Devices

    National Research Council Canada - National Science Library

    Bourham, Mohamed

    1999-01-01

    Interaction of electrothermal plasmas with solid propellants necessitates thorough understanding of plasma-propellant interface physics, momentum and energy transfer, plasma flow regimes, and mixing processes...

  14. Application of MSHIM core control strategy for westinghouse AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Onoue, Masaaki; Kawanishi, Tomohiro; Carlson, William R.; Morita, Toshio

    2003-01-01

    Westinghouse has developed a new core control strategy, in which two independently moving Rod Cluster Control Assembly (RCCA) groups are utilized for core control; one group for reactivity/temperature control, the other for axial power distribution (Axial Offset) control. This control procedure eliminates the need for Chemical Shim adjustments during power maneuvers, such as load follow, and is designated MSHIM (Mechanical Shim). This core control strategy is utilized in the AP1000. In the AP1000, it is possible to perform MSHIM load follow maneuvers for up to 95% of cycle life without changing the soluble boron concentration in the moderator. This core control strategy has been evaluated, via computer simulations, to provide appropriate margins to core and fuel design limits during normal operation maneuvers (including load follow operations) and also during anticipated Condition II accident transients. The primary benefits of MSHIM as a control strategy are as follows; Power change operation can be totally automated due to the elimination of boron concentration adjustments. Full load follow capability is achievable for up to more than 95% of cycle life. Load follow operations performed solely by mechanical devices results in a significant reduction in the boron system requirements and a significant reduction in daily effluent to be processed. (author)

  15. Interprocess communication within the DIII-D plasma control system

    International Nuclear Information System (INIS)

    Piglowski, D.A.; Penaflor, B.G.; Ferron, J.R.

    1999-06-01

    The DIII-D tokamak fusion research experiment's real-time digital plasma control system (PCS) is a complex and ever evolving system. During a plasma experiment, it is tasked with some of the most crucial functions at DIII-D. Key responsibilities of the PCS involve sub-system control, data acquisition/storage, and user interface. To accomplish these functions, the PCS is broken down into individual components (both software and hardware), each capable of handling a specific duty set. Constant interaction between these components is necessary prior, during and after a standard plasma cycle. Complicating the matter even more is that some components, mostly those which deal with user interaction, may exist remotely, that is to say they are not part of the immediate hardware which makes up the bulk of the PCS. The four main objectives of this paper are to (1) present a brief outline of the PCS hardware/software and how they relate to each other; (2) present a brief overview of a standard DIII-D plasma cycle (a shot); (3) using three sets of PCS sub-systems, describe in more detail the communication processes; and (4) evaluate the benefits and drawbacks of said systems

  16. Dielectric barrier discharge plasma actuator for flow control

    Science.gov (United States)

    Opaits, Dmitry Florievich

    Electrohydrodynamic (EHD) and magnetohydrodynamic phenomena are being widely studied for aerodynamic applications. The major effects of these phenomena are heating of the gas, body force generation, and enthalpy addition or extraction, [1, 2, 3]. In particular, asymmetric dielectric barrier discharge (DBD) plasma actuators are known to be effective EHD device in aerodynamic control, [4, 5]. Experiments have demonstrated their effectiveness in separation control, acoustic noise reduction, and other aeronautic applications. In contrast to conventional DBD actuators driven by sinusoidal voltages, we proposed and used a voltage profile consisting of nanosecond pulses superimposed on dc bias voltage. This produces what is essentially a non-self-sustained discharge: the plasma is generated by repetitive short pulses, and the pushing of the gas occurs primarily due to the bias voltage. The advantage of this non-self-sustained discharge is that the parameters of ionizing pulses and the driving bias voltage can be varied independently, which adds flexibility to control and optimization of the actuators performance. Experimental studies were conducted of a flow induced in a quiescent room air by a single DBD actuator. A new approach for non-intrusive diagnostics of plasma actuator induced flows in quiescent gas was proposed, consisting of three elements coupled together: the Schlieren technique, burst mode of plasma actuator operation, and 2-D numerical fluid modeling. During the experiments, it was found that DBD performance is severely limited by surface charge accumulation on the dielectric. Several ways to mitigate the surface charge were found: using a reversing DC bias potential, three-electrode configuration, slightly conductive dielectrics, and semi conductive coatings. Force balance measurements proved the effectiveness of the suggested configurations and advantages of the new voltage profile (pulses+bias) over the traditional sinusoidal one at relatively low

  17. Electric field bifurcation and transition in the core plasma of CHS

    International Nuclear Information System (INIS)

    Fujisawa, A.; Iguchi, H.; Sanuki, H.; Itoh, K.; Okamura, S.; Matsuoka, K.; Hamada, Y.; Itoh, S.-I.

    1997-01-01

    In the CHS heliotron/torsatron, dynamic phenomena associated with transitions in radial electric field were observed during combined ECH+NBI heated plasmas. The observations with high temporal resolution confirmed a nonlinear relation between radial electric field and radial current to cause these phenomena associated with electric field bifurcation. (author)

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

    NARCIS (Netherlands)

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

    2011-01-01

    The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma

  19. Gyrokinetic analyses of core heat transport in JT-60U plasmas with different toroidal rotation direction

    International Nuclear Information System (INIS)

    Narita, Emi; Fukuda, Takeshi; Honda, Mitsuru; Hayashi, Nobuhiko; Urano, Hajime; Ide, Shunsuke

    2015-01-01

    Tokamak plasmas with an internal transport barrier (ITB) are capable of maintaining improved confinement performance. The ITBs formed in plasmas with the weak magnetic shear and the weak radial electric field shear are often observed to be modest. In these ITB plasmas, it has been found that the electron temperature ITB is steeper when toroidal rotation is in a co-direction with respect to the plasma current than when toroidal rotation is in a counter-direction. To clarify the relationship between the direction of toroidal rotation and heat transport in the ITB region, we examine dominant instabilities using the flux-tube gyrokinetic code GS2. The linear calculations show a difference in the real frequencies; the counter-rotation case has a more trapped electron mode than the co-rotation case. In addition, the nonlinear calculations show that with this difference, the ratio of the electron heat diffusivity χ_e to the ion's χ_i is higher for the counter-rotation case than for the co-rotation case. The difference in χ_e /χ_i agrees with the experiment. We also find that the effect of the difference in the flow shear between the two cases due to the toroidal rotation direction on the linear growth rate is not significant. (author)

  20. Control of ITBs in Magnetically Confined Burning Plasmas

    Science.gov (United States)

    Panta, S. R.; Newman, D. E.; Terry, P. W.; Sanchez, R.

    2017-10-01

    In the magnetically confined burning plasma devices (in this case Tokamaks), internal transport barriers (ITBs) are those regimes in which the turbulence is suppressed by the E X B velocity shear, reducing the turbulent transport. This often occurs at a critical gradient in the profiles. The change in the transport then modifies the density and temperature profiles feeding back on the system. These transport barriers have to be controlled both to form them for improved confinement and remove them to both prevent global instabilities and to remove the ash and unnecessary impurities in the device. In this work we focus on pellet injection and modulated RF heating as a way to trigger and control the ITBs. These have an immediate consequence on density and temperature and hence pressure profiles acting as a control knob. For example, depending upon pellet size and its radial position of injection, it either helps to form or strengthen the barrier or to get rid of ITBs in the different transport channels of the burning plasmas. This transport model is then used to investigate the control and dynamics of the transport barriers in burning plasmas using pellets and RF addition to the NBI power and alpha power.

  1. Experimental Investigation on Airfoil Shock Control by Plasma Aerodynamic Actuation

    International Nuclear Information System (INIS)

    Sun Quan; Cheng Bangqin; Li Yinghong; Cui Wei; Jin Di; Li Jun

    2013-01-01

    An experimental investigation on airfoil (NACA64—215) shock control is performed by plasma aerodynamic actuation in a supersonic tunnel (Ma = 2). The results of schlieren and pressure measurement show that when plasma aerodynamic actuation is applied, the position moves forward and the intensity of shock at the head of the airfoil weakens. With the increase in actuating voltage, the total pressure measured at the head of the airfoil increases, which means that the shock intensity decreases and the control effect increases. The best actuation effect is caused by upwind-direction actuation with a magnetic field, and then downwind-direction actuation with a magnetic field, while the control effect of aerodynamic actuation without a magnetic field is the most inconspicuous. The mean intensity of the normal shock at the head of the airfoil is relatively decreased by 16.33%, and the normal shock intensity is relatively reduced by 27.5% when 1000 V actuating voltage and upwind-direction actuation are applied with a magnetic field. This paper theoretically analyzes the Joule heating effect generated by DC discharge and the Lorentz force effect caused by the magnetic field. The discharge characteristics are compared for all kinds of actuation conditions to reveal the mechanism of shock control by plasma aerodynamic actuation

  2. Robust control design for the plasma horizontal position control on J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Yu, W.Z.; Chen, Z.P.; Zhuang, G.; Wang, Z.J.

    2013-01-01

    It is extremely important for tokamak to control the plasma position during routine discharge. However, the model of plasma in tokamak usually contains much of the uncertainty, such as structured uncertainties and unmodeled dynamics. Compared with the traditional PID control approach, robust control theory is more suitable to handle this problem. In the paper, we propose a H ∞ robust control scheme to control the horizontal position of plasma during the flat-top phase of discharge on Joint Texas Experimental Tokamak (J-TEXT) tokamak. First, the model of our plant for plasma horizontal position control is obtained from the position equilibrium equations. Then the H ∞ robust control framework is used to synthesize the controller. Based on this, an H ∞ controller is designed to minimize the regulation/tracking error. Finally, a comparison study is conducted between the optimized H ∞ robust controller and the traditional PID controller in simulations. The simulation results of the H ∞ robust controller show a significant improvement of the performance with respect to those obtained with traditional PID controller, which is currently used on our machine

  3. Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

    International Nuclear Information System (INIS)

    Giannone, L.; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T.

    2009-01-01

    The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

  4. Plasma control techniques of the ASDEX feedback system

    International Nuclear Information System (INIS)

    Schneider, F.

    1981-01-01

    In the ASDEX tokamak the shots are exactly preprogrammed and most of the disturbances are reproducible. So a computer can learn from one shot how to correct the next one. With this sort of disturbance feedforward one can also introduce a 'negative delay' in the program to compensate even fast and strong disturbances withous unwanted overswing or oscillations. The feedforward in conjunction with feedback control allows production of a magnetically limited plasma from the very beginning without any wall or limiter contact. This is a reason why in ASDEX the loop voltage on breakdown can be as low as 5 V/sup 2/. The plasma column can be controlled in the vacuum vessel even after disruptions have occurred

  5. Real-time software for the COMPASS tokamak plasma control

    Czech Academy of Sciences Publication Activity Database

    Valcárcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Sartori, F.; Janky, Filip; Cahyna, Pavel; Hron, Martin; Pánek, Radomír

    2010-01-01

    Roč. 85, 3-4 (2010), s. 470-473 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition and Remote Participation for Fusion Research/7th./. Aix – en – Provence, 15.06.2009-19.06.2009] Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-time * ATCA * Data acquisition * Plasma control software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.143, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3C-4YXMP1Y-3&_user=6542793&_coverDate=07%2F31%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=9005df0735c0dbb3a93a9c154b0d09c1&searchtype=a

  6. Positron Plasma Control Techniques Applied to Studies of Cold Antihydrogen

    CERN Document Server

    Funakoshi, Ryo

    2003-01-01

    In the year 2002, two experiments at CERN succeeded in producing cold antihydrogen atoms, first ATHENA and subsequently ATRAP. Following on these results, it is now feasible to use antihydrogen to study the properties of antimatter. In the ATHENA experiment, the cold antihydrogen atoms are produced by mixing large amounts of antiprotons and positrons in a nested Penning trap. The complicated behaviors of the charged particles are controlled and monitored by plasma manipulation techniques. The antihydrogen events are studied using position sensitive detectors and the evidence of production of antihydrogen atoms is separated out with the help of analysis software. This thesis covers the first production of cold antihydrogen in the first section as well as the further studies of cold antihydrogen performed by using the plasma control techniques in the second section.

  7. User Control Interface for W7-X Plasma Operation

    International Nuclear Information System (INIS)

    Spring, A.; Laqua, H.; Schacht, J.

    2006-01-01

    The WENDELSTEIN 7-X fusion experiment will be a highly complex device operated by a likewise complex control system. The fundamental configuration of the W7-X control system follows two major design principles: It reflects the strict hierarchy of the machine set-up with a set of subordinated components, which in turn can be run autonomously during commissioning and testing. Secondly, it links the basic machine operation (mainly given by the infrastructure status and the components readiness) and the physics program execution (i.e. plasma operation) on each hierarchy level and on different time scales. The complexity of the control system implies great demands on appropriate user interfaces: specialized tools for specific control tasks allowing a dedicated view on the subject to be controlled, hiding complexity wherever possible and reasonable, providing similar operation methods on each hierarchy level and both manual interaction possibilities and a high degree of intelligent automation. The contribution will describe the operation interface for experiment control including the necessary links to the machine operation. The users of ' Xcontrol ' will be both the W7-X session leaders during plasma discharge experiments and the components' or diagnostics' operators during autonomous mode or even laboratory experiments. The main ' Xcontrol ' features, such as program composition and validation, manual and automatic control instruments, resource survey, and process monitoring, will be presented. The implementation principles and the underlying communication will be discussed. (author)

  8. Control of exogenous factors affecting plasma homovanillic acid concentration.

    Science.gov (United States)

    Davidson, M; Giordani, A B; Mohs, R C; Mykytyn, V V; Platt, S; Aryan, Z S; Davis, K L

    1987-04-01

    Measurements of plasma homovanillic acid (pHVA) concentrations appear to be a valid research strategy in psychiatric disorders in which a central dopamine (DA) abnormality has been implicated. This study provides guidance about the control of some of the exogenous factors affecting pHVA concentrations. Fasting for 14 hours eliminates the dietary effects on pHVA in healthy human subjects. Changing position, walking for 30 minutes, or smoking two cigarettes has no effect on pHVA concentrations.

  9. Fundamentals of plasma physics and controlled fusion. The third edition

    International Nuclear Information System (INIS)

    Miyamoto, Kenro

    2011-06-01

    Primary objective of this lecture note is to provide a basic text for the students to study plasma physics and controlled fusion researches. Secondary objective is to offer a reference book describing analytical methods of plasma physics for the researchers. This was written based on lecture notes for a graduate course and an advanced undergraduate course those have been offered at Department of Physics, Faculty of Science, University of Tokyo. In ch.1 and 2, basic concept of plasma and its characteristics are explained. In ch.3, orbits of ion and electron are described in several magnetic field configurations. Chapter 4 formulates Boltzmann equation of velocity space distribution function, which is the basic relation of plasma physics. From ch.5 to ch.9, plasmas are described as magnetohydrodynamic (MHD) fluid. MHD equation of motion (ch.5), equilibrium (ch.6) and diffusion and confinement time of plasma (ch.7) are described by the fluid model. Chapters 8 and 9 discuss problems of MHD instabilities whether a small perturbation will grow to disrupt the plasma or will damp to a stable state. The basic MHD equation of motion can be derived by taking an appropriate average of Boltzmann equation. This mathematical process is described in appendix A. The derivation of useful energy integral formula of axisymmetric toroidal system and the analysis of high n ballooning mode are described in app. B. From ch.10 to ch.14, plasmas are treated by kinetic theory. This medium, in which waves and perturbations propagate, is generally inhomogeneous and anisotropic. It may absorb or even amplify the wave. Cold plasma model described in ch.10 is applicable when the thermal velocity of plasma particles is much smaller than the phase velocity of wave. Because of its simplicity, the dielectric tensor of cold plasma can be easily derived and the properties of various wave can be discussed in the case of cold plasma. If the refractive index becomes large and the phase velocity of the

  10. Control of horizontal plasma position by feedforward-feedback system with digital computer in the JIPP T-II tokamak

    International Nuclear Information System (INIS)

    Toi, K.; Itoh, S.; Sakurai, K.; Matsuura, K.; Tanahashi, S.

    1980-02-01

    In the resistive shell tokamak, JIPP T-II, the control of horizontal plasma position is successfully carried out by calculating the equilibrium equation of a large-aspect-ratio tokamak plasma surrounded by a thin resistive shell of a skin time of 5.2 msec, every 1.39 msec with a digital computer. The iron core effect is also taken into account by a simple form in the equation. The required strength of vertical field is determined by the control demand composed of two groups; one is a ''feedback'' term expressed by the deviation of plasma position from the desired one and proportion-integration-differentiation correction (PID-controller), and the other is a ''feedforward'' term which is in proportion to the plasma current. The experimental results have a good agreement with the stability analysis of the control system by using the so-called Bode-diagram. By this control system, the horizontal displacement has been suppressed within 1 cm from the initiation of discharge to the termination in the high-density and low-q(a) plasma of 15 cm-radius which is obtained by both strong gas puffing and second current rise. (author)

  11. Progress and improvement of KSTAR plasma control using model-based control simulators

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Sang-hee, E-mail: hahn76@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Welander, A.S. [General Atomics, San Diego, CA (United States); Yoon, S.W.; Bak, J.G. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Eidietis, N.W. [General Atomics, San Diego, CA (United States); Han, H.S. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Humphreys, D.A.; Hyatt, A. [General Atomics, San Diego, CA (United States); Jeon, Y.M. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Johnson, R.D. [General Atomics, San Diego, CA (United States); Kim, H.S.; Kim, J. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Kolemen, E.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Penaflor, B.G.; Piglowski, D.A. [General Atomics, San Diego, CA (United States); Shin, G.W. [University of Science and Technology, Daejeon (Korea, Republic of); Walker, M.L. [General Atomics, San Diego, CA (United States); Woo, M.H. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of)

    2014-05-15

    Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (I{sub p}) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.

  12. Plasma position control in a tokamak reactor around ignition

    International Nuclear Information System (INIS)

    Carretta, U.; Minardi, E.; Bacelli, N.

    1986-01-01

    Plasma position control in a tokamak reactor in the phase approaching ignition is closely related to burn control. If ignited burn corresponds to a thermally unstable situation the plasma becomes sensitive to the thermal instability already in the phase when ignition is approached so that the trajectory in the position-pressure (R,p) space becomes effectively unpredictable. For example, schemes involving closed cycles around ignition can be unstable in the heating-cooling phases, and the deviations may be cumulative in time. Reliable plasma control in pressure-position (p, R) space is achieved by beforehand constraining the p, R trajectory rigidly with suitable feedback vertical field stabilization, which is to be established already below ignition. A scheme in which ignition is approached in a stable and automatic way by feedback stabilization on the vertical field is proposed and studied in detail. The values of the gain coefficient ensuring stabilization and the associated p and R excursions are discussed both analytically, with a 0-D approximation including non-linear effects, and numerically with a 1-D code in cylindrical geometry. Profile effects increase the excursions, in particular above ignition. (author)

  13. Plasma shape control calculations for BPX divertor design

    International Nuclear Information System (INIS)

    Strickler, D.J.; Neilson, G.H.; Jardin, S.C.; Pomphrey, N.

    1991-01-01

    The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs

  14. Core density fluctuations in reverse magnetic shear plasmas with internal transport barrier on JT-60U

    International Nuclear Information System (INIS)

    Nazikian, R.; Shinohara, K.; Yoshino, R.; Fujita, T.; Shirai, H.; Kramer, G.T.

    1999-01-01

    First measurements of the radial correlation length of density fluctuations in JT-60U plasmas with internal transport barrier (ITB) is reported. The measurements are obtained using a newly installed correlation reflectometer operating in the upper X-mode. Before transport barrier formation in the low beam power current ramp-up phase of the discharge, reflectometer measurements indicate density fluctuation levels n-tilde/n∼0.1-0.2% and radial correlation lengths 2-3 cm (k r p i ≤0.5) in the central plasma region (r/a r p i ∼3. However, fluctuation levels are considerably higher than measured near the magnetic axis. Reflectometer measurements obtained at the foot of the ITB also indicate high fluctuation levels compared to measurements in the central region of the discharge. (author)

  15. HOT PLASMA FROM SOLAR ACTIVE-REGION CORES: CONSTRAINTS FROM THE HINODE X-RAY TELESCOPE

    Energy Technology Data Exchange (ETDEWEB)

    Schmelz, J. T. [USRA, 7178 Columbia Gateway Drive, Columbia, MD 21046 (United States); Christian, G. M.; Matheny, P. O., E-mail: jschmelz@usra.edu [Physics Department, University of Memphis, Memphis, TN 38152 (United States)

    2016-12-20

    Mechanisms invoked to heat the solar corona to millions of degrees kelvin involve either magnetic waves or magnetic reconnections. Turbulence in the convection zone produces MHD waves, which travel upward and dissipate. Photospheric motions continuously build up magnetic energy, which is released through magnetic reconnection. In this paper, we concentrate on hot non-flaring plasma with temperatures of 5 MK <  T  < 10 MK because it is one of the few observables for which wave and reconnection models make different predictions. Wave models predict no (or little) hot plasma, whereas reconnection models predict it, although in amounts that are challenging to detect with current instrumentation. We used data from the X-ray Telescope (XRT) and the Atmospheric Imaging Assembly (AIA). We requested a special XRT observing sequence, which cycled through the thickest XRT filter several times per hour so we could average these images and improve the signal-to-noise. We did differential emission measure (DEM) analysis using the time-averaged thick-filter data as well as all available channels from both the XRT and AIA for regions observed on 2014 December 11. Whereas our earlier work was only able to determine that plasma with a temperature greater than 5 MK was present , we are now able to find a well-constrained DEM distribution. We have therefore added a strong observational constraint that must be explained by any viable coronal heating model. Comparing state-of-the-art wave and reconnection model predictions, we can conclude that reconnection is heating the hot plasma in these active regions.

  16. Interpreting Disruption Prediction Models to Improve Plasma Control

    Science.gov (United States)

    Parsons, Matthew

    2017-10-01

    In order for the tokamak to be a feasible design for a fusion reactor, it is necessary to minimize damage to the machine caused by plasma disruptions. Accurately predicting disruptions is a critical capability for triggering any mitigative actions, and a modest amount of attention has been given to efforts that employ machine learning techniques to make these predictions. By monitoring diagnostic signals during a discharge, such predictive models look for signs that the plasma is about to disrupt. Typically these predictive models are interpreted simply to give a `yes' or `no' response as to whether a disruption is approaching. However, it is possible to extract further information from these models to indicate which input signals are more strongly correlated with the plasma approaching a disruption. If highly accurate predictive models can be developed, this information could be used in plasma control schemes to make better decisions about disruption avoidance. This work was supported by a Grant from the 2016-2017 Fulbright U.S. Student Program, administered by the Franco-American Fulbright Commission in France.

  17. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  18. Chapter 8: Plasma operation and control [Progress in the ITER Physics Basis (PIPB)

    International Nuclear Information System (INIS)

    Gribov, Y.; Humphreys, D.; Kajiwara, K.; Lazarus, E.A.; Lister, J.B.; Ozeki, T.; Portone, A.; Shimada, M.; Sips, A.C.C.; Wesley, J.C.

    2007-01-01

    The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m -1 ), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation

  19. XSC plasma control: Tool development for the session leader

    International Nuclear Information System (INIS)

    Ambrosino, G.; Albanese, R.; Ariola, M.; Cenedese, A.; Crisanti, F.; Tommasi, G. De; Mattei, M.; Piccolo, F.; Pironti, A.; Sartori, F.; Villone, F.

    2005-01-01

    A new model-based shape controller (XSC, i.e., eXtreme Shape Controller) able to operate with high elongation and triangularity plasmas has been designed and implemented at JET in 2003. The use of the XSC needs a number of steps, which at present are not automated and therefore imply the involvement of several experts. To help the session leader in preparing an experiment, a number of software tools are needed. The paper describes the SW tools that are currently in the developing phase, and describes the new framework for the preparation of a JET experiment

  20. Real-Time Software for the Compass Tokamak Plasma Control

    Energy Technology Data Exchange (ETDEWEB)

    Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Prague (Costa Rica)

    2009-07-01

    This poster presents the flexible and high-performance real time system that guarantees the desired time cycles for plasma control on the COMPASS tokamak: 500 {mu}s for toroidal field, current, equilibrium and shaping; 50 {mu}s for fast control of the equilibrium and vertical instability. This system was developed on top of a high-performance processor and a software framework (MARTe) tailored for real-time. The preliminary measurements indicate that the time constraints will be met on the final solution. The system allows the making of modifications in the future to improve software components. (A.C.)

  1. Edge localized mode control by resonant magnetic perturbations in tokamak plasmas

    International Nuclear Information System (INIS)

    Orain, Francois

    2014-01-01

    The growth of plasma instabilities called Edge Localized Modes (ELMs) in tokamaks results in the quasi-periodic relaxation of the edge pressure profile. These relaxations induce large heat fluxes which might be harmful for the divertor in ITER, thus ELM control is mandatory in ITER. One of the promising control methods planned in ITER is the application of external resonant magnetic perturbations (RMPs), already efficient for ELM mitigation/suppression in current tokamak experiments. However a better understanding of the interaction between ELMs, RMPs and plasma flows is needed to explain the experimental results and make reliable predictions for ITER. In this perspective, non-linear modeling of ELMs and RMPs is done with the reduced MHD code JOREK, in toroidal geometry including the X-point and the Scrape-Off Layer. The initial model has been further developed to describe self-consistent plasma flows - with the addition of the bi-fluid diamagnetic drifts, the neoclassical friction and a source of parallel rotation - and to simulate the RMP penetration consistently with the plasma response. As a first step, the plasma response to RMPs (without ELMs) is studied for JET, MAST and ITER realistic plasma parameters and geometry. The general behaviour of the plasma/RMP interaction is similar for the three studied cases: RMPs are generally screened by the formation of response currents, induced by the plasma rotation on the resonant surfaces. RMPs however penetrate at the very edge where an ergodic zone is formed. The amplification of the non-resonant spectrum of the magnetic perturbations is also observed in the core. The edge ergodization induces an enhanced transport at the edge, which slightly degrades the pedestal profiles. RMPs also generate the 3D-deformation of the plasma boundary with a maximum deformation near the X-point where lobe structures are formed. Then the full dynamics of a multi-ELM cycle (without RMPs) is modeled for the first time in realistic

  2. XFEL resonant photo-pumping of dense plasmas and dynamic evolution of autoionizing core hole states

    OpenAIRE

    Rosmej, F. B.; Moinard, A.; Renner, O.; Galtier, E.; Lee, J. J.; Nagler, B.; Heimann, P. A.; Schlotter, W.; Turner, J. J.; Lee, R. W.; Makita, M.; Riley, D.; Seely, J.

    2016-01-01

    Similarly to the case of LIF (Laser-Induced Fluorescence), an equally revolutionary impact to science is expected from resonant X-ray photo-pumping. It will particularly contribute to a progress in high energy density science: pumped core hole states create X-ray transitions that can escape dense matter on a 10 fs-time scale without essential photoabsorption, thus providing a unique possibility to study matter under extreme conditions. In the first proof of principle experiment at the X-ray F...

  3. Status of the new WEST plasma control system

    International Nuclear Information System (INIS)

    Ravene, Nathalie; Nouailletas, Rémy; Signoret, Jacqueline; Guillerminet, Bernard; Treutterrer, Wolfgang; Spring, Anett; Masand, Harish; Dhongde, Jasraj; Bhandarkar, Manisha; Rapson, Chris; Laqua, Heike; Lewerentz, Marc; Moreau, Philippe; Brémond, Sylvain; Allegretti, Ludovic; Raupp, Gerhard; Werner, Andreas; Laurent, François Saint; Nardon, Eric

    2016-01-01

    The WEST (W – for Tungsten – Environment in Steady state Tokamak) project is aiming at minimizing technology and operational risks of a full tungsten actively cooled divertor on ITER. It was started in 2013 and consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. To operate the next coming operations of WEST, new controllers are required. These developments are an opportunity to develop a new Plasma Control System (PCS) architecture featuring build-in real time handling of both plasma and plants events, thus addressing key ITER needs. The Tore Supra PCS will be refurbished including a new Pulse Schedule Editor (PSE). The main idea is to use a time segmented approach to describe the pulse schedule with a full integration of event handling both on PCS and PSE. Further to detailed requirement specifications and architecture design, two software tools were selected to define and execute a whole plasma discharge defined as a set of time segments. The PCS real-time framework (RTF) is based on an upgraded version of the AUG framework, called DCS (Discharge Control System). The PSE is the Xedit application used on WEGA and under further development for W7-X facility. This paper reports on the status of the new WEST PCS developments. The on-going developments to adapt DCS to the Tore Supra Control infrastructure networks (new real-time network, chronology system and pulse supervision) will be reported. The required preparations for the use of Xedit will be presented, mainly the appropriate formal description of the WEST control system and the implementation of the mapping between the Xedit experiment configuration and DCS configuration files.

  4. Status of the new WEST plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Ravene, Nathalie, E-mail: nathalie.ravenel@gmail.com [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Nouailletas, Rémy; Signoret, Jacqueline; Guillerminet, Bernard [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Treutterrer, Wolfgang [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Spring, Anett [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Masand, Harish; Dhongde, Jasraj; Bhandarkar, Manisha [Institute for Plasma Research (IPR), Near Indira Bridge, Bhat, Gandhinagar, 382 428 Gujarat (India); Rapson, Chris [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Laqua, Heike; Lewerentz, Marc [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Moreau, Philippe; Brémond, Sylvain; Allegretti, Ludovic [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Raupp, Gerhard [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Werner, Andreas [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Laurent, François Saint; Nardon, Eric [IRFM, CEA, F-13108 Saint Paul lez Durance (France)

    2016-11-15

    The WEST (W – for Tungsten – Environment in Steady state Tokamak) project is aiming at minimizing technology and operational risks of a full tungsten actively cooled divertor on ITER. It was started in 2013 and consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. To operate the next coming operations of WEST, new controllers are required. These developments are an opportunity to develop a new Plasma Control System (PCS) architecture featuring build-in real time handling of both plasma and plants events, thus addressing key ITER needs. The Tore Supra PCS will be refurbished including a new Pulse Schedule Editor (PSE). The main idea is to use a time segmented approach to describe the pulse schedule with a full integration of event handling both on PCS and PSE. Further to detailed requirement specifications and architecture design, two software tools were selected to define and execute a whole plasma discharge defined as a set of time segments. The PCS real-time framework (RTF) is based on an upgraded version of the AUG framework, called DCS (Discharge Control System). The PSE is the Xedit application used on WEGA and under further development for W7-X facility. This paper reports on the status of the new WEST PCS developments. The on-going developments to adapt DCS to the Tore Supra Control infrastructure networks (new real-time network, chronology system and pulse supervision) will be reported. The required preparations for the use of Xedit will be presented, mainly the appropriate formal description of the WEST control system and the implementation of the mapping between the Xedit experiment configuration and DCS configuration files.

  5. Emotional learning based intelligent controller for a PWR nuclear reactor core during load following operation

    International Nuclear Information System (INIS)

    Khorramabadi, Sima Seidi; Boroushaki, Mehrdad; Lucas, Caro

    2008-01-01

    The design and evaluation of a novel approach to reactor core power control based on emotional learning is described. The controller includes a neuro-fuzzy system with power error and its derivative as inputs. A fuzzy critic evaluates the present situation, and provides the emotional signal (stress). The controller modifies its characteristics so that the critic's stress is reduced. Simulation results show that the controller has good convergence and performance robustness characteristics over a wide range of operational parameters

  6. Study of laser-imploded core plasmas with an advanced Kirkpatrick endash Baez x-ray microscope

    International Nuclear Information System (INIS)

    Kodama, R.; Shiraga, H.; Miyanaga, M.; Matsushita, T.; Nakai, M.; Azechi, H.; Mima, K.; Kato, Y.

    1997-01-01

    We developed an advanced Kirkpatrick endash Baez (AKB) x-ray microscope which consisted of two hyperbolic mirrors and two elliptic mirrors. The spatial resolution of approx-lt 3 μm was realized over ∼1 mm diam. This AKB microscope was used for x-ray imaging in laser fusion experiments. Laser absorption nonuniformity with a large wave number on a spherical solid target or a plane slab target was estimated by measurements of x-ray emission from the target surface with the microscope. The x-ray images of the imploded core plasmas were also obtained with the AKB microscope, changing the laser focus condition and the laser energy balance. copyright 1997 American Institute of Physics

  7. Controlling the Plasma-Polymerization Process of N-Vinyl-2-pyrrolidone

    DEFF Research Database (Denmark)

    Norrman, Kion; Winther-Jensen, Bjørn

    2005-01-01

    N-vinyl-2-pyrrolidone was plasma-polymerized on glass substrates using a pulsed AC plasma. Pulsed AC plasma produces a chemical surface structure different from that produced by conventional RF plasma; this is ascribed to the different power regimes used. A high degree of control over the structure...... of the chemical surface was obtained using pulsed AC plasma, as shown by ToF-SIMS. It is demonstrated how the experimental conditions to some extent control the chemical structure of the plasma-polymerized film, e.g., film thickness, density of post-plasma-polymerized oligomeric chains, and the density of intact...

  8. Radiation control in fusion plasmas by magnetic confinement

    International Nuclear Information System (INIS)

    Dachicourt, R.

    2012-10-01

    The present work addresses two important issues for the industrial use of fusion: plasma radiation control, as a part of the more general power handling issue, and high density tokamak operation. These two issues will be most critical in the demonstration reactor, called DEMO, intermediate step between ITER and a future commercial reactor. For DEMO, the need to radiate a large fraction of the power so as to limit the peak power load on the divertor will be a key constraint. High confinement will have to be combined with high radiated power fraction, and the required level of plasma purity. The main achievement of this thesis is to have shown experimental evidence of the existence of a stable plasma regime meeting the most critical requirements of a DEMO scenario: an electron density up to 40% above the Greenwald value, together with a fraction of radiated power close to 80%, with a good energy confinement and limited dilution. The plasma is additionally heated with ion cyclotron waves in a central electron heating scenario, featuring alpha particle heating. The original observations reported in this work bring highly valuable new pieces of information both to the physics of the tokamak edge layer and to the construction of an 'integrated operational scenario' required to successfully operate fusion devices. In the way for getting high density plasmas, the new observations involve the following topics. First, the formation of a poloidal asymmetry in the edge electron density profile, with a maximum density located close to toroidal pumped limiter. This asymmetry occurs inside the separatrix, with a constant plasma pressure on magnetic surfaces. Secondly, a correlative decrease of the electron temperature in the same edge region. Thirdly, the excellent coupling capabilities of the ICRH waves, up to a central line averaged electron density of 1.4 times the Greenwald density. Fourthly, a poloidally asymmetric edge radiation region, providing the dissipation of 80% of

  9. 28. Zvenigorod conference on the plasma physics and controlled thermonuclear synthesis. Theses of reports

    International Nuclear Information System (INIS)

    2001-01-01

    Theses of reports, presented at the 28th Conference on the plasma physics and controlled thermonuclear synthesis (Zvenigorod, 19-23 February 2001) are published. 246 reports were heard at the following sections: magnetic confinement, theory and experiments; inertial thermonuclear synthesis; plasma processes and physics of gas-discharge plasma; physical bases of plasma technologies. 17 reports had the summarizing character [ru

  10. Active control of magneto-hydrodynamic instabilities in hot plasmas

    CERN Document Server

    2015-01-01

    During the past century, world-wide energy consumption has risen dramatically, which leads to a quest for new energy sources. Fusion of hydrogen atoms in hot plasmas is an attractive approach to solve the energy problem, with abundant fuel, inherent safety and no long-lived radioactivity.  However, one of the limits on plasma performance is due to the various classes of magneto-hydrodynamic instabilities that may occur. The physics and control of these instabilities in modern magnetic confinement fusion devices is the subject of this book. Written by foremost experts, the contributions will provide valuable reference and up-to-date research reviews for "old hands" and newcomers alike.

  11. On the maximum Q in feedback controlled subignited plasmas

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1990-01-01

    High Q operation in feedback controlled subignited fusion plasma requires the operating temperature to be close to the ignition temperature. In the present work we discuss technological and physical effects which may restrict this temperature difference. The investigation is based on a simplified, but still accurate, 0=D analytical analysis of the maximum Q of a subignited system. Particular emphasis is given to sawtooth ocsillations which complicate the interpretation of diagnostic neutron emission data into plasma temperatures and may imply an inherent lower bound on the temperature deviation from the ignition point. The estimated maximum Q is found to be marginal (Q = 10-20) from the point of view of a fusion reactor. (authors)

  12. Formation and control of plasma potentials in TMX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T.C.; Orzechowski, T.J.; Porkolab, M.; Stallard, B.W.

    1981-05-06

    The methods to be employed to form and control plasma potentials in the TMX Upgrade tandem mirror with thermal barriers are described. ECRH-generated mirror -confined electron plasmas are used to establish a negative potential region to isolate the end-plug and central-cell celectrons. This thermal isolation will allow a higher end-plug electron temperature and an increased central-cell confining potential. Improved axial central-cell ion confinement results since higher temperature central-cell ions can be confined. This paper describes: (1) calculations of the sensitivity of barrier formation to vacuum conditions and to the presence of impurities in the neutral beams, (2) calculations of microwave penetration and absorption used to design the ECRH system, and (3) techniques to limit electron runaway to high energies by localized microwave beams and by relativistic detuning.

  13. Controlling Plasma Channels through Ultrashort Laser Pulse Filamentation

    Science.gov (United States)

    Ionin, Andrey; Seleznev, Leonid; Sunchugasheva, Elena

    2013-09-01

    A review of studies fulfilled at the Lebedev Institute in collaboration with the Moscow State University and Institute of Atmospheric Optics in Tomsk on influence of various characteristics of ultrashort laser pulse on plasma channels formed under its filamentation is presented. Filamentation of high-power laser pulses with wavefront controlled by a deformable mirror, with cross-sections spatially formed by various diaphragms and with different wavelengths was experimentally and numerically studied. An application of plasma channels formed due to filamentation of ultrashort laser pulse including a train of such pulses for triggering and guiding long electric discharges is discussed. The research was supported by RFBR Grants 11-02-12061-ofi-m and 11-02-01100, and EOARD Grant 097007 through ISTC Project 4073 P

  14. Formation and control of plasma potentials in TMX upgrade

    International Nuclear Information System (INIS)

    Simonen, T.C.; Orzechowski, T.J.; Porkolab, M.; Stallard, B.W.

    1981-01-01

    The methods to be employed to form and control plasma potentials in the TMX Upgrade tandem mirror with thermal barriers are described. ECRH-generated mirror -confined electron plasmas are used to establish a negative potential region to isolate the end-plug and central-cell celectrons. This thermal isolation will allow a higher end-plug electron temperature and an increased central-cell confining potential. Improved axial central-cell ion confinement results since higher temperature central-cell ions can be confined. This paper describes: (1) calculations of the sensitivity of barrier formation to vacuum conditions and to the presence of impurities in the neutral beams, (2) calculations of microwave penetration and absorption used to design the ECRH system, and (3) techniques to limit electron runaway to high energies by localized microwave beams and by relativistic detuning

  15. Separation control of NACA0015 airfoil using plasma actuators

    Science.gov (United States)

    Harada, Daisuke; Sakakibara, Jun

    2017-11-01

    Separation control of NACA0015 airfoil by means of plasma actuators was investigated. Plasma actuators in spanwise intermittent layout on the suction surface of the airfoil were activated with spanwise phase difference φ = 0 or φ = π in the case of dimensionless burst frequencyF+ = 6 and F+ = 0.5 at Re = 6.3 ×104 . The lift and drag of the airfoil were measured using a two component force balance. The flow around the airfoil was measured by PIV analysis. In the condition of F+ = 6 and φ = π at around stall angle, which is 10 degrees, the lift-to-drag ratio was higher than that ofF+ = 6 and φ = 0 . Therefore, it was confirmed that aerodynamic characteristics of the airfoil improved by disturbances with temporal and spatial phase difference.

  16. Plasma density remote control system of experimental advanced superconductive tokamak

    International Nuclear Information System (INIS)

    Zhang Mingxin; Luo Jiarong; Li Guiming; Wang Hua; Zhao Dazheng; Xu Congdong

    2007-01-01

    In Tokamak experiments, experimental data and information on the density control are stored in the local computer system. Therefore, the researchers have to be in the control room for getting the data. Plasma Density Remote Control System (DRCS), which is implemented by encapsulating the business logic on the client in the B/S module, conducts the complicated science computation and realizes the synchronization with the experimental process on the client. At the same time, Web Services and Data File Services are deployed for the data exchange. It is proved in the experiments that DRCS not only meets the requirements for the remote control, but also shows an enhanced capability on the data transmission. (authors)

  17. Plasma position and shape control device for thermonuclear device

    International Nuclear Information System (INIS)

    Takeuchi, Kazuhiro; Abe, Mitsushi; Kinoshita, Shigemi.

    1993-01-01

    A plasma position and shape control system is constituted with a measuring device, a quenching probability calculation section and a control calculation section. A quenching probability is calculated in the quenching probability calculation section by using a measuring data on temperature, electric current and magnetic field of superconductive coils, based on a margin upto a limit value. The control calculation section selects a control method which decreases applied voltage or current instruction value as the quenching probability of the coils is higher. Since the quenching probability of the superconductive coils can be forecast and a state of low quenching danger can be selected, the safety of the device is improved. When the quenching danger is allowed to a predetermined value, a wide operation region can be provided. (N.H.)

  18. Hierarchical Mesoporous Organosilica-Silica Core-Shell Nanoparticles Capable of Controlled Fungicide Release.

    Science.gov (United States)

    Luo, Leilei; Liang, Yucang; Erichsen, Egil Severin; Anwander, Reiner

    2018-05-17

    A new class of hierarchically structured mesoporous silica core-shell nanoparticles (HSMSCSNs) with a periodic mesoporous organosilica (PMO) core and a mesoporous silica (MS) shell is reported. The applied one-pot, two-step strategy allows rational control over the core/shell chemical composition, topology, and pore/particle size, simply by adjusting the reaction conditions in the presence of cetyltrimethylammonium bromide (CTAB) as structure-directing agent under basic conditions. The spherical, ethylene- or methylene-bridged PMO cores feature hexagonal (p6mm) or cage-like cubic symmetry (Pm3‾ n) depending on the organosilica precursor. The hexagonal MS shell was obtained by n-hexane-induced controlled hydrolysis of TEOS followed by directional co-assembly/condensation of silicate/CTAB composites at the PMO cores. The HSMSCSNs feature a hierarchical pore structure with pore diameters of about 2.7 and 5.6 nm in the core and shell domains, respectively. The core sizes and shell thicknesses are adjustable in the ranges of 90-275 and 15-50 nm, respectively, and the surface areas (max. 1300 m 2  g -1 ) and pore volumes (max. 1.83 cm 3  g -1 ) are among the highest reported for core-shell nanoparticles. The adsorption and controlled release of the fungicide propiconazole by the HSMSCSNs showed a three-stage release profile. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  19. Plasma core electron density and temperature measurements using CVI line emissions in TCABR Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, F. do, E-mail: fellypen@ifi.unicamp.br [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Centro de Componentes Semicondutores; Machida, M. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Instituto de Fisica Gleb Wataghin; Severo, J.H.F.; Sanada, E.; Ronchi, G. [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica

    2015-08-15

    In this work, we present results of electron temperature (T{sub e} ) and density (n {sub e} ) measurements obtained in Tokamak Chauffage Alfven Bresilien (TCABR) tokamak using visible spectroscopy from CVI line emissions which occurs mainly near the center of the plasma column. The presented method is based on a well-known relationship between the particle flux (Γ {sub ion}) and the photon flux (ø {sub ion}) emitted by an ion species combined with ionizations per photon atomic data provided by the atomic data and analysis structure (ADAS) database. In the experiment, we measured the photon fluxes of three different CVI spectral line emissions, 4685.2, 5290.5, and 6200.6 Å (one line per shot). Using this method it was possible to find out the temporal evolution of T{sub e} and n{sub e} in the plasma. The results achieved are in good agreement with T{sub e} and n{sub e} measurements made using other diagnostic tools. (author)

  20. Plasma Physics and Controlled Nuclear Fusion Research. Vol. II. Proceedings of a Conference on Plasma Physics and Controlled Physics Research

    International Nuclear Information System (INIS)

    1966-01-01

    Research on controlled nuclear fusion was first disclosed at the Second United Nations Conference on the Peaceful Uses of Atomic Energy, held at Geneva in 1958. From the information given, it was evident that a better understanding of the behaviour of hot dense plasmas was needed before the goal of economic energy release from nuclear fusion could be reached. The fact that research since then has been most complex and costly has enhanced the desirability of international co-operation and exchange of information and experience. Having organized its First Conference on Plasma Physics and Controlled Nuclear Fusion Research at Salzburg in 1961, the International Atomic Energy Agency again provided the means for such cooperation in organizing its Second Conference on this subject on 6-10 September, 1965, at Culham, Abingdon, Berks, England. The meeting was arranged with the generous help of the United Kingdom Atomic Energy Authority at their Culham Laboratory, where the facilities and assistance of the staff were greatly appreciated. At the meeting, which was attended by 268 participants from 26 member states and three international organizations, significant results from many experiments, including those from the new and larger machines, became available. It has now become feasible to intercorrelate data obtained from a number of similar machines; this has led to a more complete understanding of plasma behaviour. No breakthrough was reported nor had been expected towards the economical release of the energy from fusion, but there was increased understanding of the problems of production, control and containment of high-density and high-temperature plasmas

  1. A note on supersonic flow control with nanosecond plasma actuator

    Science.gov (United States)

    Zheng, J. G.; Cui, Y. D.; Li, J.; Khoo, B. C.

    2018-04-01

    A concept study on supersonic flow control using nanosecond pulsed plasma actuator is conducted by means of numerical simulation. The nanosecond plasma discharge is characterized by the generation of a micro-shock wave in ambient air and a residual heat in the discharge volume arising from the rapid heating of near-surface gas by the quick discharge. The residual heat has been found to be essential for the flow separation control over aerodynamic bodies like airfoil and backward-facing step. In this study, novel experiment is designed to utilize the other flow feature from discharge, i.e., instant shock wave, to control supersonic flow through shock-shock interaction. Both bow shock in front of a blunt body and attached shock anchored at the tip of supersonic projectile are manipulated via the discharged-induced shock wave in an appropriate manner. It is observed that drag on the blunt body is reduced appreciably. Meanwhile, a lateral force on sharp-edged projectile is produced, which can steer the body and give it an effective angle of attack. This opens a promising possibility for extending the applicability of this flow control technique in supersonic flow regime.

  2. INFERENCE OF HEATING PROPERTIES FROM “HOT” NON-FLARING PLASMAS IN ACTIVE REGION CORES. II. NANOFLARE TRAINS

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, W. T.; Bradshaw, S. J. [Department of Physics and Astronomy, Rice University, Houston, TX 77251-1892 (United States); Cargill, P. J., E-mail: will.t.barnes@rice.edu, E-mail: stephen.bradshaw@rice.edu, E-mail: p.cargill@imperial.ac.uk [Space and Atmospheric Physics, The Blackett Laboratory, Imperial College, London SW7 2BW (United Kingdom)

    2016-12-20

    Despite its prediction over two decades ago, the detection of faint, high-temperature (“hot”) emission due to nanoflare heating in non-flaring active region cores has proved challenging. Using an efficient two-fluid hydrodynamic model, this paper investigates the properties of the emission expected from repeating nanoflares (a nanoflare train) of varying frequency as well as the separate heating of electrons and ions. If the emission measure distribution (EM(T)) peaks at T = T{sub m} , we find that EM(T{sub m}) is independent of details of the nanoflare train, and EM(T) above and below T{sub m} reflects different aspects of the heating. Below T{sub m} , the main influence is the relationship of the waiting time between successive nanoflares to the nanoflare energy. Above T{sub m}, power-law nanoflare distributions lead to an extensive plasma population not present in a mono-energetic train. Furthermore, in some cases, characteristic features are present in EM(T). Such details may be detectable given adequate spectral resolution and a good knowledge of the relevant atomic physics. In the absence of such resolution we propose some metrics that can be used to infer the presence of “hot” plasma.

  3. Multi-core System Architecture for Safety-critical Control Applications

    DEFF Research Database (Denmark)

    Li, Gang

    and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...

  4. Identifying Core Competencies of Infection Control Nurse Specialists in Hong Kong.

    Science.gov (United States)

    Chan, Wai Fong; Bond, Trevor G; Adamson, Bob; Chow, Meyrick

    2016-01-01

    To confirm a core competency scale for Hong Kong infection control nurses at the advanced nursing practice level from the core competency items proposed in a previous phase of this study. This would serve as the foundation of competency assurance in Hong Kong hospitals. A cross-sectional survey design was used. All public and private hospitals in Hong Kong. All infection control nurses in hospitals of Hong Kong. The 83-item proposed core competency list established in an earlier study was transformed into a questionnaire and sent to 112 infection control nurses in 48 hospitals in Hong Kong. They were asked to rate the importance of each infection prevention and control item using Likert-style response categories. Data were analyzed using the Rasch model. The response rate of 81.25% was achieved. Seven items were removed from the proposed core competency list, leaving a scale of 76 items that fit the measurement requirements of the unidimensional Rasch model. Essential core competency items of advanced practice for infection control nurses in Hong Kong were identified based on the measurement criteria of the Rasch model. Several items of the scale that reflect local Hong Kong contextual characteristics are distinguished from the overseas standards. This local-specific competency list could serve as the foundation for education and for certification of infection control nurse specialists in Hong Kong. Rasch measurement is an appropriate analytical tool for identifying core competencies of advanced practice nurses in other specialties and in other locations in a manner that incorporates practitioner judgment and expertise.

  5. Creation and control of variably shaped plasmas in TCV

    International Nuclear Information System (INIS)

    Hofmann, F.; Lister, J.B.; Anton, M.

    1994-01-01

    During the first year of operation, the TCV tokamak has produced a large variety of plasma shapes and magnetic configurations, with 1.0≤B tor ≤1.46T, I p ≤800kA, k≤2.05, -0.7≤δ ≤0.7. A new shape control algorithm, based on a finite element reconstruction of the plasma current in real time, has been implemented. Vertical growth rates of 800 sec -1 , corresponding to a stability margin f=1.15, have been stabilized. Ohmic H-modes, with energy confinement times reaching 80ms, normalized beta (β tor aB/I p ) of 1.9 and τ E /ITER89-P of 2.4 have been obtained in single-null X-point deuterium discharges with the ion grad B drift towards the X-point. Limiter H-modes with maximum line averaged electron densities of 1.7x10 20 m -3 have been observed in D-shaped plasmas with 360kA≤I p ≤600kA. (Author)

  6. EURATOM-CEA association contributions to the 26. EPS conference on controlled fusion and plasma physics, Maastricht

    International Nuclear Information System (INIS)

    1999-10-01

    This report references the EURATOM-CEA association contributions presented at the 26. EPS conference on controlled fusion and plasma physics, in Maastricht (Netherlands) the 14-18 June 1999. Two invited papers and 24 contributed papers are proposed. They deal with: tokamak devices; particle recirculation in ergodic divertor; current profile control and MHD stability in Tore Supra discharges; edge-plasma control by the ergodic divertor; electron heat transport in stochastic magnetic layer; bolometry and radiated power; particle collection by ergodic divertor; study and simulation of plasma impurities; line shape modelling for plasma edge conditions; dynamical study of the radial structure of the fluctuations measured by reciprocating Langmuir probe in Tore Supra; up-down asymmetry of density fluctuations; Halo currents in a circular tokamak; real time measurement of the position, density, profile and current profile at Tore Supra; poloidal rotation measurement by reflectometry; interpretation of q-profile dependence of the LH power deposition profile during LHCD experiments; ICFR plasma production and optimization; improved core electron confinement; measurement of hard X-ray emission profile; modelling of shear effects on thermal and particles transport; ion turbulence; current drive generation based on autoresonance and intermittent trapping mechanisms. (A.L.B.)

  7. Measurement of core plasma temperature and rotation on W7-X made available by the x-ray imaging crystal spectrometer (XICS)

    Energy Technology Data Exchange (ETDEWEB)

    Pablant, N A; Bitter, M; Burhenn, R; Delgado-Aparicio, L; Ellis, R; Gates, D; Goto, M; Hill, K W; Langenberg, A; Lazerson, S; Mardenfeld, M; Morita, S; Neilson, G H; Oishi, T; Pedersen, T S

    2014-07-01

    A new x-ray imaging crystal spectrometer diagnostic (XICS) is currently being built for installation on W7-X. This diagnostic will contribute to the study of ion and electron thermal transport and the evolution of the radial electric field by providing high resolution temperature and rotation measurements under many plasma conditions, including ECH heated plasmas. Installation is expected before the first experimental campaign (OP1.1), making an important set of measurements available for the first W7-X plasmas. This diagnostic will also work in concert with the HR-XCS diagnostic to provide an excellent diagnostic set for core impurity transport on W7-X.

  8. To the problem of electron temperature control in plasma

    International Nuclear Information System (INIS)

    Galechyan, G.A.; Anna, P.R.

    1995-01-01

    One of the main problems in low temperature plasma is control plasma parameters at fixed values of current and gas pressure in the discharge. It is known that an increase in the intensity of sound wave directed along the positive column to values in excess of a definite threshold leads to essential rise of the temperature of electrons. However, no less important is the reduction of electron temperature in the discharge down to the value less than that in plasma in the absence external influence. It is known that to reduce the electron temperature in the plasma of CO 2 laser, easily ionizable admixture are usually introduced in the discharge area with the view of increasing the overpopulation. In the present work we shall show that the value of electron temperature can be reduced by varying of sound wave intensity at its lower values. The experiment was performed on an experimental setup consisted of the tube with length 52 cm and diameter 9.8 cm, two electrodes placed at the distance of 27 cm from each other. An electrodynamical radiator of sound wave was fastened to one of tube ends. Fastened to the flange at the opposite end was a microphone for the control of sound wave parameters. The studies were performed in range of pressures from 40 to 180 Torr and discharge currents from 40 to 110 mA. The intensity of sound wave was varied from 74 to 92 dB. The measurement made at the first resonance frequency f = 150 Hz of sound in the discharge tube, at which a quarter of wave length keep within the length of the tube. The measurement of longitudinal electric field voltage in plasma of positive column was conducted with the help of two probes according to the compensation method. Besides, the measurement of gas temperature in the discharge were taken. Two thermocouple sensors were arranged at the distance of 8 cm from the anode, one of them being installed on the discharge tube axis, the second-fixed the tube wall

  9. To the problem of electron temperature control in plasma

    Energy Technology Data Exchange (ETDEWEB)

    Galechyan, G.A. [Institute of Applied Problem of Physics, Yerevan (Armenia); Anna, P.R. [Raritan Valley Community College, Somerville, NJ (United States)

    1995-12-31

    One of the main problems in low temperature plasma is control plasma parameters at fixed values of current and gas pressure in the discharge. It is known that an increase in the intensity of sound wave directed along the positive column to values in excess of a definite threshold leads to essential rise of the temperature of electrons. However, no less important is the reduction of electron temperature in the discharge down to the value less than that in plasma in the absence external influence. It is known that to reduce the electron temperature in the plasma of CO{sub 2} laser, easily ionizable admixture are usually introduced in the discharge area with the view of increasing the overpopulation. In the present work we shall show that the value of electron temperature can be reduced by varying of sound wave intensity at its lower values. The experiment was performed on an experimental setup consisted of the tube with length 52 cm and diameter 9.8 cm, two electrodes placed at the distance of 27 cm from each other. An electrodynamical radiator of sound wave was fastened to one of tube ends. Fastened to the flange at the opposite end was a microphone for the control of sound wave parameters. The studies were performed in range of pressures from 40 to 180 Torr and discharge currents from 40 to 110 mA. The intensity of sound wave was varied from 74 to 92 dB. The measurement made at the first resonance frequency f = 150 Hz of sound in the discharge tube, at which a quarter of wave length keep within the length of the tube. The measurement of longitudinal electric field voltage in plasma of positive column was conducted with the help of two probes according to the compensation method. Besides, the measurement of gas temperature in the discharge were taken. Two thermocouple sensors were arranged at the distance of 8 cm from the anode, one of them being installed on the discharge tube axis, the second-fixed the tube wall.

  10. Improving plasma shaping accuracy through consolidation of control model maintenance, diagnostic calibration, and hardware change control

    International Nuclear Information System (INIS)

    Baggest, D.S.; Rothweil, D.A.; Pang, S.

    1995-12-01

    With the advent of more sophisticated techniques for control of tokamak plasmas comes the requirement for increasingly more accurate models of plasma processes and tokamak systems. Development of accurate models for DIII-D power systems, vessel, and poloidal coils is already complete, while work continues in development of general plasma response modeling techniques. Increased accuracy in estimates of parameters to be controlled is also required. It is important to ensure that errors in supporting systems such as diagnostic and command circuits do not limit the accuracy of plasma parameter estimates or inhibit the ability to derive accurate plasma/tokamak system models. To address this issue, we have developed more formal power systems change control and power system/magnetic diagnostics calibration procedures. This paper discusses our approach to consolidating the tasks in these closely related areas. This includes, for example, defining criteria for when diagnostics should be re-calibrated along with required calibration tolerances, and implementing methods for tracking power systems hardware modifications and the resultant changes to control models

  11. Simulation of slag control for the Plasma Hearth Project

    International Nuclear Information System (INIS)

    Power, M.A.; Carney, K.P.; Peters. G.G.

    1996-01-01

    The goal of the Plasma Hearth Project is to stabilize alpha-emitting radionuclides in a vitreous slag and to reduce the effective storage volume of actinide-containing waste for long-term burial. The actinides have been shown to partition into the vitreous slag phase of the melt. The slag composition may be changed by adding glass-former elements to ensure that this removable slag has the most desired physical and chemical properties for long-term burial. A data acquisition and control system has been designed to regulate the composition of five elements in the slag

  12. Power supplies for plasma column control in COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Havlíček, Josef; Hauptmann, R.; Peroutka, Oldřich; Tadros, Momtaz; Hron, Martin; Janky, Filip; Vondráček, Petr; Cahyna, Pavel; Mikulín, Ondřej; Šesták, David; Junek, Pavel; Pánek, Radomír

    2013-01-01

    Roč. 88, 9-10 (2013), s. 1640-1645 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * Power supplies * Feedback control * Vertical displacement * Vertical kicks Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001543#

  13. Alternatives for contaminant control during MFTF plasma buildup

    International Nuclear Information System (INIS)

    Khan, J.M.; Valby, L.E.

    1979-01-01

    The MFTF mirror device considers all low-energy species to be contaminants, since their primary effect is to erode the plasma boundary by charge-exchange reactions. Confinement for other than hydrogen isotypes is far from complete and confinement time is hardly more than transit time from the source to the end wall. The brevity of the confinement time makes it all the more necessary to prevent any contamination which might further reduce it. At Livermore, the historical solution to contaminant control has been to evaporate titanium onto cold surfaces. An alternative to this approach and its implications are considered

  14. Possibility evaluation of eliminating the saturated control fuel element from Tehran research reactor core

    International Nuclear Information System (INIS)

    Mirvakili, S.M.; Keyvani, M.; Arshi, S. Safaei; Khalafi, H.

    2012-01-01

    Highlights: ► We show safe operation of Tehran research reactor without one of its control rods. ► We propose an optimum new core configuration by fuel management calculations. ► We calculate neutronic and thermal hydraulic parameters of the new core. ► Parameters are consistent with the safety criteria. - Abstract: In this study the possibility of safe operation of Tehran research reactor (TRR) providing the elimination of one control rod is evaluated. One of the control fuel elements (CFEs) of TRR has been reached the maximum permissible burn-up and due to the impossibility of fresh fuel assembly provision under current situation, providing an optimum core configuration which satisfies safe operation conditions by applying fuel management calculations is essential. In order to ensure the safe and stable operation of recently proposed configuration for TRR core, neutronic and thermal hydraulic parameters of the new core are calculated and compared with the safety criteria. The results show good compatibility with reactor safety criteria, and provide desired shutdown margin and safety reactivity factor.

  15. Control of plasma layer in a fusion reactor correlated to DC motor control using PSO-ANFIS

    International Nuclear Information System (INIS)

    Mahapatra, Sakuntala; Daniel, Raju; Dey, Deep Narayan

    2013-01-01

    Plasma position and shape control is very crucial for the overall performance of the fusion reactor such as Tokamak. The quality of the discharge in the Saskatchewan TORus-Modified (STOR-M) tokamak is strongly related to the position of the plasma column within the discharge vessel. If the plasma column approaches too near the wall, then either minor or complete disruption occurs. Consequently it is necessary to be able to control dynamically the position of the plasma column throughout the entire discharge. Now a day's most fusion reactor employs the traditional PID controller for the confinement of plasma layer. Fuzzy logic is used for the control of Plasma layer. In this paper we have used the hybrid of PSO-ANFIS technique to control the speed of a DC motor. We have used two input parameters like speed, torque and output is firing angle. In our work first order Sugeno fuzzy model is taken with three rules and the parameters of Gaussian membership function is controlled by the PSO technique. PSO-ANFIS speed controller obtains better dynamic behavior and superior performance of the DC motor speed control. Similar approach can be correlated to the control of plasma layer. For the plasma control two inputs can be taken as plasma position ΔH and the plasma current and the single output, the control decision u(t). (author)

  16. Feedback control of chlorine inductively coupled plasma etch processing

    International Nuclear Information System (INIS)

    Lin Chaung; Leou, K.-C.; Shiao, K.-M.

    2005-01-01

    Feedback control has been applied to poly-Si etch processing using a chlorine inductively coupled plasma. Since the positive ion flux and ion energy incident upon the wafer surface are the key factors that influence the etch rate, the ion current and the root mean square (rms) rf voltage on the wafer stage, which are measured using an impedance meter connected to the wafer stage, are adopted as the controlled variables to enhance etch rate. The actuators are two 13.56 MHz rf power generators, which adjust ion density and ion energy, respectively. The results of closed-loop control show that the advantages of feedback control can be achieved. For example, with feedback control, etch rate variation under the transient chamber wall condition is reduced roughly by a factor of 2 as compared to the open-loop case. In addition, the capability of the disturbance rejection was also investigated. For a gas pressure variation of 20%, the largest etch rate variation is about 2.4% with closed-loop control as compared with as large as about 6% variation using open-loop control. Also the effect of ion current and rms rf voltage on etch rate was studied using 2 2 factorial design whose results were used to derive a model equation. The obtained formula was used to adjust the set point of ion current and rf voltage so that the desired etch rate was obtained

  17. Time course of the acute effects of core stabilisation exercise on seated postural control.

    Science.gov (United States)

    Lee, Jordan B; Brown, Stephen H M

    2017-09-20

    Core stabilisation exercises are often promoted for purposes ranging from general fitness to high-performance athletics, and the prevention and rehabilitation of back troubles. These exercises, when performed properly, may have the potential to enhance torso postural awareness and control, yet the potential for achieving immediate gains has not been completely studied. Fourteen healthy young participants performed a single bout of non-fatiguing core stabilisation exercise that consisted of repeated sets of 2 isometric exercises, the side bridge and the four-point contralateral arm-and-leg extension. Seated postural control, using an unstable balance platform on top of a force plate, was assessed before and after exercise, including multiple time points within a 20-minute follow-up period. Nine standard postural control variables were calculated at each time point, including sway displacement and velocity ranges, root mean squares and cumulative path length. Statistical analysis showed that none of the postural control variables were significantly different at any time point following completion of core stabilisation exercise. Thus, we conclude that a single bout of acute core stabilisation exercise is insufficient to immediately improve seated trunk postural control in young healthy individuals.

  18. A Key to Improved Ion Core Confinement in the JET Tokamak: Ion Stiffness Mitigation due to Combined Plasma Rotation and Low Magnetic Shear

    DEFF Research Database (Denmark)

    Mantica, P.; Challis, C.; Peeters, A.G.

    2011-01-01

    New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. Phys. Rev. Lett. 102 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implicatio...

  19. A key to improved ion core confinement in the JET tokamak : ion stiffness mitigation due to combined plasma rotation and low magnetic Shear

    NARCIS (Netherlands)

    Mantica, P.; Angioni, C.; Challis, C.; Colyer, G.; Frassinetti, L.; Hawkes, N.C.; Johnson, T.; Tsalas, M.; de Vries, P.C.; Weiland, J.; Baiocchi, B.; Beurskens, M.N.A.; Figueiredo, A.C.A.; Giroud, C.; Hobirk, J.; Joffrin, E.; Lerche, E.; Naulin, V.; Peeters, A.G.; Salmi, A.; Sozzi, C.; Strintzi, D.; Staebler, G.; Tala, T.; Van Eester, D.; Versloot, T.W.

    2011-01-01

    New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. Phys. Rev. Lett. 102 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications

  20. Model of divertor biasing and control of scrape-off layer and divertor plasmas

    International Nuclear Information System (INIS)

    Nagasaki, K.; Itoh, K.; Itoh, S.

    1991-02-01

    Analytic model of the divertor biasing is described. For the given plasma and energy sources from the core plasma, the heat and particle flux densities on the divertor plate as well as scrape-off-layer (SOL)/divertor plasmas are analyzed in a slab model. Using a two-dimensional model, the effects of the divertor biasing and SOL current are studied. The conditions to balance the plasma temperature or sheath potential on different divertor plates are obtained. Effect of the SOL current on the heat channel width is also discussed. (author)

  1. Invited and contributed papers presented at the 22. EPS conference on controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    In this report one invited and fifteen contributed papers by researchers of the `Centre de Recherche en Physique des Plasmas`, Lausanne, to the 22. EPS Conference on Controlled Fusion and Plasma Physics are assembled. figs., tabs., refs.

  2. Power supply controlled for plasma torch generation; Fuente de alimentacion controlada para la generacion de un plasma

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Z, S

    1997-12-31

    The high density of energy furnished by thermal plasma is profited in a wide range of applications, such as those related with welding fusion, spray coating and at the present in waste destruction. The waste destruction by plasma is a very attractive process because the remaining products are formed by inert glassy grains and non-toxic gases. The main characteristics of thermal plasmas are presented in this work. Techniques based on power electronics are utilized to achieve a good performance in thermal plasma generation. This work shown the design and construction of three phase control system for electric supply of thermal plasma torch, with 250 kw of capacity, as a part of the project named `Destruction of hazard wastes by thermal plasma` actually working in the Instituto Nacional de Investigaciones Nucleares (ININ). The characteristics of thermal plasma and its generation are treated in the first chapter. The A C controllers by thyristors applied in three phase arrays are described in the chapter II, talking into account the power transformer, rectifiers bank and aliasing coil. The chapter III is dedicated in the design of the trigger module which controls the plasma current by varying the trigger angle of the SCR`s; the protection and isolating unit are also presented in this chapter. The results and conclusions are discussed in chapter IV. (Author).

  3. Determination of the plasma position for its real-time control in the COMPASS tokamak

    International Nuclear Information System (INIS)

    Janky, F.; Havlicek, J.; Valcarcel, D.; Hron, M.; Horacek, J.; Kudlacek, O.; Panek, R.; Carvalho, B.B.

    2011-01-01

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  4. Determination of the plasma position for its real-time control in the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    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 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Valcarcel, D. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal); Hron, M.; Horacek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Kudlacek, O. [Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Technicka 2, 166 27 Prague (Czech Republic); Panek, R. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Carvalho, B.B. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal)

    2011-10-15

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  5. Plasma physics and controlled fusion research during half a century

    Energy Technology Data Exchange (ETDEWEB)

    Lehnert, Bo

    2001-06-01

    A review is given on the historical development of research on plasma physics and controlled fusion. The potentialities are outlined for fusion of light atomic nuclei, with respect to the available energy resources and the environmental properties. Various approaches in the research on controlled fusion are further described, as well as the present state of investigation and future perspectives, being based on the use of a hot plasma in a fusion reactor. Special reference is given to the part of this work which has been conducted in Sweden, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. Temperatures above the limit for ignition of self-sustained fusion reactions, i.e. at more than hundred million degrees, have been reached in large experiments and under conditions where the fusion power generation is comparable to the power losses. An energy producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient when being based on the present state of art. Future international research has therefore to be conducted along broad lines, with necessary ingredients of basic investigations and new ideas.

  6. Plasma physics and controlled fusion research during half a century

    International Nuclear Information System (INIS)

    Lehnert, Bo

    2001-06-01

    A review is given on the historical development of research on plasma physics and controlled fusion. The potentialities are outlined for fusion of light atomic nuclei, with respect to the available energy resources and the environmental properties. Various approaches in the research on controlled fusion are further described, as well as the present state of investigation and future perspectives, being based on the use of a hot plasma in a fusion reactor. Special reference is given to the part of this work which has been conducted in Sweden, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. Temperatures above the limit for ignition of self-sustained fusion reactions, i.e. at more than hundred million degrees, have been reached in large experiments and under conditions where the fusion power generation is comparable to the power losses. An energy producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient when being based on the present state of art. Future international research has therefore to be conducted along broad lines, with necessary ingredients of basic investigations and new ideas

  7. Optimal power and distribution control for weakly-coupled-core reactor

    International Nuclear Information System (INIS)

    Oohori, Takahumi; Kaji, Ikuo

    1977-01-01

    A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

  8. Torque decomposition and control in an iron core linear permanent magnet motor.

    NARCIS (Netherlands)

    Overboom, T.T.; Smeets, J.P.C.; Stassen, J.M.; Jansen, J.W.; Lomonova, E.

    2012-01-01

    Abstract—This paper concerns the decomposition and control of the torque produced by an iron core linear permanent magnet motor. The proposed method is based on the dq0-decomposition of the three-phase currents using Park’s transformation. The torque is decomposed into a reluctance component and two

  9. Sliding-Mode Control to Compensate PVT Variations in Dual Core Systems

    NARCIS (Netherlands)

    Pourshaghaghi, H.R.; Fatemi, S.H.; Pineda de Gyvez, J.

    2012-01-01

    In this paper, we present a novel robust sliding-mode controller for stabilizing supply voltage and clock frequency of dual core processors determined by dynamic voltage and frequency scaling (DVFS) methods in the presence of systematic and random variations. We show that maximum rejection for

  10. BOOK REVIEW: Fundamentals of Plasma Physics and Controlled Fusion

    Science.gov (United States)

    Brambilla, Marco

    1998-04-01

    Professor Kenro Miyamoto, already well known for his textbook Plasma Physics for Nuclear Fusion (MIT Press, Cambridge, MA, 1976; revised edition 1989), has now published a new book entitled Fundamentals of Plasma Physics and Controlled Fusion (Iwanami Book Service Center, Tokyo, 1997). To a large extent, the new book is a somewhat shortened and well reorganized version of its predecessor. The style, concise and matter of fact, clearly shows the origin of the text in lectures given by the author to graduate students. As announced by the title, the book is divided into two parts: the first part (about 250 pages) is a general introduction to the physics of plasmas, while the second, somewhat shorter, part (about 150 pages), is devoted to a description of the most important experimental approaches to achieving controlled thermonuclear fusion. Even in the first part, moreover, the choice of subjects is consistently oriented towards the needs of fusion research. Thus, the introduction to the behaviour of charged particles (particle motion, collisions, etc.) and to the collective description of plasmas is quite short, although the reader will get a flavour of all the most important topics and will find a number of examples chosen for their relevance to fusion applications (only the presentation of the Vlasov equation, in the second section of Chapter 4, might be criticized as so concise as to be almost misleading, since the difference between microscopic and macroscopic fields is not even mentioned). Considerably more space is devoted to the magnetohydrodynamic (MHD) description of equilibrium and stability. This part includes the solution of the Grad-Shafranov equation for circular tokamaks, a brief discussion of Pfirsch-Schlüter, neoclassical and anomalous diffusion, and two relatively long chapters on the most important ideal and resistive MHD instabilities of toroidal plasmas; drift and ion temperature gradient driven instabilities are also briefly presented. The

  11. [Automatic adjustment control system for DC glow discharge plasma source].

    Science.gov (United States)

    Wan, Zhen-zhen; Wang, Yong-qing; Li, Xiao-jia; Wang, Hai-zhou; Shi, Ning

    2011-03-01

    There are three important parameters in the DC glow discharge process, the discharge current, discharge voltage and argon pressure in discharge source. These parameters influence each other during glow discharge process. This paper presents an automatic control system for DC glow discharge plasma source. This system collects and controls discharge voltage automatically by adjusting discharge source pressure while the discharge current is constant in the glow discharge process. The design concept, circuit principle and control program of this automatic control system are described. The accuracy is improved by this automatic control system with the method of reducing the complex operations and manual control errors. This system enhances the control accuracy of glow discharge voltage, and reduces the time to reach discharge voltage stability. The glow discharge voltage stability test results with automatic control system are provided as well, the accuracy with automatic control system is better than 1% FS which is improved from 4% FS by manual control. Time to reach discharge voltage stability has been shortened to within 30 s by automatic control from more than 90 s by manual control. Standard samples like middle-low alloy steel and tin bronze have been tested by this automatic control system. The concentration analysis precision has been significantly improved. The RSDs of all the test result are better than 3.5%. In middle-low alloy steel standard sample, the RSD range of concentration test result of Ti, Co and Mn elements is reduced from 3.0%-4.3% by manual control to 1.7%-2.4% by automatic control, and that for S and Mo is also reduced from 5.2%-5.9% to 3.3%-3.5%. In tin bronze standard sample, the RSD range of Sn, Zn and Al elements is reduced from 2.6%-4.4% to 1.0%-2.4%, and that for Si, Ni and Fe is reduced from 6.6%-13.9% to 2.6%-3.5%. The test data is also shown in this paper.

  12. Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

    Energy Technology Data Exchange (ETDEWEB)

    Matsukawa, M. E-mail: matsukaw@naka.jaeri.go.jp; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T

    2003-09-01

    The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.

  13. Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

    International Nuclear Information System (INIS)

    Matsukawa, M.; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T.

    2003-01-01

    The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control

  14. Multifarious Physics Analyses of the Core Plasma Properties in a Helical DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    Miyazawa, J.; Satake, S.; Goto, T.; Seki, R.; Nunami, M.; Funaba, H.; Yamada, I.; Suzuki, C.; Sakamoto, R.; Motojima, G.; Yamada, H.; Sagara, A., E-mail: miyazawa@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki (Japan); Yokoyama, M.; Suzuki, Y.; Masaoka, Y.; Murakami, S. [Departement Nuclear Engineering, Kyoto University, Kyoto (Japan)

    2012-09-15

    Full text: Theoretical analyses on the MHD equilibrium, the neoclassical transport, and the alpha particle transport, etc., are being carried out for a helical fusion DEMO reactor named FFHR- d1, using radial profiles extrapolated from LHD. FFHR-d1 is a heliotron type DEMO reactor of which the conceptual design activity has been launched since 2010. It is possible to sustain the burning plasma without auxiliary heating (i.e., self-ignition) in FFHR-d1, since there is no need of plasma current drive in heliotron plasmas. The device size is 4 times enlarged from LHD, i.e., the major radius of helical coil center is 15.6 m, the magnetic field strength at the helical coil center is 4.7 T, and the fusion output is {approx} 3 GW. One of the distinguished subjects in FFHR-d1 compared with the former FFHR design series is the robust similarity with LHD. The arrangement of superconducting magnet coils in FFHR-d1 is similar to that of LHD, except a pair of planar poloidal coils omitted to maximize the maintenance ports. This makes reasonable to assume a similar MHD equilibrium as observed in LHD for FFHR-d1, as long as the beta profiles in these two are similar. In FFHR-d1, radial profiles of density and temperature are determined by multiplying proper enhancement factors on those obtained in LHD, according to the DPE (Direct Profile Extrapolation) method. The enhancement factors are calculated consistently with the gyro-Bohm model. Therefore, the global confinement properties as expressed in ISS95 or ISS04 are kept in FFHR-d1. A large Shafranov shift is foreseen in FFHR-d1 due to its high-beta property. This leads to deterioration in the neoclassical transport and alpha particle confinement. Effectiveness of plasma position control and/or magnetic configuration optimization has been examined to solve this problem and to check the validity of extrapolated profiles. According to these analyses, it is concluded that the self-ignition condition can be achieved in FFHR-d1 by

  15. Magnetic Detachment and Plume Control in Escaping Magnetized Plasma

    International Nuclear Information System (INIS)

    Schmit, P.F.; Fisch, N.J.

    2008-01-01

    The model of two-fluid, axisymmetric, ambipolar magnetized plasma detachment from thruster guide fields is extended to include plasmas with non-zero injection angular velocity profiles. Certain plasma injection angular velocity profiles are shown to narrow the plasma plume, thereby increasing exhaust efficiency. As an example, we consider a magnetic guide field arising from a simple current ring and demonstrate plasma injection schemes that more than double the fraction of useful exhaust aperture area, more than halve the exhaust plume angle, and enhance magnetized plasma detachment

  16. Management of complex data flows in the ASDEX Upgrade plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, Wolfgang, E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Neu, Gregor; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Cole, Richard; Lueddecke, Klaus [Unlimited Computer Systems, Iffeldorf (Germany)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Control system architectures with data-driven workflows are efficient, flexible and maintainable. Black-Right-Pointing-Pointer Signal groups provide coherence of interrelated signals and increase the efficiency of process synchronisation. Black-Right-Pointing-Pointer Sample tags indicating sample quality form the fundament of a local event handling strategy. Black-Right-Pointing-Pointer A self-organising workflow benefits from sample tags consisting of time stamp and stream activity. - Abstract: Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes. The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short. Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals

  17. Management of complex data flows in the ASDEX Upgrade plasma control system

    International Nuclear Information System (INIS)

    Treutterer, Wolfgang; Neu, Gregor; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas; Cole, Richard; Lüddecke, Klaus

    2012-01-01

    Highlights: ► Control system architectures with data-driven workflows are efficient, flexible and maintainable. ► Signal groups provide coherence of interrelated signals and increase the efficiency of process synchronisation. ► Sample tags indicating sample quality form the fundament of a local event handling strategy. ► A self-organising workflow benefits from sample tags consisting of time stamp and stream activity. - Abstract: Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes. The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short. Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals connecting process outputs and inputs. These are implemented as real-time streams of data samples

  18. Physics of the conceptual design of the ITER plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Snipes, J.A., E-mail: Joseph.Snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bremond, S. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Casper, T. [1166 Bordeaux St, Pleasanton, CA 94566 (United States); Douai, D. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Humphreys, D. [General Atomics, San Diego, CA 92186 (United States); Lister, J. [Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne (EPFL), CRPP, Lausanne CH-1015 (Switzerland); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sugihara, M., E-mail: Sugihara_ma@yahoo.co.jp [Japan (Japan); Winter, A.; Zabeo, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2014-05-15

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  19. Physics of the conceptual design of the ITER plasma control system

    International Nuclear Information System (INIS)

    Snipes, J.A.; Bremond, S.; Campbell, D.J.; Casper, T.; Douai, D.; Gribov, Y.; Humphreys, D.; Lister, J.; Loarte, A.; Pitts, R.; Sugihara, M.; Winter, A.; Zabeo, L.

    2014-01-01

    Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

  20. Real-time software for the COMPASS tokamak plasma control

    International Nuclear Information System (INIS)

    Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Sartori, F.; Janky, F.; Cahyna, P.; Hron, M.; Panek, R.

    2010-01-01

    The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 μs. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

  1. Real-time software for the COMPASS tokamak plasma control

    Energy Technology Data Exchange (ETDEWEB)

    Valcarcel, D.F., E-mail: danielv@ipfn.ist.utl.p [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P-1049-001 Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic)

    2010-07-15

    The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 {mu}s. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

  2. Study of Globus-M Tokamak Poloidal System and Plasma Position Control

    Science.gov (United States)

    Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

    2017-12-01

    In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

  3. A Testbed For Validating the LHC Controls System Core Before Deployment

    CERN Document Server

    Nguyen Xuan, J

    2011-01-01

    Since the start-up of the LHC, it is crucial to carefully test core controls components before deploying them operationally. The Testbed of the CERN accelerator controls group was developed for this purpose. It contains different hardware (PPC, i386) running various operating systems (Linux and LynxOS) and core software components running on front-ends, communication middleware and client libraries. The Testbed first executes integration tests to verify that the components delivered by individual teams interoperate, and then system tests, which verify high-level, end-user functionality. It also verifies that different versions of components are compatible, which is vital, because not all parts of the operational LHC control system can be upgraded simultaneously. In addition, the Testbed can be used for performance and stress tests. Internally, the Testbed is driven by Atlassian Bamboo, a Continuous Integration server, which builds and deploys automatically new software versions into the Test...

  4. Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schultz, B.G.

    1979-05-01

    A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

  5. EURATOM-CEA association contributions to the 26. EPS conference on controlled fusion and plasma physics, Maastricht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-10-15

    This report references the EURATOM-CEA association contributions presented at the 26. EPS conference on controlled fusion and plasma physics, in Maastricht (Netherlands) the 14-18 June 1999. Two invited papers and 24 contributed papers are proposed. They deal with: tokamak devices; particle recirculation in ergodic divertor; current profile control and MHD stability in Tore Supra discharges; edge-plasma control by the ergodic divertor; electron heat transport in stochastic magnetic layer; bolometry and radiated power; particle collection by ergodic divertor; study and simulation of pa impurities; line shape modelling for plasma edge conditions; dynamical study of the radial structure of the fluctuations measured by reciprocating Langmuir probe in Tore Supra; up-down asymmetry of density fluctuations; Halo currents in a circular tokamak; real time measurement of the position, density, profile and current profile at Tore Supra; poloidal rotation measurement by reflectometry; interpretation of q-profile dependence of the LH power deposition profile during LHCD experiments; ICFR plasma production and optimization; improved core electron confinement; measurement of hard X-ray emission profile; modelling of shear effects on thermal and particles transport; ion turbulence; current drive generation based on autoresonance and intermittent trapping mechanisms. (A.L.B.)

  6. JSPS-CAS Core University Program seminar. Proceedings of Japan-China joint seminar on atomic and molecular processes in plasma

    Energy Technology Data Exchange (ETDEWEB)

    Koike, Fumihiro [Kitasato Univ., Tokyo (Japan); Dong, Chenzhong [Northwest Normal Univ., Lanzhou (China)

    2005-02-01

    As one of the activities of JSPS-CAS Core University Program, Japan-China Joint Seminar on Atomic and Molecular Processes in Plasma was held on March 6 - 11, 2004 in Lanzhou, China. The total number of the officially registered participants was 29, in which 17 from Japan, 10 from China, and 2 from Germany. In the nuclear fusion plasma, there are quite a variety of atomic processes such as ionization, excitation, radiative recombination, non-radiative recombination (di-electronic recombination, collisional electron transfer), cascade radiation, and cascade Auger decay over the wide range of plasma temperature. The knowledge of such the processes is indispensable for the evaluation and improvement of the plasma properties, which is desirable to be investigated by international collaboration groups. The present Japan-China Joint Seminar constitutes one of such the activities to realize the above stated aim. The 21 of the presented papers are indexed individually. (J.P.N.)

  7. Proceedings of JSPS-CAS Core University Program seminar on production and steady state confinement of high performance plasmas in magnetic confinement systems

    International Nuclear Information System (INIS)

    Wan Baonian; Toi, Kazuo

    2005-09-01

    The JSPS-CAS Core University Program (CUP) seminar on 'Production and steady-state confinement of high performance plasmas in magnetic confinement systems' was held from 27 July to 29 July 2005 in Institute of Plasma Physics, the Chinese Academy of Sciences, Hefei, China. This seminar was organized in the framework of CUP in the field of plasma and nuclear fusion. About 50 persons including 20 Japanese attendees attended this seminar. Long time sustainment of high confinement and high beta plasmas is crucial for realization of an advanced nuclear fusion reactor. This seminar was motivated to summarize the results of CUP obtained in four years activities of CUP, and to extract crucial issues to be resolved near future, which must drive near and mid- term collaborations in the framework of CUP. The 32 of presented papers are indexed individually. (J.P.N.)

  8. Cold Plasmas for Biofilm Control: Opportunities and Challenges.

    Science.gov (United States)

    Gilmore, Brendan F; Flynn, Padrig B; O'Brien, Séamus; Hickok, Noreen; Freeman, Theresa; Bourke, Paula

    2018-06-01

    Bacterial biofilm infections account for a major proportion of chronic and medical device associated infections in humans, yet our ability to control them is compromised by their inherent tolerance to antimicrobial agents. Cold atmospheric plasma (CAP) represents a promising therapeutic option. CAP treatment of microbial biofilms represents the convergence of two complex phenomena: the production of a chemically diverse mixture of reactive species and intermediates, and their interaction with a heterogeneous 3D interface created by the biofilm extracellular polymeric matrix. Therefore, understanding these interactions and physiological responses to CAP exposure are central to effective management of infectious biofilms. We review the unique opportunities and challenges for translating CAP to the management of biofilms. Copyright © 2018. Published by Elsevier Ltd.

  9. A Control Method of Current Type Matrix Converter for Plasma Control Coil Power Supply

    International Nuclear Information System (INIS)

    Shimada, K.; Matsukawa, M.; Kurihara, K.; Jun-ichi Itoh

    2006-01-01

    In exploration to a tokamak fusion reactor, the control of plasma instabilities of high β plasma such as neoclassical tearing mode (NTM), resistive wall mode (RWM) etc., is the key issue for steady-state sustainment. One of the proposed methods to avoid suppressing RWM is that AC current having a phase to work for reduction the RWM growth is generated in a coil (sector coil) equipped spirally on the plasma vacuum vessel. To stabilize RWM, precise and fast real-time feedback control of magnetic field with proper amplitude and frequency is necessary. This implies that an appropriate power supply dedicated for such an application is expected to be developed. A matrix converter as one of power supply candidates for this purpose could provide a solution The matrix converter, categorized in an AC/AC direct converter composed of nine bi-directional current switches, has a great feature that a large energy storage element is unnecessary in comparison with a standard existing AC/AC indirect converter, which is composed of an AC/DC converter and a DC/AC inverter. It is also advantageous in cost and size of its applications. Fortunately, a voltage type matrix converter has come to be available at the market recently, while a current type matrix converter, which is advantageous for fast control of the large-inductance coil current, has been unavailable. On the background above mentioned, we proposed a new current type matrix converter and its control method applicable to a power supply with fast response for suppressing plasma instabilities. Since this converter is required with high accuracy control, the gate control method is adopted to three-phase switching method using middle phase to reduce voltage and current waveforms distortion. The control system is composed of VME-bus board with DSP (Digital Signal Processor) and FPGA (Field Programmable Gate Array) for high speed calculation and control. This paper describes the control method of a current type matrix converter

  10. Physical property control in core/shell inorganic nanostructures for fluorescence and magnetic targeting applications

    Science.gov (United States)

    Roberts, Stephen K.

    Nanomaterials show immense promise for the future in numerous areas of application. Properties that are unique from the bulk material and are tunable allow for innovation in material design. This thesis will focus on controlling the physical properties of core/shell nanostructures to enhance the utility of the materials. The first focus is on the impact of different solvent mixtures during the shell growth phase of SILAR based core/shell quantum dot synthesis is studied. Gaining insight into the mechanism for SILAR growth of core/shell nanoparticles allows improved synthetic yields and precursor binding, providing enhanced control to synthesis of core/shell nanoparticles. The second focus of this thesis is exploring the use of magnetic nanoparticles for magnetic drug targeting for cardiovascular conditions. Magnetic targeting for drug delivery enables increased local drug concentration, while minimizing non-specific interactions. In order to be effective for magnetic targeting, it must be shown that low magnetic strength is sufficient to capture flowing nanoparticles. By demonstrating the binding of a therapeutic agent to the surface at medicinal levels, the viability for use as a nanoparticle drug delivery system is improved.

  11. Utilization of control rod drive (CRD) system for long term core cooling

    International Nuclear Information System (INIS)

    Huerta B, A.

    1991-01-01

    In this paper we consider an application of Probabilistic Risk Assessment (PRA) to risk management. Foreseeable risk management strategies to prevent core damage are constrained by the availability of first line systems as well as support systems. The actual trend in the evaluation of risk management options can be performed in a number of ways. An example is the identification of back-up systems which could be used to perform the same safety functions. In this work we deal with the evaluation of the feasibility, for BWR's, to use the Control Rod Drive system to maintain an adequate reactor core long term cooling in some accident sequences. This preliminary evaluation is carried out as a part of the Internal Events Analysis for Laguna Verde Nuclear Power Plant (LVNPP) that is currently under way by the Mexican Nuclear Regulatory Body. This analysis addresses the evaluation and incorporation of all the systems, including the safety related and the back-up non safety related systems, that are available for the operator in order to prevent core damage. As a part of this analysis the containment venting capability is also evaluated as a back-up of the containment heat removal function. This will prevent the primary containment overpressurization and loss of certain core cooling systems. A selection of accident sequences in which the Control Rod Drive system could be used to mitigate the accident and prevent core damage are discussed. A personal computer transient analysis code is used to carry out thermohydraulic simulations in order to evaluate the Control Rod Drive system performance, the corresponding results are presented. Finally, some preliminary conclusions are drawn. (author). 9 refs, 5 figs

  12. PLGA-lecithin-PEG core-shell nanoparticles for controlled drug delivery.

    Science.gov (United States)

    Chan, Juliana M; Zhang, Liangfang; Yuet, Kai P; Liao, Grace; Rhee, June-Wha; Langer, Robert; Farokhzad, Omid C

    2009-03-01

    Current approaches to encapsulate and deliver therapeutic compounds have focused on developing liposomal and biodegradable polymeric nanoparticles (NPs), resulting in clinically approved therapeutics such as Doxil/Caelyx and Genexol-PM, respectively. Our group recently reported the development of biodegradable core-shell NP systems that combined the beneficial properties of liposomal and polymeric NPs for controlled drug delivery. Herein we report the parameters that alter the biological and physicochemical characteristics, stability, drug release properties and cytotoxicity of these core-shell NPs. We further define scalable processes for the formulation of these NPs in a reproducible manner. These core-shell NPs consist of (i) a poly(D,L-lactide-co-glycolide) hydrophobic core, (ii) a soybean lecithin monolayer, and (iii) a poly(ethylene glycol) shell, and were synthesized by a modified nanoprecipitation method combined with self-assembly. Preparation of the NPs showed that various formulation parameters such as the lipid/polymer mass ratio and lipid/lipid-PEG molar ratio controlled NP physical stability and size. We encapsulated a model chemotherapy drug, docetaxel, in the NPs and showed that the amount of lipid coverage affected its drug release kinetics. Next, we demonstrated a potentially scalable process for the formulation, purification, and storage of NPs. Finally, we tested the cytotoxicity using MTT assays on two model human cell lines, HeLa and HepG2, and demonstrated the biocompatibility of these particles in vitro. Our data suggest that the PLGA-lecithin-PEG core-shell NPs may be a useful new controlled release drug delivery system.

  13. Studies on performances of the control system of plasma position and shape

    International Nuclear Information System (INIS)

    Aikawa, Hiroshi; Tsuzuki, Naohisa; Kimura, Toyoaki; Ogata, Atsushi; Ninomiya, Hiromasa

    1978-09-01

    Performance in the control system of plasma position and shape is determined by estimating the disturbing field, system functions and load variation of the controlled object. Various stray fields are considered as disturbing field. Plasma internal inductance and poloidal beta are taken into consideration as load variation of the controlled object. The required performance is obtained through considerations of plasma equilibrium, stability, impurity concentration and sensors accuracy. The results are described as requests to the poloidal power supply system. (author)

  14. A Heterogeneous Multi-core Architecture with a Hardware Kernel for Control Systems

    DEFF Research Database (Denmark)

    Li, Gang; Guan, Wei; Sierszecki, Krzysztof

    2012-01-01

    Rapid industrialisation has resulted in a demand for improved embedded control systems with features such as predictability, high processing performance and low power consumption. Software kernel implementation on a single processor is becoming more difficult to satisfy those constraints....... This paper presents a multi-core architecture incorporating a hardware kernel on FPGAs, intended for high performance applications in control engineering domain. First, the hardware kernel is investigated on the basis of a component-based real-time kernel HARTEX (Hard Real-Time Executive for Control Systems...

  15. MCNP evaluation of top node control rod depletion below the core in KKL

    International Nuclear Information System (INIS)

    Beran, Tâm; Seltborg, Per; Lindahl, Sten-Örjan; Bieli, Roger; Ledergerber, Guido

    2014-01-01

    In previous studies, there has been identified a significant discrepancy in the BWR control rod top node depletion between the two core simulator nodal codes POLCA7 and PRESTO-2, which indicates that there is a large general uncertainty in nodal codes in calculating the top node depletion of fully withdrawn control rods. In this study, the stochastic Monte Carlo code MCNP has been used to calculate the top node control rod depletion for benchmarking the nodal codes. By using the TIP signal obtained from an extended TIP campaign below the core performed in the KKL reactor, the MCNP model has been verified by comparing the axial profile between the TIP data and the gamma flux calculated by MCNP. The MCNP results have also been compared with calculations from POLCA7, which was found to yield slightly higher depletion rates than MCNP. It was also found that the 10 B depletion in the top node is very sensitive to the exact axial location of the control rod top when it is fully withdrawn. By using the MCNP results, the neutron flux model below the core in the nodal codes can be improved by implementing an exponential function for the neutron flux. (author)

  16. Operation manual for the core flow test loop zone power-supply controller

    Energy Technology Data Exchange (ETDEWEB)

    Harper, R.E.

    1981-11-01

    The core flow test loop, which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at ORNL, is a high-pressure, high-temperature, out-of-reactor helium circulation system that is being constructed to permit study of the performance at steady-state and transient conditions of simulated segments of core assemblies for a GCFR demonstration plant. The simulated core segments, which are divided into zones, contain electrical heating elements to simulate the heat generated by fission. To control the power which is applied to a zone, a novel multitapped transformer and zone power control system have been designed and built which satisfy stringent design criteria. The controller can match power output to demand to within better than +-1% over a 900:1 dynamic range and perform full-power transients within 1 s. The power is applied in such a way as to minimize the electromagnetic interference at the bandwidth of the loop instrumentation, and the controller incorporates several error detection techniques, making it inherently fail-safe. The operation manual describes the specifications, operating instructions, error detection capabilities, error recovery, troubleshooting, calibration and QA procedures, and maintenance requirements. Also included are sections on the theory of operation, circuitry description, and a complete set of schematics.

  17. A microfluidic chip for blood plasma separation using electro-osmotic flow control

    International Nuclear Information System (INIS)

    Jiang, Hai; Weng, Xuan; Chon, Chan Hee; Wu, Xudong; Li, Dongqing

    2011-01-01

    In this paper, a microfluidic-based chip with two straight microchannels and five branch microchannels was designed and tested to separate blood plasma from a small sample of fresh human blood. The electro-osmotic flow method was used to control the separation of blood plasma. Blood cell removal and blood plasma extraction were realized in experiments. The efficiency of extracting blood plasma can be as high as 26%

  18. Twentyseventh European physical society conference on controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    Igitkhanov, Y.

    2000-01-01

    The twentyseventh European physical society conference on controlled fusion and plasma physics was held in Budapest, 12-16 June 2000. About 10 invited papers were presented, covering a wide range of problems in plasma physics, including confinement and transport issues in fusion devices, astrophysics and industrial application of plasmas. More than 100 papers were presented on plasma theory and experiments from tokamaks and stellarators. Some of the ITER-relevant issues covered are described in this newsletter

  19. XXXII Zvenigorod conference on the plasma physics and controlled thermonuclear synthesis. Theses of reports

    International Nuclear Information System (INIS)

    2005-01-01

    Theses of the reports, presented at the XXXII International conference on the plasma physics and controlled thermonuclear synthesis (Zvenigorod, 14-18 February 2005) are published. The total number of reports is 322, including 16 summarizing ones. The other reports are distributed by the following sections: magnetic confinement of high-temperature plasma (88 reports), inertial thermonuclear fusion (65), physical processes in low-temperature plasma (99) and physical bases of the plasma and beam technologies (54) [ru

  20. A key to improved ion core confinement in the JET tokamak: ion stiffness mitigation due to combined plasma rotation and low magnetic shear.

    Science.gov (United States)

    Mantica, P; Angioni, C; Challis, C; Colyer, G; Frassinetti, L; Hawkes, N; Johnson, T; Tsalas, M; deVries, P C; Weiland, J; Baiocchi, B; Beurskens, M N A; Figueiredo, A C A; Giroud, C; Hobirk, J; Joffrin, E; Lerche, E; Naulin, V; Peeters, A G; Salmi, A; Sozzi, C; Strintzi, D; Staebler, G; Tala, T; Van Eester, D; Versloot, T

    2011-09-23

    New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.

  1. Conceptual design of plasma position control of SST-1 tokamak using vertical field coil

    International Nuclear Information System (INIS)

    Gulati, Hitesh Kumar; Patel, Kiritkumar B.; Dhongde, Jasraj

    2015-01-01

    SST-1 (Steady State Superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼ 500 ms. SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s. Based on the solution of Grad-Shafranov equation the shift of plasma column center from geometrical centre of vacuum chamber is measured using various magnetic probes and flux loops installed in the machine. The closed feedback loop uses plasma current (Ip), Delta R as feedback signal and manipulate the vertical field current (Ivf). The discharge starts with feed forward loop using initially provided reference then the active feedback starts after discharge by few msec once plasma column is completely formed. The feedback loop time is of the order of 10 msec. The primary objective is to acquire plasma position control related signals, compute plasma position and generate position correction signal for VF coil power supply, communicate correction to VF coil power supply and modify VF power supply output in a deterministic time span. In this we present the methodology used for plasma horizontal displacement control using vertical field and discuss the preliminary results. (author)

  2. Control and balance of nuclear matters used for core fabrication of Super Phenix

    International Nuclear Information System (INIS)

    Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

    1987-05-01

    The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

  3. Performance Improvement of a Magnetized Coaxial Plasma Gun by adopting Iron-core Bias Coil and New Pre-Ionization System

    Science.gov (United States)

    Edo, Takahiro; Asai, T.; Tanaka, F.; Yamada, S.; Hosozawa, A.; Gota, H.; Roche, T.; Allfrey, I.; Matsumoto, T.

    2017-10-01

    A magnetized coaxial plasma gun (MCPG) is a device used to generate a compact toroid (CT), which has a spheromak-like configuration. A typical MCPG consists of a set of axisymmetric cylindrical electrodes, bias coil, and gas-puff valves. In order to expand the CT operating range, the distributions of the bias magnetic field and neutral gas have been investigated. We have developed a new means of generating stuffing flux. By inserting an iron core into the bias coil, the magnetic field increases dramatically; even a small current of a few Amps produces a sufficient bias field. According to a simulation result, it was also suggested that the radial distribution of the bias field is easily controlled. The ejected CT and the target FRC are cooled by excess neutral gas that typical MCPGs require to initiate a breakdown; therefore, we have adopted a miniature gun as a new pre-ionization (PI) system. By introducing this PI system, the breakdown occurs at lower neutral gas density so that the amount of excess neutral gas can be reduced.

  4. The symbol coding language for the BUTs processor of in-core reactor control systems

    International Nuclear Information System (INIS)

    Vorob'ev, D.M.; Golovanov, M.N.; Levin, G.L.; Parfenova, T.K.; Filatov, V.P.

    1978-01-01

    A symbolic coding language is described; it has been developed for automation of making up programs for in-core control systems. The systems use the ideology of the CAMAC-VECTOR system and include the BUTs-20 processor. The symbolic coding language has been developed as a programming language of the ASSEMBLER type. Operators of instructions and pseudo-instructions, the rules of reading in the text of the source program, and operator record formats are considered

  5. Generalized perturbation theory error control within PWR core-loading pattern optimization

    International Nuclear Information System (INIS)

    Imbriani, J.S.; Turinsky, P.J.; Kropaczek, D.J.

    1995-01-01

    The fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for PWR reactors. The code couples the optimization technique of simulated annealing (SA) with a generalized perturbation theory (GPT) model for evaluating core physics characteristics. To ensure the accuracy of the GPT predictions, as well as to maximize the efficient of the SA search, a GPT error control method has been developed

  6. Development of three methods for control rod position monitoring based on fixed in-core neutron detectors

    International Nuclear Information System (INIS)

    Peng, Xingjie; Li, Qing; Wang, Kan

    2015-01-01

    Highlights: • Three methods are utilized separately to unfold the control rod position from the fixed in-core neutron detector measurements. • Fixed in-core neutron detector measurements are simulated by neutronics code SMART. • Numerical results show that all these methods can unfold the control rod position accurately. • Two correction strategies are proposed to correct the simulated fixed in-core detector signals. - Abstract: Nuclear reactor core power distribution on-line monitoring system is very important in core surveillance, and this system should have the ability to indicate some abnormal conditions, such as the unacceptable control rod misalignment. In this study, the methodologies of radial basis function neural network (RBFNN), group method of data handling (GMDH) and Levenberg–Marquardt (LM) algorithm are utilized separately to unfold the control rod position from the fixed in-core neutron detector measurements. For using these methods, a large number of in-core detector signals corresponding to various known rod positions are needed. These data can be generated by an advanced core calculation code. In this study, the neutronics code SMART was used. The simulation results show that all these methods can unfold the control rod position accurately, and the performance comparison shows that the regularized RBFNN performs best. Two correction strategies are proposed to correct the simulated fixed in-core detector signals and improve the rod position monitoring accuracy when there are mismatches between actual physical factors and modeled physical factors

  7. Worldwide collaborative efforts in plasma control software development

    International Nuclear Information System (INIS)

    Penaflor, B.G.; Ferron, J.R.; Walker, M.L.; Humphreys, D.A.; Leuer, J.A.; Piglowski, D.A.; Johnson, R.D.; Xiao, B.J.; Hahn, S.H.; Gates, D.A.

    2008-01-01

    This presentation will describe the DIII-D collaborations with various tokamak experiments throughout the world which have adapted custom versions of the DIII-D plasma control system (PCS) software for their own use. Originally developed by General Atomics for use on the DIII-D tokamak, the PCS has been successfully installed and used for the NSTX experiment in Princeton, the MAST experiment in Culham UK, the EAST experiment in China, and the Pegasus experiment in the University of Wisconsin. In addition to these sites, a version of the PCS is currently being developed for use by the KSTAR tokamak in Korea. A well-defined and robust PCS software infrastructure has been developed to provide a common foundation for implementing the real-time data acquisition and feedback control codes. The PCS infrastructure provides a flexible framework that has allowed the PCS to be easily adapted to fulfill the unique needs of each site. The software has also demonstrated great flexibility in allowing for different computing, data acquisition and real-time networking hardware to be used. A description of the current PCS software architecture will be given along with experiences in developing and supporting the various PCS installations throughout the world

  8. CORE STABILIZATION EXERCISES AFTER ACL RECONSTRUCTION SURGERY PROVIDES BETTER OUTCOMES: A RANDOMIZED CONTROLLED TRIAL

    Directory of Open Access Journals (Sweden)

    Dilpreet Kaur

    2015-12-01

    Full Text Available Background: Decreased core stability displaces center of gravity away from base of support reducing activity participation of athlete. Present study was conducted to study the effect of core stabilization exercises after reconstruction surgery of ACL on functional outcomes. Methods: 30 subjects following 5 months of ACL reconstruction were randomly assigned to either group that performed (study group or did not performed (control group additional core stabilization exercises in conjugation with standard rehabilitation protocol. Outcome measures were: activity level using Tegnar activity level scale and functional performance using triple hop test. Outcome measures were compared at day 1 and day 42 of the treatment. Result: Significant improvement was seen in the study group for Tegnar score with mean difference changing from 4.5 to 1.5 from day 1 to day 42 of treatment (p=0.039 while the control group showed improvement in mean difference changing from 3.8 to 1.4 (p=.045 from day 1 to day 42 of treatment. Highly significant improvement was seen in the study group for triple hop test with mean difference changing from 25 to 6.7 (p<.001 compared to the control group with mean difference changing from 15.2 to 9.7(p=.005 from day 1 to day 42 of treatment. Conclusions: Both the groups showed improvement for activity level and functional performance but highly significant improvement was seen in the study group for functional performance. Core stabilization exercises in conjugation with the standard ACL rehabilitation protocol results better improvement in the triple hop test.

  9. A method of taking control rod history into account in core simulation calculations for BWR'S

    International Nuclear Information System (INIS)

    Hojerup, C.F.; Nonbol, E.

    1990-01-01

    The problem of taking control rod history into account in core simulator codes using precalculated cross sections has been examined, and two methods have been devised and tested. The very demanding first method, using the accumulated control rod in burn-up as a parameter, turned out to be even more inaccurate than the much less demanding second method, which only requires two full burn-up histories, one with the control rod in all the time, and another with the control rod out all the time. From the analysis it can be seen that the proper treatment of the control rod history is quite important, both for the cross sections, as several per cent on the reactivity are at stake, as for the pin powers, which for some pins are very much affected

  10. Application of optimal control theory to laser heating of a plasma in a solenoidal magnetic field

    International Nuclear Information System (INIS)

    Neal, R.D.

    1975-01-01

    Laser heating of a plasma column confined by a solenoidal magnetic field is studied via modern optimal control techniques. A two-temperature, constant pressure model is used for the plasma so that the temperature and density are functions of time and location along the plasma column. They are assumed to be uniform in the radial direction so that refraction of the laser beam does not occur. The laser intensity used as input to the column at one end is taken as the control variable and plasma losses are neglected. The localized behavior of the plasma heating dynamics is first studied and conventional optimal control theory applied. The distributed parameter optimal control problem is next considered with minimum time to reach a specified final ion temperature criterion as the objective. Since the laser intensity can only be directly controlled at the input end of the plasma column, a boundary control situation results. The problem is unique in that the control is the boundary value of one of the state variables. The necessary conditions are developed and the problem solved numerically for typical plasma parameters. The problem of maximizing the space-time integral of neutron production rate in the plasma is considered for a constant distributed control problem where the laser intensity is assumed fixed at maximum and the external magnetic field is taken as a control variable

  11. Nonequilibrium phenomena and determination of plasma parameters in the hot core of the cathode region in free-burning arc discharges

    International Nuclear Information System (INIS)

    Kuehn, Gerrit; Kock, Manfred

    2007-01-01

    We present spectroscopic measurements of plasma parameters (electron density n e , electron temperature T e , gas temperature T g , underpopulation factor b) in the hot-core region in front of the cathode of a low-current, free-burning arc discharge in argon under atmospheric pressure. The discharge is operated in the hot-core mode, creating a hot cathode region with plasma parameters similar to high-current arcs in spite of the fact that we use comparatively low currents (less than 20 A). We use continuum emission and (optically thin) line emission to determine n e and T e . We apply relaxation measurements based on a power-interruption technique to investigate deviations from local thermodynamic equilibrium (LTE). These measurements let us determine the gas temperature T g . All measurements are performed side-on with charge-coupled-device cameras as detectors, so that all measured plasma parameters are spatially resolved after an Abel inversion. This yields the first ever spatially resolved observation of the non-LTE phenomena of the hot core in the near-cathode region of free-burning arcs. The results only partly coincide with previously published predictions and measurements in the literature

  12. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  13. Reconfigurable ATCA hardware for plasma control and data acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, B.B., E-mail: bernardo@ipfn.ist.utl.p [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Batista, A.J.N.; Correia, M.; Neto, A.; Fernandes, H.; Goncalves, B.; Sousa, J. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal)

    2010-07-15

    The IST/EURATOM Association is developing a new generation of control and data acquisition hardware for fusion experiments based on the ATCA architecture. This emerging open standard offers a significantly higher data throughput over a reliable High Availability (HA) mechanical and electrical platform. One of this ATCA boards has 32 galvanically isolated ADC channels (18 bit) each mounted on a swappable plug-in card, 8 DAC channels (16 bit), 8 digital I/O channels and embeds a high performance XILINX Virtex 4 family field programmable gate array (FPGA). The specific modular and configurable hardware design enables adaptable utilization of the board in dissimilar applications. The first configuration, specially developed for tokamak plasma Vertical Stabilization, consists of a Multiple-Input-Multiple-Output (MIMO) controller that is capable of feedback loops faster than 1 ms using a multitude of input signals fed from different boards communicating through the Aurora{sup TM} point-to-point protocol. Massive parallel algorithms can be implemented on the FPGA either with programmed digital logic, using a HDL hardware description language, or within its internal silicon PowerPC{sup TM} running a full fledged real-time operating system. The second board configuration is dedicated for transient recording of the entire 32 channels at 2 MSamples/s to the on-board 512 MB DDR2 memory. Signal data retrieval is accelerated by a DMA-driven PCI Express{sup TM} x1 Interface to the ATCA system controller, providing an overall throughput in excess of 100 MB/s. This paper illustrates these developments and discusses possible configurations for foreseen applications.

  14. Controllable synthesis and characterization of novel copper-carbon core-shell structured nanoparticles

    International Nuclear Information System (INIS)

    Zhai, Jing; Tao, Xia; Pu, Yuan; Zeng, Xiao-Fei; Chen, Jian-Feng

    2011-01-01

    Highlights: → We reported a facile, green and cheap hydrothermal method to obtain novel copper-carbon core-shell nanoparticles. → The as-formed particles with controllable size and morphology are antioxidant. → The particles with organic-group-loaded surfaces and protective shells are expected to be applied in fields of medicine, electronics, sensors and lubricant. -- Abstract: A facile hydrothermal method was developed for preparing copper-carbon core-shell structured particles through a reaction at 160 o C in which glucose, copper sulfate pentahydrate and cetyltrimethylammonium bromide were used as starting materials. The original copper-carbon core-shell structured particles obtained were sized of 100-250 nm. The thickness of carbonaceous shells was controlled ranging from 25 to 100 nm by adjusting the hydrothermal duration time and the concentrations of glucose in the process. Products were characterized with transmission electron microscopy, X-ray diffraction, energy dispersive spectroscopy, Fourier transform infrared spectroscopy. Since no toxic materials were involved in the preparation, particles with stable carbonaceous framework and reactive surface also showed promising applications in medicine, electronics, sensors, lubricant, etc.

  15. Controllable synthesis and characterization of novel copper-carbon core-shell structured nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Jing [Sin-China Nano Technology Center, Key Lab for Nanomaterials, Ministry of Education, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, No. 15 Beisanhuan Dong Lu, Beijing 100029 (China); Tao, Xia; Pu, Yuan; Zeng, Xiao-Fei [Sin-China Nano Technology Center, Key Lab for Nanomaterials, Ministry of Education, Beijing University of Chemical Technology, Beijing 100029 (China); Chen, Jian-Feng, E-mail: chenjf@mail.buct.edu.cn [Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, No. 15 Beisanhuan Dong Lu, Beijing 100029 (China)

    2011-06-15

    Highlights: {yields} We reported a facile, green and cheap hydrothermal method to obtain novel copper-carbon core-shell nanoparticles. {yields} The as-formed particles with controllable size and morphology are antioxidant. {yields} The particles with organic-group-loaded surfaces and protective shells are expected to be applied in fields of medicine, electronics, sensors and lubricant. -- Abstract: A facile hydrothermal method was developed for preparing copper-carbon core-shell structured particles through a reaction at 160 {sup o}C in which glucose, copper sulfate pentahydrate and cetyltrimethylammonium bromide were used as starting materials. The original copper-carbon core-shell structured particles obtained were sized of 100-250 nm. The thickness of carbonaceous shells was controlled ranging from 25 to 100 nm by adjusting the hydrothermal duration time and the concentrations of glucose in the process. Products were characterized with transmission electron microscopy, X-ray diffraction, energy dispersive spectroscopy, Fourier transform infrared spectroscopy. Since no toxic materials were involved in the preparation, particles with stable carbonaceous framework and reactive surface also showed promising applications in medicine, electronics, sensors, lubricant, etc.

  16. Determination of power peak factor using control rods, ex-core detectors and neural networks

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado

    2005-01-01

    This work presents a methodology based on the artificial neural network technique to predict in real time the power peak factor in a form that can be implemented in reactor protection systems. The neural network inputs were those available in the reactor protection systems, namely, the axial and quadrant power differences obtained from measured ex-core detector signals, and the position of control rods. The response of ex core detector signals was measured in experiments especially performed in the IPEN/MB-01 zero-power reactor. Several reactor states with different power density distribution were obtained by positioning the control rods in different configurations. The power distribution and its peak factor were calculated for each of these reactor states using the Citation code. The obtained results show that the power peak factor correlates well with the control rod position and the quadrant power difference, and with a lesser degree with the axial power differences. The data presented an inherent organisation and could be classified into different classes of power peak factor behaviour as a function of position of control rods, axial power difference and quadrant power difference. The RBF networks were able to identify classes and interpolate the power peak factor values. The relative error for the power peak factor estimation ranged from 0.19 % to 0.67 %, less than the one that was obtained performing a power density distribution map with in-core detectors. It was observed that the positions of control rods bear the detailed and localised information about the power density distribution, and that the axial and the quadrant power difference describe its global variations in the axial and radial directions. The results showed that the RBF and MLP networks produced similar results, and that a neural network correlation can be implemented in power reactor protection systems. (author)

  17. Control oriented modeling and simulation of the sawtooth instability in nuclear fusion tokamak plasmas

    NARCIS (Netherlands)

    Witvoet, G.; Westerhof, E.; Steinbuch, M.; Doelman, N.J.; Baar, de M.R.

    2009-01-01

    Tokamak plasmas in nuclear fusion are subject to various instabilities. A clear example is the sawtooth instability, which has both positive and negative effects on the plasma. To optimize between these effects control of the sawtooth period is necessary. This paper presents a simple control

  18. VME multiprocessor system for plasma control at the JT-60 Upgrade

    International Nuclear Information System (INIS)

    Kimura, T.; Kurihara, K.; Takahashi, M.; Kawamata, Y.; Akasaka, H.; Matsukawa, M.

    1989-01-01

    In this paper design and preliminary tests are reported of a VME multiprocessor system for the JT-60 Upgrade plasma control utilizing three MC88100 based RISC computers and VME buses. The design of the VME system was stimulated by faster and more accurate computation requirements for the plasma position and shape control

  19. Controlling laser ablation plasma with external electrodes. Application to sheath dynamics study and beam physics

    International Nuclear Information System (INIS)

    Isono, Fumika; Nakajima, Mitsuo; Hasegawa, Jun; Kawamura, Tohru; Horioka, Kazuhiko

    2013-01-01

    The potential of laser ablation plasma was controlled successfully by using external ring electrodes. We found that an electron sheath is formed at the plasma boundary, which plays an important role in the potential formation. When the positively biased plasma reaches a grounded grid, electrons in the plasma are turned away and ions are accelerated, which leads to the formation of a virtual anode between the grid and an ion probe. We think that this device which can raise the plasma potential up to order of kV can be applied to the study of sheath dynamics and to a new type of ion beam extraction. (author)

  20. Feedback control of plasma equilibrium with control system aided by personal computer on the JIPP T-IIU tokamak

    International Nuclear Information System (INIS)

    Tsuzuki, T.; Toi, K.; Matsuura, K.

    1991-04-01

    A feedback control system aided by a personal computer is developed to maintain plasma position on the required position in the JIPP T-IIU tokamak. The personal computer enables to adjust various control parameters easily. In this control system, a control demand for driving the power supply of feedback controlled vertical field coils is composed to be proportional to a total plasma current. This system has been successfully employed throughout the discharge where the plasma current substantially changes from zero to hundreds of kiloamperes, because the feedback control can be done, being independent of the plasma current. The analysis of this feedback control system taken into account of digital sampling agrees well with the experimental results. (author)

  1. Controllers for high-performance nuclear fusion plasmas

    NARCIS (Netherlands)

    Baar, de M.R.

    2012-01-01

    A succesful nuclear fusion reactor will confine plasma at hig temperatures and densities, with low thermal losses. The workhorse of the nuclear fusion community is the tokamak, a toroidal device in which plasmas are confined by poloidal and toroidal magnetic fields. Ideally, the confirming magnetic

  2. ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

    2004-01-01

    An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

  3. The segmented non-uniform dielectric module design for uniformity control of plasma profile in a capacitively coupled plasma chamber

    International Nuclear Information System (INIS)

    Xia, Huanxiong; Xiang, Dong; Yang, Wang; Mou, Peng

    2014-01-01

    Low-temperature plasma technique is one of the critical techniques in IC manufacturing process, such as etching and thin-film deposition, and the uniformity greatly impacts the process quality, so the design for the plasma uniformity control is very important but difficult. It is hard to finely and flexibly regulate the spatial distribution of the plasma in the chamber via controlling the discharge parameters or modifying the structure in zero-dimensional space, and it just can adjust the overall level of the process factors. In the view of this problem, a segmented non-uniform dielectric module design solution is proposed for the regulation of the plasma profile in a CCP chamber. The solution achieves refined and flexible regulation of the plasma profile in the radial direction via configuring the relative permittivity and the width of each segment. In order to solve this design problem, a novel simulation-based auto-design approach is proposed, which can automatically design the positional sequence with multi independent variables to make the output target profile in the parameterized simulation model approximate the one that users preset. This approach employs an idea of quasi-closed-loop control system, and works in an iterative mode. It starts from initial values of the design variable sequences, and predicts better sequences via the feedback of the profile error between the output target profile and the expected one. It never stops until the profile error is narrowed in the preset tolerance

  4. Overview of the preliminary design of the ITER plasma control system

    Science.gov (United States)

    Snipes, J. A.; Albanese, R.; Ambrosino, G.; Ambrosino, R.; Amoskov, V.; Blanken, T. C.; Bremond, S.; Cinque, M.; de Tommasi, G.; de Vries, P. C.; Eidietis, N.; Felici, F.; Felton, R.; Ferron, J.; Formisano, A.; Gribov, Y.; Hosokawa, M.; Hyatt, A.; Humphreys, D.; Jackson, G.; Kavin, A.; Khayrutdinov, R.; Kim, D.; Kim, S. H.; Konovalov, S.; Lamzin, E.; Lehnen, M.; Lukash, V.; Lomas, P.; Mattei, M.; Mineev, A.; Moreau, P.; Neu, G.; Nouailletas, R.; Pautasso, G.; Pironti, A.; Rapson, C.; Raupp, G.; Ravensbergen, T.; Rimini, F.; Schneider, M.; Travere, J.-M.; Treutterer, W.; Villone, F.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

    2017-12-01

    An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

  5. Control of tokamak plasma current and equilibrium with hybrid poloidal field coils

    International Nuclear Information System (INIS)

    Shimada, Ryuichi

    1982-01-01

    A control method with hybrid poloidal field system is considered, which comprehensively implements the control of plasma equilibrium and plasma current, those have been treated independently in Tokamak divices. Tokamak equilibrium requires the condition that the magnetic flux function value on plasma surface must be constant. From this, the current to be supplied to each coil is determined. Therefore, each coil current is the resultant of the component related to plasma current excitation and the component required for holding equilibrium. Here, it is intended to show a method by which the current to be supplied to each coil can easily be calculated by the introduction of hybrid control matrix. The text first considers the equilibrium of axi-symmetrical plasma and the equilibrium magnetic field outside plasma, next describes the determination of current using the above hybrid control matrix, and indicates an example of controlling Tokamak plasma current and equilibrium by the hybrid poloidal field coils. It also shows that the excitation of plasma current and the maintenance of plasma equilibrium can basically be available with a single power supply by the appropriate selection of the number of turns of each coil. These considerations determine the basic system configuration as well as decrease the installed capacity of power source for the poloidal field of a Tokamak fusion reactor. Finally, the actual configuration of the power source for hybrid poloidal field coils is shown for the above system. (Wakatsuki, Y.)

  6. The effectiveness of core stabilization exercise in adolescent idiopathic scoliosis: A randomized controlled trial.

    Science.gov (United States)

    Gür, Gözde; Ayhan, Cigdem; Yakut, Yavuz

    2017-06-01

    Core stabilization training is used to improve postural balance in musculoskeletal problems. The purpose of this study was to investigate the effectiveness of stabilization training in adolescent idiopathic scoliosis. A randomized controlled trial, pretest-posttest design. In total, 25 subjects with adolescent idiopathic scoliosis were randomly divided into two groups: stabilization group ( n = 12) and control group ( n = 13). The stabilization group received core stabilization in addition to traditional rehabilitation, and the control group received traditional rehabilitation for 10 weeks. Assessment included Cobb's angle on radiograph, apical vertebral rotation in Adam's test, trunk asymmetry (Posterior Trunk Symmetry Index), cosmetic trunk deformity (Trunk Appearance Perception Scale), and quality of life (Scoliosis Research Society-22 questionnaire). Inter-group comparisons showed significantly greater improvements in the mean change in lumbar apical vertebral rotation degree and the pain domain of Scoliosis Research Society-22 in the stabilization group than those in the control group ( p adolescent idiopathic scoliosis. Clinical relevance Stabilization exercises are more effective in reducing rotation deformity and pain than traditional exercises in the conservative rehabilitation of adolescent idiopathic scoliosis. These improvements suggest that stabilization training should be added to rehabilitation programs in adolescent idiopathic scoliosis.

  7. Modelling of profile control with LH wave injection in the HL-2A single-null divertor plasma

    International Nuclear Information System (INIS)

    Gao Qingdi; Yuan Baoshan; Li Fangzhu; Wang, Aike; Budny, R.V.

    2005-01-01

    In the HL-2A tokamak a single-null divertor configuration has been established. The separatrix of the single-null diverted plasma was identified with a filament model, and the determined striking area on the target plate is in agreement with the measurements of electric probe array. Higher power LH wave (1.5MW) is injected to the diverted plasma with a nearly symmetric spectrum. Dominant electron heating and current profile control are investigated with numerical simulation. Plasma heating by electron Landau interaction results in operation scenarios of preferentially dominant electron heating. Due to the off-axis driven current, an optimized q-profile is formed, and an enhanced confinement regime with steep electron temperature gradient is produced. The clear decrease of the electron thermal conductivity in the LH power deposition region shows that an electron-ITB is developed. When higher LH power injects into the target plasma that is heated by NBI (0.5MW), the ion temperature has a large increment in addition to the high increase of electron temperature. The temperature profiles indicate that an enhanced core confinement is established with both ion-ITB and electron-ITB developed. (author)

  8. Low temperature synthesis of silicon quantum dots with plasma chemistry control in dual frequency non-thermal plasmas.

    Science.gov (United States)

    Sahu, Bibhuti Bhusan; Yin, Yongyi; Han, Jeon Geon; Shiratani, Masaharu

    2016-06-21

    The advanced materials process by non-thermal plasmas with a high plasma density allows the synthesis of small-to-big sized Si quantum dots by combining low-temperature deposition with superior crystalline quality in the background of an amorphous hydrogenated silicon nitride matrix. Here, we make quantum dot thin films in a reactive mixture of ammonia/silane/hydrogen utilizing dual-frequency capacitively coupled plasmas with high atomic hydrogen and nitrogen radical densities. Systematic data analysis using different film and plasma characterization tools reveals that the quantum dots with different sizes exhibit size dependent film properties, which are sensitively dependent on plasma characteristics. These films exhibit intense photoluminescence in the visible range with violet to orange colors and with narrow to broad widths (∼0.3-0.9 eV). The observed luminescence behavior can come from the quantum confinement effect, quasi-direct band-to-band recombination, and variation of atomic hydrogen and nitrogen radicals in the film growth network. The high luminescence yields in the visible range of the spectrum and size-tunable low-temperature synthesis with plasma and radical control make these quantum dot films good candidates for light emitting applications.

  9. Shape-controlled synthesis of Au@Pd core-shell nanoparticles and their corresponding electrochemical properties

    KAUST Repository

    Song, Hyon Min

    2012-01-01

    The shape-controlled synthesis of Au@Pd core-shell nanoparticles (NPs) was successfully achieved through the emulsion phase generated during the phase transfer from organic to aqueous medium. Contrary to conventional epitaxial growth for obtaining core-shell structures, this method does not require high temperatures and does not have shape restrictions. © 2012 The Royal Society of Chemistry.

  10. Surface Plasma Arc by Radio-Frequency Control Study (SPARCS)

    International Nuclear Information System (INIS)

    Ruzic, David N.

    2013-01-01

    This paper is to summarize the work carried out between April 2012 and April 2013 for development of an experimental device to simulate interactions of o-normal detrimental events in a tokamak and ICRF antenna. The work was mainly focused on development of a pulsed plasma source using theta pinch and coaxial plasma gun. This device, once completed, will have a possible application as a test stand for high voltage breakdown of an ICRF antenna in extreme events in a tokamak such as edge-localized modes or disruption. Currently, DEVeX does not produce plasma with high temperature enough to requirement for an ELM simulator. However, theta pinch is a good way to produce high temperature ions. The unique characteristic of plasma heating by a theta pinch is advantageous for an ELM simulator due to its effective ion heating. The objective of the proposed work, therefore, is to build a test facility using the existing theta pinch facility in addition to a coaxial plasma gun. It is expected to produce a similar pulsed-plasma heat load to the extreme events in tokamaks and to be applied for studying interactions of hot plasma and ICRF antennas

  11. Electromagnetic Wave Transmittance Control using Anisotropic Plasma Lattice

    Science.gov (United States)

    Matlis, Eric; Corke, Thomas; Hoffman, Anthony

    2017-11-01

    Experiments of transmission through a lattice array of plasma columns have shown an absorption band close to the plasma frequency at 14 GHz. The beam was oriented at a 35° incident angle to the planar plasma cell. These experiments were designed to determine if the observed absorption was the result of the isotropic plasma medium or that of an anisotropic metamaterial. Transmission of the microwave energy was not consistent with an isotropic material in which absorption would monotonically increase below the plasma frequency. The experimental results are supported by an anisotropic model which was developed for the plasma permittivity using an effective medium approximation. The plasma columns were modeled as uniform rods with permittivity described by a Drude model while the components of the permittivity tensor was calculated using the Maxwell-Garnett effective medium theory. Electron densities of n = 4 x1012 cm-3 were assumed which is consistent with prior experimental measurements. This model confirms the existence of non-zero imaginary wave vector k in a narrow region centered about 14 GHz.

  12. Control of plasma poloidal shape and position in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

  13. Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Albrecht, H.; Malinauskas, A.P.

    1978-01-01

    A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

  14. Magnetophoresis behaviour at low gradient magnetic field and size control of nickel single core nanobeads

    Energy Technology Data Exchange (ETDEWEB)

    Benelmekki, M., E-mail: benelmekki@fisica.uminho.p [Centro de Fisica, Universidade do Minho, Braga (Portugal); Montras, A. [Sepmag Tecnologies, Parc Tecnologic del Valles, Barcelona (Spain); Martins, A.J.; Coutinho, P.J.G. [Centro de Fisica, Universidade do Minho, Braga (Portugal); Martinez, Ll.M. [Sepmag Technologies, Atlanta, GA (United States)

    2011-08-15

    Magnetic separation of organic compounds, proteins, nucleic acids and other biomolecules, and cells from complex reaction mixtures is becoming the most suitable solution for large production in bioindustrial purification and extraction processes. Optimal magnetic properties can be achieved by the use of metals. However, they are extremely sensitive to oxidation and degradation under atmospheric conditions. In this work Ni nanoparticles are synthesised by conventional solution reduction process with the addition of a non-ionic surfactant as a surface agent. The nanoparticles were surfacted in citric acid and then coated with silica to form single core Ni nanobeads. A magnetophoresis study at different magnetic field gradients and at the different steps of synthesis route was performed using Horizontal Low Gradient Magnetic Field (HLGMF) systems. The reversible aggregation times are reduced to a few seconds, allowing a very fast separation process. - Research highlights: Monodispersed single core Ni-silica core-shell structures were prepared. Control of Ni nanoparticles size was achieved using a non-ionic surfactant. Magnetophoresis at different magnetic field gradients was monitored. Magnetophoresis at different steps of synthesis route was performed. Attractive magnetic interactions overcome electrostatic repulsions.

  15. Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

    International Nuclear Information System (INIS)

    Welander, A.S.; Deranian, R.D.; Humphreys, D.A.; Leuer, J.A.; Walker, M.L.

    2005-01-01

    Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII-D, the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution

  16. Modeling and control of plasma rotation for NSTX using neoclassical toroidal viscosity and neutral beam injection

    Energy Technology Data Exchange (ETDEWEB)

    Goumiri, I. R. [Princeton Univ., NJ (United States). Mechanical and Aerospace Dept.; Rowley, C. W. [Princeton Univ., NJ (United States). Mechanical and Aerospace Dept.; Sabbagh, S. A. [Columbia Univ., New York, NY (United States). Dept. of Applied Physics and Applied Mathematics; Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gerhardt, S. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Boyer, M. D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Andre, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kolemen, E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Taira, K. [Florida State Univ, Dept Mech Engn, Tallahassee, FL USA.

    2016-02-19

    A model-based feedback system is presented to control plasma rotation in a magnetically confined toroidal fusion device, to maintain plasma stability for long-pulse operation. This research uses experimental measurements from the National Spherical Torus Experiment (NSTX) and is aimed at controlling plasma rotation using two different types of actuation: momentum from injected neutral beams and neoclassical toroidal viscosity generated by three-dimensional applied magnetic fields. Based on the data-driven model obtained, a feedback controller is designed, and predictive simulations using the TRANSP plasma transport code show that the controller is able to attain desired plasma rotation profiles given practical constraints on the actuators and the available measurements of rotation.

  17. Optimization of in-core fuel management and control rod strategy in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1975-01-01

    An in-core fuel management problem is formulated for the equilibrium fuel cycle in an N-region nuclear reactor model. The formulation shows that the infinite multiplication factor k infinity requisite for newly charged fuel can be separated into two terms - one corresponding to the average k infinity at the end of the cycle and the other representing the direct contribution of the shuffling scheme and control rod programming. This formulation is applied to a three-region cylindrical reactor to obtain simultaneous optimization of shuffling and control rod programming. It is demonstrated that this formulation aids greatly in gaining a better understanding of the effects of changes in the shuffling scheme and control rod programming on equilibrium fuel cycle performance. (auth.)

  18. Prevention and control of rheumatic heart disease: Overcoming core challenges in resource-poor environments

    Science.gov (United States)

    Dougherty, Scott; Beaton, Andrea; Nascimento, Bruno R; Zühlke, Liesl J; Khorsandi, Maziar; Wilson, Nigel

    2018-01-01

    Rheumatic heart disease (RHD) has long receded as a significant threat to public health in high-income countries. In low-resource settings, however, the specter of RHD remains unabated, as exemplified by recent data from the Global Burden of Diseases Study. There are many complex reasons for this ongoing global disparity, including inadequate data on disease burden, challenges in effective advocacy, ongoing poverty and inequality, and weak health systems, most of which predominantly affect developing nations. In this review, we discuss how each of these acts as a core challenge in RHD prevention and control. We then examine key lessons learnt from successful control programs in the past and highlight resources that have been developed to help create strong national RHD control programs. PMID:29440834

  19. Core elements of hospital antibiotic stewardship programs from the Centers for Disease Control and Prevention.

    Science.gov (United States)

    Pollack, Loria A; Srinivasan, Arjun

    2014-10-15

    The proven benefits of antibiotic stewardship programs (ASPs) for optimizing antibiotic use and minimizing adverse events, such as Clostridium difficile and antibiotic resistance, have prompted the Centers for Disease Control and Prevention (CDC) to recommend that all hospitals have an ASP. This article summarizes Core Elements of Hospital Antibiotic Stewardship Programs, a recently released CDC document focused on defining the infrastructure and practices of coordinated multidisciplinary programs to improve antibiotic use and patient care in US hospitals. Published by Oxford University Press on behalf of the Infectious Diseases Society of America 2014. This work is written by (a) US Government employee(s) and is in the public domain in the US.

  20. Heterogenous treatment of water gaps and control rods in core calculations

    International Nuclear Information System (INIS)

    Lindahl, S.Oe.

    1984-08-01

    The conventional fuel assembly homogenization process introduces errors in the gross power distribution of the order of 10%. A method to mitigate the homogenization error is proposed. The gaps in between the fuel pin domains, containing water, control rod blades, fuel channel shroud, etc, are described by their transmission and reflection properties while the fuel pin regions are represented by conventional, homogenized cross sections. A nodal equation, incorporating the gap transmission and reflection functions, is set up in a general form. By numerical testing on 2D core problems it is shown that the explicit gap treatment is capable of reducing the homogenization error by an order of magnitude. (author)

  1. Device for moving fuel elements and control bars from one place to another in reactor core

    International Nuclear Information System (INIS)

    Hoffmeister, B.

    1976-01-01

    A description is given of a device, including a gripper, for manipulation of fuel elements and control bars, or rods, in a reactor core in which a vertical mast is provided in which the gripper is nonrotatably reciprocable. The mast is rotatable on a work platform and latch elements are provided to halt the mast in selected rotated positions on the platform. Abutment means are also provided to limit the rotary movement of the mast on the platform to 360 0 . 11 Claims, 6 drawing figures

  2. Transition of RF internal antenna plasma by gas control

    Energy Technology Data Exchange (ETDEWEB)

    Hamajima, Takafumi; Yamauchi, Toshihiko; Kobayashi, Seiji; Hiruta, Toshihito; Kanno, Yoshinori [Advanced Institute of Industrial Technology, 1-10-40 HigashiOhi, Shinagawa-ku, Tokyo, 140-0011 (Japan); Japan Atomic Energy Agency, 2-4 Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan)

    2012-07-11

    The transition between the capacitively coupled plasma (CCP) and the inductively coupled plasma (ICP) was investigated with the internal radio frequency (RF) multi-turn antenna. The transition between them showed the hysteresis curve. The radiation power and the period of the self-pulse mode became small in proportion to the gas pressure. It was found that the ICP transition occurred by decreasing the gas pressure from 400 Pa.

  3. A Plasma Control and Gas Protection System for Laser Welding of Stainless Steel

    DEFF Research Database (Denmark)

    Juhl, Thomas Winther; Olsen, Flemming Ove

    1997-01-01

    A prototype shield gas box with different plasma control nozzles have been investigated for laser welding of stainless steel (AISI 316). Different gases for plasma control and gas protection of the weld seam have been used. The gas types, welding speed and gas flows show the impact on process...... stability and protection against oxidation. Also oxidation related to special conditions at the starting edge has been investigated. The interaction between coaxial and plasma gas flow show that the coaxial flow widens the band in which the plasma gas flow suppresses the metal plasma. In this band the welds...... are oxide free. With 2.7 kW power welds have been performed at 4000 mm/min with Ar / He (70%/30%) as coaxial, plasma and shield gas....

  4. Plasma control system for 'Day-One' operation of KSTAR tokamak

    International Nuclear Information System (INIS)

    Hahn, Sang-hee; Walker, M.L.; Kim, Kukhee; Ahn, H.S.; Penaflor, B.G.; Piglowski, D.A.; Johnson, R.D.; Choi, Jaehoon; Lee, Dong-keun; Kim, Jayhyun; Yoon, S.W.; Seo, Seong-Heon; Kim, H.T.; Kim, K.P.; Lee, T.G.; Park, M.K.; Bak, J.G.; Lee, S.G.; Nam, Y.U.; Eidietis, N.W.

    2009-01-01

    A complete plasma control system (PCS) has been developed for KSTAR's first plasma campaign as a collaborative project with the DIII-D team. The KSTAR real time plasma control system is based on a conceptual design by Jhang and Choi [Hogun Jhang, I.S. Choi, Fusion Engineering and Design 73 (2005) 35-49] and consists of a fast real-time computer/communication cluster and software derived from the GA-PCS [Penaflor, B.G., et.al., Fusion Engineering and Design, 83 (2) (2008) 176]. The system has been used for simulation testing, poloidal field (PF) coil power supply commissioning and first plasma control. The seven sets of up-down symmetric, superconducting PF coil/power supply systems have been successfully tested. Reflective memory (RFM) is utilized as the primary actuator/PCS real-time communication layer and PCS synchronization with KSTAR timing system and slower control devices is achieved through an EPICS implementation. Consistent feedback loop times of 100 microseconds has been achieved during PF coil power supply testing and first plasma commissioning. Here we present the 'Day-One' plasma control system in its final form for the first plasma experimental campaign of KSTAR and describe how the system has been utilized during magnet commissioning and plasma startup experiments.

  5. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    International Nuclear Information System (INIS)

    HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

    2002-01-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

  6. Shape-controlled synthesis of Au@Pd core-shell nanoparticles and their corresponding electrochemical properties

    KAUST Repository

    Song, Hyon Min; Anjum, Dalaver H.; Khashab, Niveen M.

    2012-01-01

    The shape-controlled synthesis of Au@Pd core-shell nanoparticles (NPs) was successfully achieved through the emulsion phase generated during the phase transfer from organic to aqueous medium. Contrary to conventional epitaxial growth for obtaining

  7. Status and prospects of activities on algorithms and methods in WWER-1000 core control

    International Nuclear Information System (INIS)

    Filimonov, P.; Krainov, Y.; Proselkov, V.

    1994-01-01

    On the basis of long-term operational experience and investigations the problems of WWER-1000 reactor control are discussed. Such control is needed for WWER-1000, as well as for its Western analog PWR, for suppressing the axially instable power density field resulted from non-equilibrium redistribution of Xe-135 nuclei in the reactor core. It has been found that an adequate assessment of the reactor state and the prediction of its response to various control actions is essential for the control of power density distribution. For this purpose a computerized operator's adviser with a reactor simulator realizing a physical reactor model based on BIPR-7 code is used. The operation experience of WWER-1000 shows that the available control algorithms allow, with a fair degree of assurance, the prevention of intensive xenon oscillations and the stabilization of the axial offset. But in connection with the renunciation of half-length control rods a new algorithm is under development which makes use of full-length control rods for suppressing the intensive xenon oscillations in the descending phase. A new method based on BIPR-7 and PERMAK codes is also being developed for estimating the value and rate of linear power rating change of the fuel elements in power cycling. 12 figs., 7 refs

  8. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  9. Status and prospects of activities on algorithms and methods in WWER-1000 core control

    Energy Technology Data Exchange (ETDEWEB)

    Filimonov, P; Krainov, Y; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    On the basis of long-term operational experience and investigations the problems of WWER-1000 reactor control are discussed. Such control is needed for WWER-1000, as well as for its Western analog PWR, for suppressing the axially instable power density field resulted from non-equilibrium redistribution of Xe-135 nuclei in the reactor core. It has been found that an adequate assessment of the reactor state and the prediction of its response to various control actions is essential for the control of power density distribution. For this purpose a computerized operator`s adviser with a reactor simulator realizing a physical reactor model based on BIPR-7 code is used. The operation experience of WWER-1000 shows that the available control algorithms allow, with a fair degree of assurance, the prevention of intensive xenon oscillations and the stabilization of the axial offset. But in connection with the renunciation of half-length control rods a new algorithm is under development which makes use of full-length control rods for suppressing the intensive xenon oscillations in the descending phase. A new method based on BIPR-7 and PERMAK codes is also being developed for estimating the value and rate of linear power rating change of the fuel elements in power cycling. 12 figs., 7 refs.

  10. Radiation control in the core shroud replacement project of Fukushima-Daiichi NPS Unit no.2

    International Nuclear Information System (INIS)

    Kokubun, Yasunori; Haraguchi, Kazuyuki; Yoshizawa, Yuji; Yamada, Yasuo

    2000-01-01

    In Fukushima-Daiichi NPS Unit no.2, the core shroud replacement was made following that of Unit no.3. This project involves replacement of wide-ranging equipment, with the project extending over a long period of time. This was expected to increase the dose equivalent of workers. Accordingly, various measures to lower the dose equivalent were planned and implemented. We outline radiation controls implemented during the project period. The shroud replacement project was a preventive maintenance project which consisted of replacing the core shroud and other internals with those less susceptible to stress corrosion cracking. Problems related to radiation control during the replacement project of Unit no.3 the year before last were summarized. We studied, planned, and implemented measures to be reflected in the project for Unit no.2. This was done to lower the dose equivalent as much as possible while paying due attention to safety and economy. For radiation control during the project for Unit no.2, experiments with Unit no.3 were fully exploited and any effective measures taken at that time were adopted in this project. Problems pointed out after that project with Unit no.3 resulted in new or improved measures being taken with Unit no.2. Measures taken over from the project with Unit no.3; a. Daily analysis of difference between expected and actual dose equivalents b. Dose reduction measures, chemical decontamination, temporary shield, flushing, etc.; New or improved measures; a. Dose reduction measures: Mechanical removal of radiation sources, strengthening of shield, etc.; b. Automatic remote control system; c. Use of new protective devices. With measures implemented as described above, the dose equivalent during shroud replacement of Unit no.2 was reduced by about 30% when compared with that (11.5 persons · Sv) in the case of Unit no.3. Implemented radiation controls will be checked and reviewed in future for reflection in projects with other units. (author)

  11. Feedback control modeling of plasma position and current during intense heating in ISX-B

    International Nuclear Information System (INIS)

    Charlton, L.A.; Swain, D.W.; Neilson, G.H.

    1979-08-01

    The ISX-B Tokamak at ORNL is designed to have 1.8 MW (and eventually 3 MW) of neutral beam power injected to heat the plasma. This power may raise the anti β of the plasma to over 5% in less than 50 msec if the plasma is MHD stable. The results of a numerical simulation of the feedback control system and poloidal coil power supplies necessary to control the resulting noncircular (D-shaped or elliptical) plasma are presented. The resulting feedback control system is shown to be straightforward, although nonlinear voltage-current dependence is assumed in the power supplies. The required power supplied to the poloidal coils in order to contain the plasma under the high heating rates is estimated

  12. Control rod drop accident analysis for the mixed core project in Ling Ao NPS

    International Nuclear Information System (INIS)

    Zhang Shishun; Zhou Zhou; Xiao Min

    2004-01-01

    AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

  13. A Heterogeneous Multi-core Architecture with a Hardware Kernel for Control Systems

    DEFF Research Database (Denmark)

    Li, Gang; Guan, Wei; Sierszecki, Krzysztof

    2012-01-01

    Rapid industrialisation has resulted in a demand for improved embedded control systems with features such as predictability, high processing performance and low power consumption. Software kernel implementation on a single processor is becoming more difficult to satisfy those constraints. This pa......Rapid industrialisation has resulted in a demand for improved embedded control systems with features such as predictability, high processing performance and low power consumption. Software kernel implementation on a single processor is becoming more difficult to satisfy those constraints......). Second, a heterogeneous multi-core architecture is investigated, focusing on its performance in relation to hard real-time constraints and predictable behavior. Third, the hardware implementation of HARTEX is designated to support the heterogeneous multi-core architecture. This hardware kernel has...... several advantages over a similar kernel implemented in software: higher-speed processing capability, parallel computation, and separation between the kernel itself and the applications being run. A microbenchmark has been used to compare the hardware kernel with the software kernel, and compare...

  14. Use of in-core self-powered detectors in the nuclear reactor control system

    International Nuclear Information System (INIS)

    Mitel'man, M.G.; Andreev, L.G.; Batenin, I.V.

    1975-01-01

    The paper describes the results of experiments conducted at the Obninsk atomic power station on the establishment of an integrated system for the control and automatic regulation of the distribution of energy liberation. The pick-ups used for this system are direct-charge detectors with rhodium emitters. The system, which consists of 12 integral DPZ-7 detectors distributed evenly on 2 mutually perpendicular diamaters and 2 assemblies, each containing 4 1n direct-charge detectors, monitors the distribution of neutron flow density according to core height. Increased signals from the detectors were fed into the summation block, in which the outgoing signal was also adjusted. The half-life of the rhodium was calculated and the non-inertia of the entire system ascertained. The total operating time of the system for the automatic control of reactor power is approximately 1 h. The influence of interference was not determined. During the experiment, the power of the reactor remained constant, i.e. its regulation by the direct-charge detectors inside the core was effective. (author)

  15. Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

    Directory of Open Access Journals (Sweden)

    Li Gang

    2016-01-01

    Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

  16. Use of plasma waves to create in Tokamaks quasi-stationary conditions required for controlled fusion

    International Nuclear Information System (INIS)

    Moreau, D.

    1993-04-01

    In this thesis are studied the coupling of hybrid waves to the plasma, multijunction antennas, hybrid wave stochastic propagation, fast wave current drive and lower-hybrid current drive experiments in Tore Supra and Jet. The possibility of decoupling current density profile and temperature give one more degree of freedom for the control of plasma in a configuration which is not very flexible

  17. Plasma density control in real-time on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Janky, F., E-mail: filip.janky.work@gmail.com [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 in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Havlicek, J. [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 in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Varavin, M.; Zacek, F.; Seidl, J.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)

    2015-10-15

    Highlights: • We fitted length of the chord of the interferometry crossing plasma in the different plasma scenarios. • We add correction to the actual length of the chord of the interferometry according to plasma shape and position in real-time code. • We used this correction to control plasma density in real-time. - Abstract: The electron density on COMPASS is measured using 2 mm microwave interferometer. Interferometer signal is used as an input for the feedback control loop, running under the MARTe real-time framework. Two different threads are used to calculate (fast 50 μs thread) and to control (slow 500 μs thread) the electron density. The interferometer measures a line averaged density along a measurement chord. This paper describes an approach to control the line-averaged electron density in a real-time loop, using a correction to the real plasma shape, the plasma position, and non-linear effects of the electron density measurement at high densities. Newly developed real-time electron density control give COMPASS the chance to control the electron density more accurately which is essential for parametric scans for diagnosticians, for physics experiments and also for achieving plasma scenarios with H-mode.

  18. Periodical plasma structures controlled by external magnetic field

    Science.gov (United States)

    Schweigert, I. V.; Keidar, M.

    2017-11-01

    The plasma of Hall thruster type in external magnetic field is studied in 2D3V kinetic simulations using PIC MCC method. The periodical structure with maxima of electron and ion densities is formed and becomes more pronounced with increase of magnetic field incidence angle in the plasma. These ridges of electron and ion densities are aligned with the magnetic field vector and shifted relative each other. This leads to formation of two-dimensional double-layers structure in cylindrical plasma chamber. Depending on Larmor radius and Debye length up to nineteen potential steps appear across the oblique magnetic field. The electrical current gathered on the wall is associated with the electron and ion density ridges.

  19. Divertor heat flux control and plasma-material interaction

    International Nuclear Information System (INIS)

    Kikuchi, Yusuke; Nagata, Masayoshi; Sawada, Keiji; Takamura, Shuichi; Ueda, Yoshio

    2014-01-01

    Development of reliable radiative-cooling divertors is essential in DEMO reactor because it uses low-activation materials with low heat removal and the plasma heat flux exhausted from the confined region is 5 times as large as in ITER. It is important to predict precisely the heat and particle flux toward the divertor plate by simulation. In this present article, theoretical and experimental data of the reflection, secondary emission and surface recombination coefficients of the divertor plate by ion bombardment are given and their effects on the power transmission coefficient are discussed. In addition, some topics such as the erosion process of the divertor plate by ELM and the plasma disruption, the thermal shielding due to the vapor layer on the divertor plate and the formation of fuzz structure on W by helium plasma irradiation, are described. (author)

  20. Periodical plasma structures controlled by external magnetic field

    Science.gov (United States)

    Schweigert, I. V.; Keidar, M.

    2017-06-01

    The characteristics of two-dimensional periodical structures in a magnetized plasma are studied using kinetic simulations. Ridges (i.e. spikes in electron and ion density) are formed and became more pronounced with an increase of magnetic field incidence angle in the plasma volume in the cylindrical chamber. These ridges are shifted relative to each other, which results in the formation of a two-dimensional double-layer structure. Depending on Larmor radius and Debye length up to 19 potential steps appear across the oblique magnetic field. The electrical current gathered into the channels is associated with the electron and ion density ridges.

  1. Feedback control of horizontal position and plasma surface shape in a non-circular tokamak

    International Nuclear Information System (INIS)

    Moriyama, Shin-ichi; Nakamura, Kazuo; Nakamura, Yukio; Itoh, Satoshi

    1986-01-01

    The linear model for the coupled horizontal position and plasma surface shape control in the non-circular tokamak device was described. It enables us to estimate easily the displacement and the distortion due to the changes in plasma pressure and current density distribution. The PI-controller and the optimal regulator were designed with the linear model. Transient-response analysis of the control system in the TRIAM-1M tokamak showed that the optimal regulator is superior to the PI-controller with regard to the mutual-interference between the position control system and the elongation control system. (author)

  2. Dynamics and feedback control of plasma equilibrium position in a tokamak

    International Nuclear Information System (INIS)

    Burenko, O.

    1983-01-01

    A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. 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. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems

  3. Organization of the in-core control system connection with the M-6000 computer

    International Nuclear Information System (INIS)

    Golovanov, M.N.; Duma, V.R.; Levin, G.L.; Filatov, V.P.

    1978-01-01

    Problems of organizing communication of a digital computer with the equipment of the in-core control system (CC) are discussed. Three possible modes of joint operation of the CC equipment and the digital computer are given. The off-line control device provides data collection, preliminary processing and recording servicing of peripheral requests, and data exchange with the digital computer; computer-controlled operation of the equipment makes it possible to control input-output operations of the CCS equipment, and also to retain the working capacity of the CCS system when the off-line control device is failed; during file exchange the data are transferred between the computer and the CCS equipment. Requirements for the communication unit design are drawn up. An analysis of existing methods of a digital computer interface with the equipment is presented, and substantiation of the proposed variant of connection of the communication unit directly to the branch highway is given. Operation of the CCS equipment under various conditions is considered. The flowsheet and description of the interface of the M-6000 computer with the CCS equipment are given

  4. JSPS-CAS Core University Program seminar. Proceedings of Japan-China joint seminar on atomic and molecular processes in plasma

    International Nuclear Information System (INIS)

    Koike, Fumihiro; Dong Chenzhong

    2010-02-01

    As one of the activities of JSPS-CAS Core University Program, Japan-China Joint Seminar on Atomic and Molecular Processes in Plasma was held on October 26 - 31, 2009 in Xi'an, China. The total number of the officially registered participants was 54, in which 18 from Japan, 35 from China, and 1 from USA. And this seminar is an extension of the last two seminars that were held on March 6 - 11, 2004 in Lanzhou, China, and on October 6 - 12, 2007 in Dunhuang, China. In the nuclear fusion plasma, there are quite a variety of atomic processes such as ionization, excitation, radiative recombination, non-radiative recombination (di-electronic recombination, collisional electron transfer), cascade radiation, and cascade Auger decay over the wide range of plasma temperature. The knowledge of those processes is indispensable for the evaluation and improvement of the plasma properties. Because of the diversity of the subject, it is desirable to investigate them by international collaboration groups. The present seminar may contribute to realize the above stated aim; especially it has given an opportunity for the collaborative workers to illustrate their achievements. This seminar summarizes the collaborative researches for the last decade and propose the issues for the future prospect. The 30 of the presented papers are indexed individually. (J.P.N.)

  5. JSPS-CAS core university program seminar. Proceedings of Japan-China joint seminar on atomic and molecular processes in plasma

    International Nuclear Information System (INIS)

    Koike, Fumihiro; Dong Chenzhong

    2008-03-01

    As one of the activities of JSPS-CAS Core University Program, Japan-China Joint Seminar on Atomic and Molecular Processes in Plasma was held on October 8 - 12, 2007 in Dunhuang, China. The total number of the officially registered participants was 41, in which 12 from Japan, 25 from China, and 4 from EU. And this seminar is an extension of the last seminar that was held on March 6 - 11, 2004 in Lanzhou, China. In the nuclear fusion plasma, there are quite a variety of atomic processes such as ionization, excitation, radiative recombination, non-radiative recombination (di-electronic recombination, collisional electron transfer), cascade radiation, and cascade Auger decay over the wide range of plasma temperature. The knowledge of such processes is indispensable for the evaluation and improvement of the plasma properties, which is desirable to be investigated by international collaboration groups. The present seminar constitutes one of such activities to realize the above stated aim; especially it has given an opportunity for the collaborative workers to illustrate their achievements. The 32 of the presented papers are indexed individually. (J.P.N.)

  6. Real time determination and control of the plasma localisation and internal inductance in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Saint-Laurent, F. E-mail: stlauren@drfc.cad.cea.fr; Martin, G

    2001-10-01

    The study of long-duration high-power discharges need an efficient real time control of the plasma parameters, especially the plasma position when RF heating systems are used. On Tore Supra, recent improvements have been carried out (i) for the poloidal interpolation and the radial extrapolation of the magnetic measurements, (ii) for a better feedback matrix converting the radial errors of the plasma position to voltage values for the poloidal generators, and (iii) for a very fast solution to find the plasma parameters from the knowledge of its surface. The plasma edge localisation is now controlled with a precision better than 1 cm and controlled within a few millimetres uncertainty for several tenths of seconds. Moreover, for advanced tokamak scenarios, a precise real time determination of safety factor, poloidal beta, internal inductance, Shafranov shift as well as the online computation of the electron density and current density profiles are now available on Tore Supra. These quantities compare well with results from batch calculations using an equilibrium code. To fulfil the new requirements of plasma control for the CIEL project, a local control of the plasma edge position and curvature is planned for the near future.

  7. 2001 activity report of the development and research line in controlled thermonuclear fusion of the Plasma Associated Laboratory

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto

    2002-01-01

    The year 2001 activities of the controlled thermonuclear fusion research line of the Plasma Associated Laboratory at the National Institute for Space Research - Brazil are reported. The report approaches the staff, participation in congresses, goals for the year 2002 and papers on Tokamak plasmas, plasma diagnostic, bootstraps, plasma equilibrium and diagnostic

  8. An intelligent nuclear reactor core controller for load following operations, using recurrent neural networks and fuzzy systems

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.

    2003-01-01

    In the last decade, the intelligent control community has paid great attention to the topic of intelligent control systems for nuclear plants (core, steam generator...). Papers mostly used approximate and simple mathematical SISO (single-input-single-output) model of nuclear plants for testing and/or tuning of the control systems. They also tried to generalize theses models to a real MIMO (multi-input-multi-output) plant, while nuclear plants are typically of complex nonlinear and multivariable nature with high interactions between their state variables and therefore, many of these proposed intelligent control systems are not appropriate for real cases. In this paper, we designed an on-line intelligent core controller for load following operations, based on a heuristic control algorithm, using a valid and updatable recurrent neural network (RNN). We have used an accurate 3-dimensional core calculation code to represent the real plant and to train the RNN. The results of simulation show that this intelligent controller can control the reactor core during load following operations, using optimum control rod groups manoeuvre and variable overlapping strategy. This methodology represents a simple and reliable procedure for controlling other complex nonlinear MIMO plants, and may improve the responses, comparing to other control systems

  9. Application of non-thermal plasmas to pollution control

    International Nuclear Information System (INIS)

    Penetrante, B.M.; Vogtlin, G.E.; Bardsley, J.N.; Vitello, P.A.; Wallman, P.H.

    1993-06-01

    Non-thermal plasma techniques can be used to destroy many types of hazardous molecules. They are particularly efficient when the toxic materials are present in very small concentrations. This paper discusses three particular applications of non-thermal plasmas: (1) decomposition of hydrogen sulfide (H 2 S), (2) removal of trichloroethylene (TCE), and (3) removal of nitrogen oxides (NO x ). Emphasis is placed on the energy cost for implementing the decomposition or removal of these pollutants. Some of the factors affecting the energy cost are discussed. The authors discuss in detail their work at LLNL on pulsed plasma processing for the treatment of NO x in diesel engine exhaust. The results suggest that their plasma reactor can remove up to 70% of NO with relatively high initial concentrations (up to 500 ppM) at a power consumption cost of 2.5% for an engine with an output of 14 kW and an exhaust gas flow rate of 1,200 liters per minute

  10. Application of non-thermal plasmas to pollution control

    International Nuclear Information System (INIS)

    Pen