text:
Energy Technology Data Exchange (ETDEWEB)
Wootton, A.J.
This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported.
Texas Experimental Tokamak. Technical progress report, April 1990--April 1993
Energy Technology Data Exchange (ETDEWEB)
Wootton, A.J.
This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported.
Texas Experimental Tokamak, a plasma research facility: Technical progress report
Energy Technology Data Exchange (ETDEWEB)
Wootton, A.J.
In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively.
Measurement of the electron and ion temperatures by the x-ray imaging crystal spectrometer on joint Texas experimental tokamak
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Yan, W.; Chen, Z. Y., E-mail: firstname.lastname@example.com; Huang, D. W.; Tong, R. H.; Wang, S. Y.; Wei, Y. N.; Ma, T. K.; Zhuang, G. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China); Jin, W. [Center of Interface Dynamics for Sustainability, China Academy of Engineering Physics, Chengdu, Sichuan 610200 (China); Lee, S. G. [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Shi, Y. J. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)
An x-ray imaging crystal spectrometer has been developed on joint Texas experimental tokamak for the measurement of electron and ion temperatures from the K{sub α} spectra of helium-like argon and its satellite lines. A two-dimensional multi-wire proportional counter has been applied to detect the spectra. The electron and ion temperatures have been obtained from the Voigt fitting with the spectra of helium-like argon ions. The profiles of electron and ion temperatures show the dependence on electron density in ohmic plasmas.
Observations of zonal flows in electrode biasing experiments on the Joint Texas Experimental tokamak
Shen, H. G.; Lan, T.; Chen, Z. P.; Kong, D. F.; Zhao, H. L.; Wu, J.; Sun, X.; Liu, A. D.; Xie, J. L.; Li, H.; Ding, W. X.; Liu, W. D.; Yu, C. X.; Xu, M.; Sun, Y.; Liu, H.; Wang, Z. J.; Zhuang, G.
Zonal flows (ZFs) are observed during the electrode biasing (EB) high confinement mode (H-mode) using Langmuir probe arrays on the edge of J-TEXT tokamak. The long-distance correlation characteristics of floating potentials and interactions with turbulence are studied. During positive biasing H-mode, either the geodesic acoustic mode or low frequency ZF increases. Strong suppression of radial transport by ZFs is found in the low frequency region. The components of the radial particle flux without and with EB are compared in the frequency domain. The interaction between ZFs and ambient turbulence is also discussed. The results show that the rate of ZFs' shear is comparable with that of E × B shear, suggesting that ZFs could be the trigger of the biasing H-mode.
The Texas Experimental Tokamak: A plasma research facility. A proposal submitted to the Department of Energy in response to Program Notice 95-10: Innovations in toroidal magnetic confinement systems
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The Fusion Research Center (FRC) at the University Texas will operate the tokamak TEXT-U and its associated systems for experimental research in basic plasma physics. While the tokamak is not innovative, the research program, diagnostics and planned experiments are. The fusion community will reap the benefits of the success in completing the upgrades (auxiliary heating, divertor, diagnostics, wall conditioning), developing diverted discharges in both double and single null configurations, exploring improved confinement regimes including a limiter H-mode, and developing unique, critical turbulence diagnostics. With these new regimes, the authors are poised to perform the sort of turbulence and transport studies for which the TEXT group has distinguished itself and for which the upgrade was intended. TEXT-U is also a facility for collaborators to perform innovative experiments and develop diagnostics before transferring them to larger machines. The general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The emphasis of the proposed research is to provide well-documented plasmas which will be used to suggest and evaluate theories, to explore control techniques, to develop advanced diagnostics and analysis techniques, and to extend current drive techniques. Up to 1 MW of electron cyclotron heating (ECH) will be used not only for heating but as a localized, perturbative tool. Areas of proposed research are: (1) core turbulence and transport; (2) edge turbulence and transport; (3) turbulence analysis; (4) improved confinement; (5) ECH physics; (6) Alfven wave current drive; and (7) diagnostic development.
TIBER: Tokamak Ignition/Burn Experimental Research. Final design report
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Henning, C.D.; Logan, B.G.; Barr, W.L.; Bulmer, R.H.; Doggett, J.N.; Johnson, B.M.; Lee, J.D.; Hoard, R.W.; Miller, J.R.; Slack, D.S.
The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs. (WRF)
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Meglicki, Z
We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs.
Status of neutron diagnostics on the experimental advanced superconducting tokamak
Energy Technology Data Exchange (ETDEWEB)
Zhong, G. Q.; Hu, L. Q., E-mail: email@example.com; Pu, N.; Zhou, R. J.; Xiao, M.; Cao, H. R.; Li, K.; Huang, J.; Xu, G. S.; Wan, B. N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States); Fan, T. S.; Peng, X. Y.; Du, T. F.; Ge, L. J. [School of Physics and State Key Laboratory of Nuclear Physics and Technology, Peking University, Chengfu Road 201, 100871 Beijing (China)
Neutron diagnostics have become a significant means to study energetic particles in high power auxiliary heating plasmas on the Experimental Advanced Superconducting Tokamak (EAST). Several kinds of neutron diagnostic systems have been implemented for time-resolved measurements of D-D neutron flux, fluctuation, emission profile, and spectrum. All detectors have been calibrated in laboratory, and in situ calibration using {sup 252}Cf neutron source in EAST is in preparation. A new technology of digitized pulse signal processing is adopted in a wide dynamic range neutron flux monitor, compact recoil proton spectrometer, and time of flight spectrometer. Improvements will be made continuously to the system to achieve better adaptation to the EAST's harsh γ-ray and electro-magnetic radiation environment.
Experimental observations of driven and intrinsic rotation in tokamak plasmas
Rice, J. E.
Experimental observations of driven and intrinsic rotation in tokamak plasmas are reviewed. For momentum sources, there is direct drive from neutral beam injection, lower hybrid and ion cyclotron range of frequencies waves (including mode conversion flow drive), as well as indirect \\mathbf{j}× \\mathbf{B} forces from fast ion and electron orbit shifts, and toroidal magnetic field ripple loss. Counteracting rotation drive are sinks, such as from neutral drag and toroidal viscosity. Many of these observations are in agreement with the predictions of neo-classical theory while others are not, and some cases of intrinsic rotation remain puzzling. In contrast to particle and heat fluxes which depend on the relevant diffusivity and convection, there is an additional term in the momentum flux, the residual stress, which can act as the momentum source for intrinsic rotation. This term is independent of the velocity or its gradient, and its divergence constitutes an intrinsic torque. The residual stress, which ultimately responds to the underlying turbulence, depends on the confinement regime and is a complicated function of collisionality, plasma shape, and profiles of density, temperature, pressure and current density. This leads to the rich intrinsic rotation phenomenology. Future areas of study include integration of these many effects, advancement of quantitative explanations for intrinsic rotation and development of strategies for velocity profile control.
Tokamak experimental power reactor conceptual design. Volume I
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A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H/sub 2/O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters.
Comparison between 2D turbulence model ESEL and experimental data from AUG and COMPASS tokamaks
DEFF Research Database (Denmark)
Ondac, Peter; Horacek, Jan; Seidl, Jakub
In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtain...... for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath.......In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtained...... by reciprocating probe measurements from the two machines. Agreements were found in radial profiles of mean plasma potential and temperature, and in a level of density fluctuations. Disagreements, however, were found in the level of plasma potential and temperature fluctuations. This implicates a need...
Overview of experimental results on the HL-2A tokamak
Yan, L. W.; Duan, X. R.; Ding, X. T.; Dong, J. Q.; Yang, Q. W.; Liu, Yi; Zou, X. L.; Liu, D. Q.; Xuan, W. M.; Chen, L. Y.; Rao, J.; Song, X. M.; Huang, Y.; Mao, W. C.; Wang, Q. M.; Li, Q.; Cao, Z.; Li, B.; Cao, J. Y.; Lei, G. J.; Zhang, J. H.; Li, X. D.; Chen, W.; Cheng, J.; Cui, C. H.; Cui, Z. Y.; Deng, Z. C.; Dong, Y. B.; Feng, B. B.; Gao, Q. D.; Han, X. Y.; Hong, W. Y.; Huang, M.; Ji, X. Q.; Kang, Z. H.; Kong, D. F.; Lan, T.; Li, G. S.; Li, H. J.; Li, Qing; Li, W.; Li, Y. G.; Liu, A. D.; Liu, Z. T.; Luo, C. W.; Mao, X. H.; Pan, Y. D.; Peng, J. F.; Shi, Z. B.; Song, S. D.; Song, X. Y.; Sun, H. J.; Wang, A. K.; Wang, M. X.; Wang, Y. Q.; Xiao, W. W.; Xie, Y. F.; Yao, L. H.; Yao, L. Y.; Yu, D. L.; Yuan, B. S.; Zhao, K. J.; Zhong, G. W.; Zhou, J.; Zhou, Y.; Yan, J. C.; Yu, C. X.; Pan, C. H.; Liu, Yong; HL-2A Team
The physics experiments on the HL-2A tokamak have been focused on confinement improvement, particle and thermal transport, zonal flow and turbulence, filament characteristics, energetic particle induced modes and plasma fuelling efficiency since 2008. ELMy H-mode discharges are achieved in a lower density regime using a combination of NBI heating with ECRH. The power threshold is found to increase with a decrease in density, almost independent of the launching order of the ECRH and NBI heating power. The pedestal density profiles in the H-mode discharges are measured. The particle outward convection is observed during the pump-out transient phase with ECRH. The negative density perturbation (pump-out) is observed to propagate much faster than the positive one caused by out-gassing. The core electron thermal transport reduction triggered by far off-axis ECRH switch-off is investigated. The coexistence of low frequency zonal flow (LFZF) and geodesic acoustic mode (GAM) is observed. The dependence of the intensities of LFZFs and GAMs on the safety factor and ECRH power is identified. The 3D spatial structures of plasma filaments are measured in the boundary plasma and large-scale structures along a magnetic field line analysed for the first time. The beta-induced Alfvén eigenmodes (BAEs), excited by large magnetic islands (m-BAE) and by energetic electrons (e-BAE), are observed. The results for the study of fuelling efficiency and penetration characteristics of supersonic molecular beam injection (SMBI) are described.
Directory of Open Access Journals (Sweden)
Peter Ondac
Full Text Available In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtained by reciprocating probe measurements from the two machines. Agreements were found in radial profiles of mean plasma potential and temperature, and in a level of density fluctuations. Disagreements, however, were found in the level of plasma potential and temperature fluctuations. This implicates a need for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath.
First experimental results with the Current Limit Avoidance System at the JET tokamak
Energy Technology Data Exchange (ETDEWEB)
De Tommasi, G. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Galeani, S. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Jachmich, S. [Association EURATOM-Belgian State, Koninklijke Militaire School - Ecole Royale Militaire, B-1000 Brussels (Belgium); Joffrin, E. [IRFM-CEA, Centre de Cadarache, 13108 Saint-paul-lez-Durance (France); Lennholm, M. [EFDA Close Support Unit, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); European Commission, B-1049 Brussels (Belgium); Lomas, P.J. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Neto, A.C. [Associazione EURATOM-IST, Instituto de Plasmas e Fusao Nuclear, IST, 1049-001 Lisboa (Portugal); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Via Claudio 21, 80125 Napoli (Italy); McCullen, P. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Pironti, A. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Rimini, F.G. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Sips, A.C.C. [European Commission, B-1049 Brussels (Belgium); Varano, G.; Vitelli, R. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Zaccarian, L. [CNRS, LAAS, 7 Avenue du Colonel Roche, F-31400 Toulouse (France); Universitè de Toulouse, LAAS, F-31400 Toulouse (France)
The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.
Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report
Energy Technology Data Exchange (ETDEWEB)
Buncher, B.R.; Chi, J.W.H.; Fernandez, R.
This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended.
Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak
DEFF Research Database (Denmark)
Liu, S. C.; Guo, H. Y.; Xu, G. S.
Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected...... parameters are measured by reciprocating probes at the outer midplane, showing that the electron temperature and density increase but the parallel Mach number decreases significantly due to the gas injection. Effects of poloidal E × B drifts and parallel SOL flows on the divertor asymmetry observed in EAST...
Energy Technology Data Exchange (ETDEWEB)
OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, they have made significant progress in developing the building blocks needed for AT operation: (1) they have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. The authors have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiated power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet
Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma
Xu, Liqing; Zhang, Jizong; Chen, Kaiyun; Hu, Liqun; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin; Zhu, Yubao
Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey-predator model was found to reproduce the fishbone nonlinear process well.
Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor
Energy Technology Data Exchange (ETDEWEB)
Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.
As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield.
First results from solid state neutral particle analyzer on experimental advanced superconducting tokamak
Energy Technology Data Exchange (ETDEWEB)
Zhang, J. Z.; Zhao, J. L.; Wan, B. N.; Li, J. G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhu, Y. B., E-mail: firstname.lastname@example.com; Heidbrink, W. W. [Department of Physics and Astronomy, University of California, Irvine, California 92697 (United States)
Full function integrated, compact solid state neutral particle analyzers (ssNPA) based on absolute extreme ultraviolet silicon photodiode have been successfully implemented on the experimental advanced superconducting tokamak to measure energetic particle. The ssNPA system has been operated in advanced current mode with fast temporal and spatial resolution capabilities, with both active and passive charge exchange measurements. It is found that the ssNPA flux signals are increased substantially with neutral beam injection (NBI). The horizontal active array responds to modulated NBI beam promptly, while weaker change is presented on passive array. Compared to near-perpendicular beam, near-tangential beam brings more passive ssNPA flux and a broader profile, while no clear difference is observed on active ssNPA flux and its profile. Significantly enhanced intensities on some ssNPA channels have been observed during ion cyclotron resonant heating.
Studies for the ion cyclotron range of frequency heating in a tokamak fusion experimental device
Energy Technology Data Exchange (ETDEWEB)
Saigusa, Mikio [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
Ion cyclotron range of frequency heating has been investigated as an efficient additional plasma heating and non-inductive current driving methods in a tokamak type fusion experimental device. At first, an ICRF antenna coupling code was developed for the estimation of the coupling properties of phased antenna array, so that the ICRF antennas were designed for JT-60 and JT-60U ICRF heating systems using the coupling codes. The ICRF heating experiments had been performed in JT-60 and JT-60U. The coupling properties of ICRF antenna, the physics of peripheral plasma and energy confinement by ICRF heating in various heating regimes have been investigated. Next, the Toroidicity induced Alfven Eigen (TAE) mode have been studied using minority ICRF heating for producing energetic ions which can excite TAE mode. The TAE mode could be suppressed by current profile control using current ramp operation and lower hybrid current drive. (author) 74 refs.
Experimental and theoretical study of particle transport in the TCV Tokamak
Energy Technology Data Exchange (ETDEWEB)
Fable, E.
The main scope of this thesis work is to compare theoretical models with experimental observations on particle transport in particular regimes of plasma operation from the Tokamak à Configuration Variable (TCV) located at CRPP–EPFL in Lausanne. We introduce the main topics in Tokamak fusion research and the challenging problems in the first Chapter. A particular attention is devoted to the modelling of heat and particle transport. In the second Chapter the experimental part is presented, including an overview of TCV capabilities, a brief review of the relevant diagnostic systems, and a discussion of the numerical tools used to analyze the experimental data. In addition, the numerical codes that are used to interpret the experimental data and to compare them with theoretical predictions are introduced. The third Chapter deals with the problem of understanding the mechanisms that regulate the transport of energy in TCV plasmas, in particular in the electron Internal Transport Barrier (eITB) scenario. A radial transport code, integrated with an external module for the calculation of the turbulence-induced transport coefficients, is employed to reproduce the experimental scenario and to understand the physics at play. It is shown how the sustainment of an improved confinement regime is linked to the presence of a reversed safety factor profile. The improvement of confinement in the eITB regime is visible in the energy channel and in the particle channel as well. The density profile shows strong correlation with the temperature profile and has a large local logarithmic gradient. This is an important result obtained from the TCV eITB scenario analysis and is presented in the fourth Chapter. In the same chapter we present the estimate of the particle diffusion and convection coefficients obtained from density transient experiments performed in the eITB scenario. The theoretical understanding of the strong correlation between density and temperature observed in the e
Upgrades of the high resolution imaging x-ray crystal spectrometers on experimental advanced superconducting tokamak.
Lu, B; Wang, F; Shi, Y; Bitter, M; Hill, K W; Lee, S G; Fu, J; Li, Y; Wan, B
Two imaging x-ray crystal spectrometers, the so-called "poloidal" and "tangential" spectrometers, were recently implemented on experimental advanced superconducting tokamak (EAST) to provide spatially and temporally resolved impurity ion temperature (T(i)), electron temperature (T(e)) and rotation velocity profiles. They are derived from Doppler width of W line for Ti, the intensity ratio of Li-like satellites to W line for Te, and Doppler shift of W line for rotation. Each spectrometer originally consisted of a spherically curved crystal and a two-dimensional multi-wire proportional counter (MWPC) detector. Both spectrometers have now been upgraded. The layout of the tangential spectrometer was modified, since it had to be moved to a different port, and the spectrometer was equipped with two high count rate Pilatus detectors (Model 100 K) to overcome the count rate limitation of the MWPC and to improve its time resolution. The poloidal spectrometer was equipped with two spherically bent crystals to record the spectra of He-like and H-like argon simultaneously and side by side on the original MWPC. These upgrades are described, and new results from the latest EAST experimental campaign are presented.
Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak
Energy Technology Data Exchange (ETDEWEB)
Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Tritz, K. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)
A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.
Design study of toroidal magnets for tokamak experimental power reactors. [NbTi alloys
Energy Technology Data Exchange (ETDEWEB)
Stekly, Z.J.J.; Lucas, E.J. (eds.)
This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed.
Analysis of pedestal gradient characteristic on the Experimental Advanced Superconducting Tokamak
Wang, Teng Fei; Han, Xiao Feng; Zang, Qing; Xiao, Shu Mei; Tian, Bao Gang; Hu, Ai Lan; Zhao, Jun Yu
A pedestal database was built based on type I edge localized mode H-modes in the Experimental Advanced Superconducting Tokamak. The most common functional form hyperbolic tangent function (tanh) method is used to analyze pedestal characteristics. The pedestal gradient scales linearly with its pedestal top and the normalized pedestal pressure gradient α shows a strong correlation with electron collisionality. The connection among pedestal top value, gradient, and width is established with the normalized pedestal pressure gradient. In the core region of the plasma, the nature of the electron temperature stiffness reflects a proportionality between core and pedestal temperature while the increase proportion is lower than that expected in the high temperature region. However, temperature profile stiffness is limited or even disappears at the edge of the plasma, while the gradient length ratio ( ηe ) on the pedestal is important. The range of ηe is from 0.5 to 2, varying with the plasma parameters. The pedestal temperature brings a more significant impact on ηe than pedestal density.
MHD activity in the ISX-B tokamak: experimental results and theoretical interpretation
Energy Technology Data Exchange (ETDEWEB)
Carreras, B.A.; Dunlap, J.L.; Bell, J.D.; Charlton, L.A.; Cooper, W.A.; Dory, R.A.; Hender, T.C.; Hicks, H.R.; Holmes, J.A.; Lynch, V.E.
The observed spectrum of MHD fluctuations in the ISX-B tokamak is clearly dominated by the n=1 mode when the q=1 surface is in the plasma. This fact agrees well with theoretical predictions based on 3-D resistive MHD calculations. They show that the (m=1; n=1) mode is then the dominant instability. It drives other n=1 modes through toroidal coupling and n>1 modes through nonlinear couplings. These theoretically predicted mode structures have been compared in detail with the experimentally measured wave forms (using arrays of soft x-ray detectors). The agreement is excellent. More detailed comparisons between theory and experiment have required careful reconstructions of the ISX-B equilibria. The equilibria so constructed have permitted a precise evaluation of the ideal MHD stability properties of ISX-B. The present results indicate that the high ..beta.. ISX-B equilibria are marginally stable to finite eta ideal MHD modes. The resistive MHD calculations also show that at finite ..beta.. there are unstable resistive pressure driven modes.
Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report
Energy Technology Data Exchange (ETDEWEB)
This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn/sub 3/ conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended.
Development and experimental evaluation of theoretical models for ion cyclotron resonance frequency heating of tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Mantsinen, M. [Helsinki Univ. of Technology, Espoo (Finland). Dept. of Technical Physics
Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in
New dual gas puff imaging system with up-down symmetry on experimental advanced superconducting tokamak
DEFF Research Database (Denmark)
Liu, S. C.; Shao, L. M.; Zweben, S. J.
advanced superconducting tokamak (EAST). The two views are up-down symmetric about the midplane and separated by a toroidal angle of 66.6 degrees. A linear manifold with 16 holes apart by 10 mm is used to form helium gas cloud at the 130x130 mm (radial versus poloidal) objective plane. A fast camera...
First results obtained from the soft x-ray pulse height analyzer on experimental advanced superconducting tokamak.
Xu, P; Lin, S Y; Hu, L Q; Duan, Y M; Zhang, J Z; Chen, K Y; Zhong, G Q
An assembly of soft x-ray pulse height analyzer system, based on silicon drift detector (SDD), has been successfully established on the experimental advanced superconducting tokamak (EAST) to measure the spectrum of soft x-ray emission (E=1-20 keV). The system, including one 15-channel SDD linear array, is installed on EAST horizontal port C. The time-resolved radial profiles of electron temperature and K(alpha) intensities of metallic impurities have been obtained with a spatial resolution of around 7 cm during a single discharge. It was found that the electron temperatures derived from the system are in good agreement with the values from Thomson scattering measurements. The system can also be applied to the measurement of the long pulse discharge for EAST. The diagnostic system is introduced and some typical experimental results obtained from the system are also presented.
Outward particle transport by coherent mode in the H-mode pedestal in the Experimental Advanced Superconducting Tokamak (EAST)
Zhang, T.; Han, X.; Gao, X.; Liu, H. Q.; Shi, T. H.; Liu, J. B.; Liu, Y.; Kong, D. F.; Liu, Z. X.; Qu, H.; Xiang, H. M.; Geng, K. N.; Wang, Y. M.; Wen, F.; Zhang, S. B.; Ling, B. L.; the EAST Team
A coherent mode (CM) in the edge pedestal region has been observed on different fluctuation quantities, including density fluctuation, electron temperature fluctuation and magnetic fluctuation in H mode plasma on the Experimental Advanced Superconducting Tokamak (EAST) tokamak. Measurements at different poloidal positions show that the local poloidal wavenumber is smallest at the outboard midplane and will increase with poloidal angle. This poloidal asymmetry is consistent with the flute-like assumption (i.e. k// ˜ 0) from which the toroidal mode number of the mode has been estimated as between 12 and 17. It was further found that the density fluctuation amplitude of the CM also demonstrated poloidal asymmetry. The appearance of a CM can clearly decrease or even stop the increase in the edge density, while the disappearance of a CM will lead to an increase in the pedestal density and density gradient. Statistical analysis showed there was a trend that as the CM mode amplitude increased, the rate of increase of the edge density decreased and the particle flux (Γdiv) onto the divertor plate increased. The CM sometimes showed burst behavior, and these bursts led bursts on Γdiv with a time of about 230 μs, which is close to the time for particle flow from the outer midplane to the divertor targets along the scrape-off layer magnetic field line. This evidence showed that the CM had an effect on the outward transport of particles.
Experimental and theoretical study of quasicoherent fluctuations in enhanced D(alpha) plasmas in the Alcator C-Mod tokamak.
Mazurenko, A; Porkolab, M; Mossessian, D; Snipes, J A; Xu, X Q; Nevins, W M
A comparison of experimental measurements and theoretical studies of the quasicoherent (QC) mode, observed at high densities during enhanced D(alpha) (EDA) H mode in the Alcator C-Mod tokamak, are reported. The QC mode is a high frequency ( approximately 100 kHz) nearly sinusoidal fluctuation in density and magnetic field, localized in the steep density gradient ("pedestal") at the plasma edge, with typical wave numbers k(R) approximately 3-6 cm(-1), k(theta) approximately 1.3 cm(-1) (midplane). It is proposed here that the QC mode is a form of resistive ballooning mode known as the resistive X-point mode, in reasonable agreement with predictions by the BOUT (boundary-plasma turbulence) code.
Observations of compound sawteeth in ion cyclotron resonant heating plasma using ECE imaging on experimental advanced superconducting tokamak
Hussain, Azam; Zhao, Zhenling; Xie, Jinlin; Zhu, Ping; Liu, Wandong; Ti, Ang
The spatial and temporal evolutions of compound sawteeth were directly observed using 2D electron cyclotron emission imaging on experimental advanced superconducting tokamak. The compound sawtooth consists of partial and full collapses. After partial collapse, the hot core survives as only a small amount of heat disperses outwards, whereas in the following full collapse a large amount of heat is released and the hot core dissipates. The presence of two q = 1 surfaces was not observed. Instead, the compound sawtooth occurs mainly at the beginning of an ion cyclotron resonant frequency heating pulse and during the L-H transition phase, which may be related to heat transport suppression caused by a decrease in electron heat diffusivity.
Design and characterization of a 32-channel heterodyne radiometer for electron cyclotron emission measurements on experimental advanced superconducting tokamak.
Han, X; Liu, X; Liu, Y; Domier, C W; Luhmann, N C; Li, E Z; Hu, L Q; Gao, X
A 32-channel heterodyne radiometer has been developed for the measurement of electron cyclotron emission (ECE) on the experimental advanced superconducting tokamak (EAST). This system collects X-mode ECE radiation spanning a frequency range of 104-168 GHz, where the frequency coverage corresponds to a full radial coverage for the case with a toroidal magnetic field of 2.3 T. The frequency range is equally spaced every 2 GHz from 105.1 to 167.1 GHz with an RF bandwidth of ~500 MHz and the video bandwidth can be switched among 50, 100, 200, and 400 kHz. Design objectives and characterization of the system are presented in this paper. Preliminary results for plasma operation are also presented.
Experimental particle physics research at Texas Tech University
Energy Technology Data Exchange (ETDEWEB)
Akchurin, Nural [Texas Tech Univ., Lubbock, TX (United States); Lee, Sung-Won [Texas Tech Univ., Lubbock, TX (United States); Volobouev, Igor [Texas Tech Univ., Lubbock, TX (United States); Wigmans, Richard [Texas Tech Univ., Lubbock, TX (United States)
The high energy physics group at Texas Tech University (TTU) concentrates its research efforts on the Compact Muon Solenoid (CMS) experiment at the Large Hadron Collider (LHC) and on generic detector R&D for future applications. Our research programs have been continuously supported by the US Department of Energy for over two decades, and this final report summarizes our achievements during the last grant period from May 1, 2012 to March 31, 2016. After having completed the Run 1 data analyses from the CMS detector, including the discovery of the Higgs boson in July 2012, we concentrated on commissioning the CMS hadron calorimeter (HCAL) for Run 2, performing analyses of Run 2 data, and making initial studies and plans for the second phase of upgrades in CMS. Our research has primarily focused on searches for Beyond Standard Model (BSM) physics via dijets, monophotons, and monojets. We also made significant contributions to the analyses of the semileptonic Higgs decays and Standard Model (SM) measurements in Run 1. Our work on the operations of the CMS detector, especially the performance monitoring of the HCAL in Run 1, was indispensable to the experiment. Our team members, holding leadership positions in HCAL, have played key roles in the R&D, construction, and commissioning of these detectors in the last decade. We also maintained an active program in jet studies that builds on our expertise in calorimetry and algorithm development. In Run 2, we extended some of our analyses at 8 TeV to 13 TeV, and we also started to investigate new territory, e.g., dark matter searches with unexplored signatures. The objective of dual-readout calorimetry R&D was intended to explore (and, if possible, eliminate) the obstacles that prevent calorimetric detection of hadrons and jets with a comparable level of precision as we have grown accustomed to for electrons and photons. The initial prototype detector was successfully tested at the SPS/CERN in 2003-2004 and evolved over the
Development of a high-speed vacuum ultraviolet (VUV) imaging system for the Experimental Advanced Superconducting Tokamak
Zhou, Fan; Ming, Tingfeng; Wang, Yumin; Wang, Zhijun; Long, Feifei; Zhuang, Qing; Li, Guoqiang; Liang, Yunfeng; Gao, Xiang
A high-speed vacuum ultraviolet (VUV) imaging system for edge plasma studies is being developed on the Experimental Advanced Superconducting Tokamak (EAST). Its key optics is composed of an inverse type of Schwarzschild telescope made of a set of Mo/Si multilayer mirrors, a micro-channel plate (MCP) equipped with a P47 phosphor screen and a high-speed camera with CMOS sensors. In order to remove the contribution from low-energy photons, a Zr filter is installed in front of the MCP detector. With this optics, VUV photons with a wavelength of 13.5 nm, which mainly come from the line emission from intrinsic carbon (C vi: n = 4-2 transition) or the Ly-α line emission from injected Li iii on the EAST, can be selectively measured two-dimensionally with both high temporal and spatial resolutions. At present, this system is installed to view the plasma from the low field side in a horizontal port in the EAST. It has been operated routinely during the 2016 EAST experiment campaign, and the first result is shown in this work. To roughly evaluate the system performance, synthetic images are created. And it indicates that this system mainly measures the edge localized emissions by comparing the synthetic images and experimental data.
High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor
Energy Technology Data Exchange (ETDEWEB)
Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)
This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)
Experimental studies and modelling of high radiation and high density plasmas in the ASDEX upgrade tokamak
Energy Technology Data Exchange (ETDEWEB)
Casali, Livia
Fusion plasmas contain impurities, either intrinsic originating from the wall, or injected willfully with the aim of reducing power loads on machine components by converting heat flux into radiation. The understanding and the prediction of the effects of these impurities and their radiation on plasma performances is crucial in order to retain good confinement. In addition, it is important to understand the impact of pellet injection on plasma performance since this technique allows higher core densities which are required to maximise the fusion power. This thesis contributes to these efforts through both experimental investigations and modelling. Experiments were conducted at ASDEX Upgrade which has a full-W wall. Impurity seeding was applied to H-modes by injecting nitrogen and also medium-Z impurities such as Kr and Ar to assess the impact of both edge and central radiation on confinement. A database of about 25 discharges has been collected and analysed. A wide range of plasma parameters was achieved up to ITER relevant values such as high Greenwald and high radiation fractions. Transport analyses taking into account the radiation distribution reveal that edge localised radiation losses do not significantly impact confinement as long as the H-mode pedestal is sustained. N seeding induces higher pedestal pressure which is propagated to the core via profile stiffness. Central radiation must be limited and controlled to avoid confinement degradation. This requires reliable control of the impurity concentration but also possibilities to act on the ELM frequency which must be kept high enough to avoid an irreversible impurity accumulation in the centre and the consequent radiation collapse. The key role of the f{sub ELM} is confirmed also by the analysis of N+He discharges. Non-coronal effects affect the radiation of low-Z impurities at the plasma edge. Due to the radial transport, the steep temperature gradients and the ELM flush out, a local equilibrium cannot be
Advanced Laboratory at Texas State University: Error Analysis, Experimental Design, and Research Experience for Undergraduates
Ventrice, Carl
Physics is an experimental science. In other words, all physical laws are based on experimentally observable phenomena. Therefore, it is important that all physics students have an understanding of the limitations of certain experimental techniques and the associated errors associated with a particular measurement. The students in the Advanced Laboratory class at Texas State perform three detailed laboratory experiments during the semester and give an oral presentation at the end of the semester on a scientific topic of their choosing. The laboratory reports are written in the format of a ``Physical Review'' journal article. The experiments are chosen to give the students a detailed background in error analysis and experimental design. For instance, the first experiment performed in the spring 2009 semester is entitled Measurement of the local acceleration due to gravity in the RFM Technology and Physics Building. The goal of this experiment is to design and construct an instrument that is to be used to measure the local gravitational field in the Physics Building to an accuracy of ±0.005 m/s^2. In addition, at least one of the experiments chosen each semester involves the use of the research facilities within the physics department (e.g., microfabrication clean room, surface science lab, thin films lab, etc.), which gives the students experience working in a research environment.
Experimental studies of toroidal correlations of plasma density fluctuations along the magnetic field lines in the T-10 tokamak and first results of numerical modeling
Buldakov, M. A.; Vershkov, V. A.; Isaev, M. Yu; Shelukhin, D. A.
The antenna system of reflectometry diagnostics at the T-10 tokamak allows to study long-range toroidal correlations of plasma density fluctuations along the magnetic field lines. The antenna systems are installed in two poloidal cross-sections of the vacuum chamber separated by a 90° angle in the toroidal direction. The experiments, which were conducted at the low field side, showed that the high level of toroidal correlations is observed only for quasi-coherent fluctuations. However, broadband and stochastic low frequency fluctuations are not correlated. Numerical modeling of the plasma turbulence structure in the T-10 tokamak was conducted to interpret the experimental results and take into account non-locality of reflectometry measurements. In the model used, it was assumed that the magnitudes of density fluctuations are constant along the magnetic field lines. The 2D full-wave Tamic-RTH code was used to model the reflectometry signals. High level of correlations for quasi-coherent fluctuations was obtained during the modeling, which agrees with the experimental observations. However, the performed modeling also predicts high level of correlations for broadband fluctuations, which contradicts the experimental data. The modeling showed that the effective reflection radius, from which the information on quasi-coherent plasma turbulence is obtained, is shifted outwards from the reflection radius by approximately 7 mm.
A fast-time-response extreme ultraviolet spectrometer for measurement of impurity line emissions in the Experimental Advanced Superconducting Tokamak.
Zhang, Ling; Morita, Shigeru; Xu, Zong; Wu, Zhenwei; Zhang, Pengfei; Wu, Chengrui; Gao, Wei; Ohishi, Tetsutarou; Goto, Motoshi; Shen, Junsong; Chen, Yingjie; Liu, Xiang; Wang, Yumin; Dong, Chunfeng; Zhang, Hongmin; Huang, Xianli; Gong, Xianzu; Hu, Liqun; Chen, Junlin; Zhang, Xiaodong; Wan, Baonian; Li, Jiangang
A flat-field extreme ultraviolet (EUV) spectrometer working in the 20-500 Å wavelength range with fast time response has been newly developed to measure line emissions from highly ionized tungsten in the Experimental Advanced Superconducting Tokamak (EAST) with a tungsten divertor, while the monitoring of light and medium impurities is also an aim in the present development. A flat-field focal plane for spectral image detection is made by a laminar-type varied-line-spacing concave holographic grating with an angle of incidence of 87°. A back-illuminated charge-coupled device (CCD) with a total size of 26.6 × 6.6 mm(2) and pixel numbers of 1024 × 255 (26 × 26 μm(2)/pixel) is used for recording the focal image of spectral lines. An excellent spectral resolution of Δλ0 = 3-4 pixels, where Δλ0 is defined as full width at the foot position of a spectral line, is obtained at the 80-400 Å wavelength range after careful adjustment of the grating and CCD positions. The high signal readout rate of the CCD can improve the temporal resolution of time-resolved spectra when the CCD is operated in the full vertical binning mode. It is usually operated at 5 ms per frame. If the vertical size of the CCD is reduced with a narrow slit, the time response becomes faster. The high-time response in the spectral measurement therefore makes possible a variety of spectroscopic studies, e.g., impurity behavior in long pulse discharges with edge-localized mode bursts. An absolute intensity calibration of the EUV spectrometer is also carried out with a technique using the EUV bremsstrahlung continuum at 20-150 Å for quantitative data analysis. Thus, the high-time resolution tungsten spectra have been successfully observed with good spectral resolution using the present EUV spectrometer system. Typical tungsten spectra in the EUV wavelength range observed from EAST discharges are presented with absolute intensity and spectral identification.
Experimental measurement of magnetic field null in the vacuum chamber of KTM tokamak based on matrix of 2D Hall sensors
Energy Technology Data Exchange (ETDEWEB)
Shapovalov, G.; Chektybayev, B., E-mail: firstname.lastname@example.com; Sadykov, A.; Skakov, M.; Kupishev, E.
Experimental technique of measurement of magnetic field null region inside of the KTM tokamak vacuum chamber has been developed. Square matrix of 36 2D Hall sensors, which used in the technique, allows carrying out direct measurements of poloidal magnetic field dynamics in the vacuum chamber. To better measuring accuracy, Hall sensor's matrix was calibrated with commercial Helmholtz coils and in situ measurement of defined magnetic field from poloidal and toroidal coils. Standard KTM Data-Acquisition System has been used to collect data from Hall sensors. Experimental results of measurement of magnetic field null in the vacuum chamber of KTM are shown in the paper. Additionally results of the magnetic field null reconstruction from signals of inductive total flux loops are shown in the paper.
Radiation-driven m = 2 island formation and dynamics near density limit in experimental advanced superconducting tokamak ohmic plasma
Xu, Liqing; Duan, Yanmin; Chen, Kaiyun; Zhao, Hailin; Luo, Zhenping; Zheng, Zhen; Liu, Yong; Liu, Haiqing; Chen, Yingjie; Yi, Yuan; Hu, Liqun; Du, Hongfei; Shi, Tonghui
A radiation-driven m = 2 island was observed in the experimental advanced superconducting tokamak (EAST) ohmic plasma, near the density limit. The mode onset occurs when the the ohmic heating input is less than the radiative cooling loss, which agrees with the mode onset behavior of the thermo-resistive model. The evolution of the equilibrium during the mode process was obtained using the ONETWO transport code, with inputs comprising the experimental electron temperature and density profiles. A large m = 2 island can drive an m = 1 sideband mode, which leads to an internal crash that appears as a large change in temperature that occurs not only in the q = 2 region but also in the core.
Experimental observations of mode-converted ion cyclotron waves in a tokamak plasma by phase contrast imaging.
Nelson-Melby, E; Porkolab, M; Bonoli, P T; Lin, Y; Mazurenko, A; Wukitch, S J
The process of mode conversion, whereby an externally launched electromagnetic wave converts into a shorter wavelength mode(s) in a thermal plasma near a resonance in the index of refraction, is particularly important in a multi-ion species plasma near the ion cyclotron frequency. Using phase contrast imaging techniques (PCI), mode-converted electromagnetic ion cyclotron waves have been detected for the first time in the Alcator C-Mod tokamak near the H-3He ion-ion hybrid resonance region during high power rf heating experiments. The results agree with theoretical predictions.
Experimental study of the MHD activity associated to the mode m=2, n=1 in the Tore Supra tokamak; Etude experimentale de l`activite MHD associee au mode m=2, n=1 dans le tokamak Tore Supra
Energy Technology Data Exchange (ETDEWEB)
Turlur, S.
In tokamaks such as Tore Supra, the plasma confinement magnetic structure can be severely affected when Magnetohydrodynamic (M.H.D.) instabilities are destabilized. Experimentally, these instabilities are detected as magnetic fluctuations with captors located against the inner wall of the vacuum vessel. Fourier analysis provides amplitude, frequency and wave numbers of magnetic modes. In case of fast or transient phenomena, the analysis of magnetic fluctuations is completed using the singular value decomposition. In this dissertation, these analysis techniques are used to study two specific examples of M.H.D. activity related to the m = 2, n = 1 mode. On Tore Supra, the onset of this mode have strong consequences on the stability of partially or fully non inductive discharges. A regular and persistent sawtooth-like regime is observed on the electronic temperature leading to a significant degradation of the central confinement. Heat exhaust and particle balance are also essential parameters to achieve stationary discharges. On Tore Supra, these are studied with the ergodic divertor which produces stochastic magnetic field lines at the plasma edge. For optimal operating conditions of the ergodic divertor, the growth of the m = 2, N = 1 mode can lead to sudden destruction of magnetic equilibrium. For both cases, understanding and characterization of mechanisms leading to the observed m = 2, n = 1 M.H.D. activity are fundamental to obtain stationary discharges. (author). 115 refs.
Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra
Energy Technology Data Exchange (ETDEWEB)
Costanzo, L
The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a
Experimental measurements of energy transfer and nonlinear interaction in turbulence at the sino-united spherical tokamak
Chai, Song; Xu, Yuhong; Gao, Zhe; Wang, Wenhao; Liu, Yangqing; Tan, Yi
The characteristics of the energy transfer and nonlinear coupling among edge electromagnetic turbulence have been dedicatedly studied in various discharge stages at the sino-united spherical tokamak using multiple Langmuir and magnetic probe arrays. The wavelet bispectral analysis and the modified Kim's method are applied to investigate turbulence properties and their linear growth/damping and nonlinear energy transfer rates, along with multi-field turbulence interactions. The results show diverse features in the linear growth and nonlinear energy transfer between multi-field fluctuations during the current ramp-up, stationary, and internal connection event discharge phases. The diversity implies the importance to develop more sophisticated multi-field models to directly estimate the energy transfer rate among multiple turbulent fields.
Experimental study of reversed shear Alfven eigenmodes during the current ramp in the Alcator C-Mod tokamak
Energy Technology Data Exchange (ETDEWEB)
Edlund, E M; Kramer, G J [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Porkolab, M; Lin, Y; Tsujii, N; Wukitch, S J [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Lin, L, E-mail: email@example.com [University of California Los Angeles, Los Angeles, CA 90095 (United States)
Experiments conducted in the Alcator C-Mod tokamak have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs during the current ramp provides a constraint on the evolution of q{sub min}, a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive bounds on the adiabatic index, a measure of the plasma compressibility. This scaling places the adiabatic index at 1.40 {+-} 0.15 and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7/4.
Experimental Study of Reversed Shear Alfven Eigenmodes During The Current Ramp In The Alcator C-Mod Tokamak
Energy Technology Data Exchange (ETDEWEB)
Edlund, E. M.; Porkolab, M.; Kramer, G. J.; Lin, L.; Lin, Y.; Tsuji, N.; Wukitch, S. J.
Experiments conducted in the Alcator C-Mod tokamak at MIT have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs throughout the current ramp provides a constraint on the evolution of qmin, a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive a measure of the adiabatic index, a measure of the plasma compressibility. This scaling bounds the adiabatic index at 1.40 ± 0:15 used in MHD models and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7~4.
Fabrication and Characterization of Samples for a Material Migration Experiment on the Experimental Advanced Superconducting Tokamak (EAST).
Energy Technology Data Exchange (ETDEWEB)
Wampler, William R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Van Deusen, Stuart B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
This report documents work done for the ITER International Fusion Energy Organization (Sponsor) under a Funds-In Agreement FI 011140916 with Sandia National Laboratories. The work consists of preparing and analyzing samples for an experiment to measure material erosion and deposition in the EAST Tokamak. Sample preparation consisted of depositing thin films of carbon and aluminum onto molybdenum tiles. Analysis consists of measuring the thickness of films before and after exposure to helium plasma in EAST. From these measurements the net erosion and deposition of material will be quantified. Film thickness measurements are made at the Sandia Ion Beam Laboratory using Rutherford backscattering spectrometry and nuclear reaction analysis, as described in this report. This report describes the film deposition and pre-exposure analysis. Results from analysis after plasma exposure will be given in a subsequent report.
Degraded Confinement in Tokamak Experiments
Schüller, F. C.
After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional heating will be given. Some examples of improved confinement modes will be discussed. Fluctuation
An enhanced tokamak startup model
Goswami, Rajiv; Artaud, Jean-François
The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.
Prospects for Tokamak Fusion Reactors
Energy Technology Data Exchange (ETDEWEB)
Sheffield, J.; Galambos, J.
This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.
Advanced Tokamak Scenarios for the FIRE Burning Plasma Experiment
Energy Technology Data Exchange (ETDEWEB)
C.E. Kessel; D. Ignat; T.K. Mau
The advanced tokamak (AT) capability of the Fusion Ignition Research Experiment (FIRE) burning plasma experiment is examined with 0-D systems analysis, equilibrium and ideal-MHD stability, radio-frequency current-drive analysis, and full discharge dynamic simulations. These analyses have identified the required parameters for attractive burning advanced tokamak plasmas, and indicate that these are feasible with the present progress on existing experimental tokamaks.
Song, Yuntao; Du, Shijun
Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study
Tokamak plasma self-organization-synergetics of magnetic trap plasmas
Razumova, K. A.; Andreev, V. F.; Eliseev, L. G.; Kislov, A. Y.; La Haye, R. J.; Lysenko, S. E.; Melnikov, A. V.; Notkin, G. E.; Pavlov, Y. D.; Kantor, M. Y.
Analysis of a wide range of experimental results in plasma magnetic confinement investigations shows that in most cases, plasmas are self-organized. In the tokamak case, it is realized in the self-consistent pressure profile, which permits the tokamak plasma to be macroscopically MHD stable.
Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor.
Singh, M J; De Esch, H P L
This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.
First observation of a new zonal-flow cycle state in the H-mode transport barrier of the experimental advanced superconducting Tokamak
DEFF Research Database (Denmark)
Xu, G.S.; Wang, H. Q.; Wan, B. N.
A new turbulence-flow cycle state has been discovered after the formation of a transport barrier in the H-mode plasma edge during a quiescent phase on the EAST superconducting tokamak. Zonal-flow modulation of high-frequency-broadband (0.05-1MHz) turbulence was observed in the steep-gradient regi...
Degraded confinement and turbulence in tokamak experiments
Schüller, F. C.
After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional heating will be given. Some examples of improved confinement modes will be discussed. Fluctuation
Degraded confinement and turbulence in tokamak experiments
Hogeweij, G. M. D.
After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional healing will be given. Some examples of improved confinement; modes will be discussed.
Advanced commercial tokamak study
Energy Technology Data Exchange (ETDEWEB)
Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.
Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs.
Characterization of the Tokamak Novillo in cleaning regime; Caracterizacion del Tokamak Novillo en regimen de limpieza
Energy Technology Data Exchange (ETDEWEB)
Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E
In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip{sub t} like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I{sub (p)}t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)
Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)
Azizov, E. A.
The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.
UCLA Tokamak Program Close Out Report.
Energy Technology Data Exchange (ETDEWEB)
Taylor, Robert John [UCLA/retired
The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.
Spontaneous generation of rotation in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Parra Diaz, Felix [Oxford University
Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.
Benchmarking Tokamak edge modelling codes
Contributors To The Efda-Jet Work Programme; Coster, D. P.; Bonnin, X.; Corrigan, G.; Kirnev, G. S.; Matthews, G.; Spence, J.; Contributors to the EFDA-JET work programme
Tokamak edge modelling codes are in widespread use to interpret and understand existing experiments, and to make predictions for future machines. Little direct benchmarking has been done between the codes, and the users of the codes have tended to concentrate on different experimental machines. An important validation step is to compare the codes for identical scenarios. In this paper, two of the major edge codes, SOLPS (B2.5-Eirene) and EDGE2D-NIMBUS are benchmarked against each other. A set of boundary conditions, transport coefficients, etc. for a JET plasma were chosen, and the two codes were run on the same grid. Initially, large differences were seen in the resulting plasmas. These differences were traced to differing physics assumptions with respect to the parallel heat flux limits. Once these were switched off in SOLPS, or implemented and switched on in EDGE2D-NIMBUS, the remaining differences were small.
Effect of limiter currents on plasma equilibrium and stability in a tokamak
Belashov, V. I.; Gribov, Yu. V.; Putvinskij, S. V.; Brevnov, N. N.
The results of theoretical and experimental research of currents between diaphragms limiting plasma cord in tokamak on plasma equilibrium and stability with an arbitrary form of transverse cross section are presented. It is shown that plasma cord behaviour depends on applied voltage polarity. The phenomena considered can be important for tokamaks in which fast plasma compression in a big radius is invisaged.
Tokamak foundation in USSR/Russia 1950-1990
Smirnov, V. P.
In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.
Basic Physics of Tokamak Transport Final Technical Report.
Energy Technology Data Exchange (ETDEWEB)
Sen, Amiya K.
The goal of this grant has been to study the basic physics of various sources of anomalous transport in tokamaks. Anomalous transport in tokamaks continues to be one of the major problems in magnetic fusion research. As a tokamak is not a physics device by design, direct experimental observation and identification of the instabilities responsible for transport, as well as physics studies of the transport in tokamaks, have been difficult and of limited value. It is noted that direct experimental observation, identification and physics study of microinstabilities including ITG, ETG, and trapped electron/ion modes in tokamaks has been very difficult and nearly impossible. The primary reasons are co-existence of many instabilities, their broadband fluctuation spectra, lack of flexibility for parameter scans and absence of good local diagnostics. This has motivated us to study the suspected tokamak instabilities and their transport consequences in a simpler, steady state Columbia Linear Machine (CLM) with collisionless plasma and the flexibility of wide parameter variations. Earlier work as part of this grant was focused on both ITG turbulence, widely believed to be a primary source of ion thermal transport in tokamaks, and the effects of isotope scaling on transport levels. Prior work from our research team has produced and definitively identified both the slab and toroidal branches of this instability and determined the physics criteria for their existence. All the experimentally observed linear physics corroborate well with theoretical predictions. However, one of the large areas of research dealt with turbulent transport results that indicate some significant differences between our experimental results and most theoretical predictions. Latter years of this proposal were focused on anomalous electron transport with a special focus on ETG. There are several advanced tokamak scenarios with internal transport barriers (ITB), when the ion transport is reduced to
Vertical compact torus injection into the STOR-M tokamak
Liu, Dazhi
Central fuelling is a fundamental issue in the next generation tokamak-ITER (International Thermonuclear Experimental Reactor). It is essential for optimization of the bootstrap current which is proportional to the pressure gradient of trapped particles. The conventional fusion reactor fuelling techniques, such as gas puffing and cryogenic pellet injection, are considered inadequate to fulfill this goal due to premature ionization caused by high plasma temperature and density. Compact Torus (CT) injection is a promising fuelling technique for central fuelling a reactor-grade tokamak. An accelerated CT is expected to penetrate into the core region and deposit fuel there provided the CT kinetic energy density exceeds the magnetic energy density in a target plasma. This process is complicated and involves CT penetration into an external magnetic field, a CT stopping mechanism, magnetic reconnection, and excitation of plasma waves. CTs can be injected at different angles with respect to the tokamak toroidal magnetic field, either horizontally or vertically. Normally, CTs are injected radially in the mid-plane of a tokamak. In this configuration, CTs will undergo a decelerating force due to the gradient of the tokamak toroidal magnetic field. CTs will stop inside the tokamak chamber or bunce back depending on the relation between kinetic energy density of injected CTs and the tokamak toroidal magnetic field energy density. In the case of vertical injection, deeper penetration is expected due to the absence of the gradient of the tokamak toroidal field in that direction. Experimental investigations on vertical CT injection into a tokamak will be of great significance. The aim of this thesis is to experimentally investigate the feasibility of vertical CT injection into a tokamak and effects of CTs on tokamak plasma confinements. The Saskatchewan Torus-Modified (STOR-M) tokamak is currently the only tokamak equipped with a CT injector in the world. Vertical CT injection
Particle transport in tokamak plasmas, theory and experiment
Energy Technology Data Exchange (ETDEWEB)
Angioni, C [Max-Planck Institut fuer Plasmaphysik, IPP-EURATOM Association, D-85748 Garching (Germany); Fable, E; Maslov, M; Weisen, H [Centre de Recherches en Physique des Plasmas, Association EURATOM-Confederation Suisse, EPFL, 1015 Lausanne (Switzerland); Greenwald, M [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA (United States); Peeters, A G [Centre for Fusion, Space and Astrophysics, University of Warwick, CV4 7AL, Coventry (United Kingdom); Takenaga, H [Japan Atomic Energy Agency, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)
The physical processes producing electron particle transport in the core of tokamak plasmas are described. Starting from the gyrokinetic equation, a simple analytical derivation is used as guidance to illustrate the main mechanisms driving turbulent particle convection. A review of the experimental observations on particle transport in tokamaks is presented and the consistency with the theoretical predictions is discussed. An overall qualitative agreement, and in some cases even a specific quantitative agreement, emerges between complex theoretical predictions and equally complex experimental observations, exhibiting different dependences on plasma parameters under different regimes. By these results, the direct connection between macroscopic transport properties and the character of microscopic turbulence is pointed out, and an important confirmation of the paradigm of microinstabilities and turbulence as the main cause of transport in the core of tokamaks is obtained. Finally, the impact of these results on the prediction of the peaking of the electron density profile in a fusion reactor is illustrated.
A control approach for plasma density in tokamak machines
Energy Technology Data Exchange (ETDEWEB)
Boncagni, Luca, E-mail: email@example.com [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)
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].
DEFF Research Database (Denmark)
Kjellberg, Kurt
Indtryk fra et besøg på SLA, Special Libraries Associations årlige konference, San Antonio, Texas, USA, 9.-13. juni 2001. "An Information Odyssey: Seizing the Competitive Advantage"......Indtryk fra et besøg på SLA, Special Libraries Associations årlige konference, San Antonio, Texas, USA, 9.-13. juni 2001. "An Information Odyssey: Seizing the Competitive Advantage"...
Measurement of electron density profile by microwave reflectometry on tokamaks
Energy Technology Data Exchange (ETDEWEB)
Simonet, F.
A new method for measuring the electron density spatial profile has been successfully tested on the tokamak of Fontenay aux Roses (TFR). This method is based on the total reflection experienced by a wave of frequency F on the layer where F = F/sub p/e(r). The experimental results show that the maximum electron density in the discharge is also easily measured and that accurate determination of a density profile can be obtained with a time resolution of 5 ms. This diagnostic is well adapted to all fusion devices where access to the total plasma cross section is limited, particularly for large tokamaks.
Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...
Spherical tokamak development in Brazil
Energy Technology Data Exchange (ETDEWEB)
Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
The ETE spherical Tokamak project
Energy Technology Data Exchange (ETDEWEB)
Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] [and others]. E-mail: email@example.com
This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)
Spherical tokamak development in Brazil
Energy Technology Data Exchange (ETDEWEB)
Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)
The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
The ARIES tokamak reactor study
Energy Technology Data Exchange (ETDEWEB)
The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.
Bibliography of fusion product physics in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Hively, L. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sigmar, D. J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category.
Moving Divertor Plates in a Tokamak
Energy Technology Data Exchange (ETDEWEB)
S.J. Zweben, H. Zhang
Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.
Fusion potential for spherical and compact tokamaks
Energy Technology Data Exchange (ETDEWEB)
Sandzelius, Mikael
The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.
Response of plasma rotation to resonant magnetic perturbations in J-TEXT tokamak
Yan, W.; Chen, Z. Y.; Huang, D. W.; Hu, Q. M.; Shi, Y. J.; Ding, Y. H.; Cheng, Z. F.; Yang, Z. J.; Pan, X. M.; Lee, S. G.; Tong, R. H.; Wei, Y. N.; Dong, Y. B.; J-TEXT Team
The response of plasma toroidal rotation to the external resonant magnetic perturbations (RMP) has been investigated in Joint Texas Experimental Tokamak (J-TEXT) ohmic heating plasmas. For the J-TEXT's plasmas without the application of RMP, the core toroidal rotation is in the counter-current direction while the edge rotation is near zero or slightly in the co-current direction. Both static RMP experiments and rotating RMP experiments have been applied to investigate the plasma toroidal rotation. The core toroidal rotation decreases to lower level with static RMP. At the same time, the edge rotation can spin to more than 20 km s‑1 in co-current direction. On the other hand, the core plasma rotation can be slowed down or be accelerated with the rotating RMP. When the rotating RMP frequency is higher than mode frequency, the plasma rotation can be accelerated to the rotating RMP frequency. The plasma confinement is improved with high frequency rotating RMP. The plasma rotation is decelerated to the rotating RMP frequency when the rotating RMP frequency is lower than the mode frequency. The plasma confinement also degrades with low frequency rotating RMP.
STARFIRE: a commercial tokamak reactor
Energy Technology Data Exchange (ETDEWEB)
The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor.
Tokamak Plasmas: Mirnov coil data analysis for tokamak ADITYA
Indian Academy of Sciences (India)
The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...
Tokamak Plasmas: Internal magnetic field measurement in tokamak ...
Indian Academy of Sciences (India)
The theory of the measurement and a detailed design of the Zeeman polarimeter constructed to measure the poloidal field profile in the ADITYA tokamak are presented. The Fabry-Perot which we have employed in our design, with photodiode arrays followed by lock-in detection of the polarization signal, allows the ...
MDSplus integration at TCABR tokamak: Current status
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Sá, W.P. de, E-mail: firstname.lastname@example.com; Ronchi, G., E-mail: email@example.com
Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.
Microtearing mode (MTM) turbulence in JIPPT-IIU tokamak plasmas
Hamada, Y.; Watari, T.; Nishizawa, A.; Yamagishi, O.; Narihara, K.; Ida, K.; Kawasumi, Y.; Ido, T.; Kojima, M.; Toi, K.; the JIPPT-IIU Group
Magnetic, density and potential fluctuations up to 500 kHz at several spatial points have been observed in the core region of JIPPT-IIU tokamak plasmas using a heavy ion beam probe. The frequency spectra of the density and magnetic oscillations are found to be similar, whereas there are large differences in the phase, coherence and frequency dependences deduced from signals at adjacent sample volumes. These differences allow us to ascribe the detected magnetic fluctuations to the microtearing mode (MTM) by simple dispersion relations of the MTM in collisionless and intermediate regimes. The frequency-integrated level of magnetic fluctuations around 150 kHz (100-200 kHz) is \\tilde{{B}}r /Bt ≈ 1× 10-4 , a level high enough for the ergodization of the magnetic surface and enhanced electron heat loss as derived by Rechester and Rosenbluth (1978 Phys. Rev. Lett. 40 38). This level is consistent with the measurements performed using cross-polarization scattering of microwaves in the Tore Supra tokamak. Our results are the first direct experimental verification of the MTM in the core region of tokamak plasmas, which has been recently observed in gyrokinetic simulations using a very fine mesh in tokamak and ST plasmas.
High-pressure, flux-conserving tokamak equilibria
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Dory, R.A.; Peng, Y.K.M.
Magnetohydrodynamic (MHD) tokamak equilibria are found with values of ..beta.. up to 20 percent and prescribed MHD safety factor values (e.g., q(axis) = 1 and q(edge) = 4.8) for tokamaks with aspect ratio A = 4 and D-shaped cross section. If such equilibria could be attained experimentally, they would be very attractive for decreasing the projected costs of tokamak power reactors substantially. In the flux-conserving tokamak (FCT) model, where rapid heating is applied to an already relatively hot plasma, these high ..beta.. equilibria are achievable. We study the quasi-static evolution of FCT equilibria as ..beta.. increases. An operating window is found in the pressure profile width w/sub p/: for high ..beta.. the values of w/sub p/ must lie between 0.40 and 0.55 of the plasma minor width. Within this window, plasma current and poloidal ..beta.. increase monotonically with ..beta... For fixed plasma boundary, significant poloidal surface currents are induced, but these can be eliminated by small increases in the plasma minor radius, the pressure profile width, and the vacuum toroidal field.
Gyrokinetic continuum simulations of turbulence in the Texas Helimak
Bernard, T. N.; Shi, E. L.; Hammett, G. W.; Hakim, A.; Taylor, E. I.
We have used the Gkeyll code to perform 3x-2v full-f gyrokinetic continuum simulations of electrostatic plasma turbulence in the Texas Helimak. The Helimak is an open field-line experiment with magnetic curvature and shear. It is useful for validating numerical codes due to its extensive diagnostics and simple, helical geometry, which is similar to the scrape-off layer region of tokamaks. Interchange and drift-wave modes are the main turbulence mechanisms in the device, and potential biasing is applied to study the effect of velocity shear on turbulence reduction. With Gkeyll, we varied field-line pitch angle and simulated biased and unbiased cases to study different turbulent regimes and turbulence reduction. These are the first kinetic simulations of the Helimak and resulting plasma profiles agree fairly well with experimental data. This research demonstrates Gkeyll's progress towards 5D simulations of the SOL region of fusion devices. Supported by the U.S. DOE SCGSR program under contract DE-SC0014664, the Max-Planck/Princeton Center for Plasma Physics, the SciDAC Center for the Study of Plasma Microturbulence, and DOE contract DE-AC02-09CH11466.
This is a draft Landsat Data Mosaic, which contains remote sensing information for Montgomery County, Texas Harris County, Texas Fort Bend County, Texas Brazoria County, Texas Galveston County, and Texas Imagery dates on the following dates: October 6, 1999 and September 29, 200...
Sliding Mode Control of a Tokamak Transformer
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Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.
A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.
Application of avalanche photodiode for soft X-ray pulse-height analyses in the Ht-7 tokamak
Shi Yue Jiang; Hu Li Qun; Sun Yan Jun; LiuSheng; Ling Bil
An avalanche photodiode (APD) has been used as soft X-ray energy pulse-height analysis system for the measurement of the electron temperature on the HT-7 tokamak. The experimental results obtained with the APD with its inferior energy resolution show a little difference compared to the conventional high energy-resolution Si (Li) detector. Both numerical analysis and experimental results prove that the APD is good enough for application of the electron temperature measurement in tokamaks.
Characterization of the Novillo Tokamak in main discharge regime; Caracterizacion del Tokamak Novillo en regimen de descarga principal
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Lopez C, R.; Melendez L, L.; Chavez A, E.; Colunga S, S.; Valencia A, R.; Gaytan G, E
The analytical procedure to carry out the establishment of the discharge in a Tokamak including: a) Ionization, b) Diffusion losses, recombination, union, drift speed, spurious fields, and c) Electric field is presented. In an experimental way a procedure settles down by means of which it is characterized the plasma, specially a new characteristic discharge parameter is settled down and it is the plasma current by the duration of the (I{sub p}t) discharge. (Author)
Transport of dust particles in tokamak devices
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Pigarov, A.Yu. [University of California at San Diego, La Jolla, CA (United States)]. E-mail: email@example.com; Smirnov, R.D. [University of California at San Diego, La Jolla, CA (United States); Krasheninnikov, S.I. [University of California at San Diego, La Jolla, CA (United States); Rognlien, T.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Rosenberg, M. [University of California at San Diego, La Jolla, CA (United States); Soboleva, T.K. [UNAM, Mexico, DistritoFederal (Mexico)
Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.
Plasma position control in the STOR-M tokamak: A fuzzy logic approach
Morelli, Jordan Edwin
Adequate control of the position of the plasma column within the STOR-M tokamak is a chief requirement in order for experimental quality discharges to be obtained. Optimal control over tokamak discharge parameters, including the plasma position, is very difficult to achieve. This is due in large part to the difficulty in modelling the tokamak discharge parameters, as they are highly nonlinear and time varying in nature. The difficulty of modelling the tokamak discharge parameters suggests that a control system, such as a fuzzy logic based controller, which does not require a system model may be well suited to the control of fusion plasma. In order to improve the quality of control over the plasma position within the STOR-M tokamak, the existing analog PID controller was modified. These modifications facilitate the application of a digital controller by a personal computer via the Advantech PCL-711B data acquisition card. The performance of the modified plasma position controller and an Arbitrary Signal Generator developed by the author was evaluated. This modified plasma position controller was applied successfully to the STOR-M tokamak during both normal mode and A.C. mode operation. In both cases, the modified controller provided adequate control over the position of the plasma column within the discharge chamber. Furthermore, the modified controller was more convenient to optimize than the original, existing analog PID controller. By taking advantage of the modifications that were made to the plasma position controller, a fuzzy logic controller was developed by the author. The fuzzy logic based plasma position controller was also successfully applied to the STOR-M tokamak during both normal mode and A.C. operation. The fuzzy controller was demonstrated to reliably provide a higher degree of control over the position of the plasma column within the STOR-M tokamak than the modified PID controller.
Features of self-organized plasma physics in tokamaks
Razumova, K. A.
The history of investigations the role of self-organization processes in tokamak plasma confinement is presented. It was experimentally shown that the normalized pressure profile is the same for different tokamaks. Instead of the conventional Fick equation, where the thermal flux is proportional to a pressure gradient, processes in the plasma are well described by the Dyabilanin's energy balance equation, in which the heat flux is proportional to the difference of normalized gradients for self-consistent and real pressure profiles. The transport coefficient depends on the values of heat flux, which compensates distortion of the pressure profile with external impacts. Radiative cooling of the plasma edge decreases the heat flux and improves the confinement.
Stability and heating of a poloidal divertor tokamak
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Biddle, A. P.; Dexter, R. N.; Holly, D. T.; Lipschultz, B.; Osborne, T. H.; Prager, S. C.; Shepard, D.A., Sprott, J.C.; Witherspoon, F. D.
Five experimental studies - two stability and three heating investigations - have been carried out on Tokapole II, a Tokamak with a four node poloidal divertor. First, discharges have been attained with safety factor q as low as 0.6 over most of the column without degradation of confinement, and correlation of helical instability onset with current profile shape is being studied. Second, the axisymmetric instability has been investigated in detail for various noncircular cross-sectional shapes, and results have been compared with a numerical stability code adapted to the Tokapole machine. Third, application of high power fast wave ion cyclotron resonance heating doubles the ion temperature and permits observation of heating as a function of harmonic number and spatial location of the resonance. Fourth, low power shear Alfven wave propagation is underway to test the applicability of this heating method to tokamaks. Fifth, preionization by electron cyclotron heating has been employed to reduce the startup loop voltage by approx. 60%.
Three-dimensional analysis of tokamaks and stellarators.
Garabedian, Paul R
The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project.
Three-dimensional analysis of tokamaks and stellarators
Garabedian, Paul R.
The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807
Three-dimensional equilibria in axially symmetric tokamaks
Garabedian, Paul R.
The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of α particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158
Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bremond, S.
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.
Statistical analysis of first period of operation of FTU Tokamak; Analisi statistica del primo periodo di operazioni del Tokamak FTU
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Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia
On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted.
Diagnostics for the Rijnhuizen Tokamak Project
Donne, A. J. H.
The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel
The disruptive instability in Tokamak plasmas
Salzedas, F.J.B.
Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative
Electron density and temperature determination in a Tokamak plasma using light scattering; Determinacion de la densidad y temperatura electronicas en un Tokamak mediante difusion luminosa
Energy Technology Data Exchange (ETDEWEB)
Perez-Navarro Gomerz, A.; Zurro Hernandez, B.
A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs.
Natural current profiles in a tokamak
Energy Technology Data Exchange (ETDEWEB)
Taylor, J.B.
In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described.
Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks
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Valencia A, R
As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission
Liquid nitrogen cooling considerations of the Compact Ignition Tokamak
Energy Technology Data Exchange (ETDEWEB)
Dabiri, A.E.
A simple model was developed to estimate the cooldown time between pulses of toroidal field (TF) coils of the Compact Ignition Tokamak (CIT) using liquid nitrogen. Good agreement was obtained between the analysis results and those measured in the early fusion experimental devices. A cooldown time of about 1 h would reduce the TF coil temperature to about 80 K. An R and D experimental program is required to determine the actual cooldown time between pulses, an issue in the conceptual design of the CIT.
Improved confinement induced by tangential injection of compact torus into the Saskatchewan Torus-Modified (STOR-M) tokamak
Xiao, C.; Hirose, A.; Sen, S.
Compact torus injection into the Saskatchewan Torus-Modified [Phys. Fluids B 4, 3277 (1992)] tokamak discharges has triggered improved confinement characterized by an increase in the electron density by more than twofold, 30% reduction in the Hα radiation level, significant suppression of floating potential fluctuations and m=2 Mirnov oscillations. In this paper, we present detailed experimental setup and results, as well as an extended theory explaining the mechanism for triggering improved confinement in a tokamak by compact torus injection.
Plasma recombination and molecular effects in tokamak divertors and divertor simulators
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Krasheninnikov, S.I.; Pigarov, A.Y.; Knoll, D.A.; LaBombard, B.; Lipschultz, B.; Sigmar, D.J.; Soboleva, T.K.; Terry, J.L.; Wising, F. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)]|[Department of Physics, The College of William and Mary, Williamsburg, Virginia 23187 (United States)]|[Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)]|[Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Mexico D.F. (Mexico)]|[Institute for Electromagnetic Field Theory, Chalmers University of Technology, S-41296 Gothenburg (Sweden)
Analysis of the experimental data from tokamaks and linear divertor simulators leads to the conclusion that plasma recombination is a crucial element of plasma detachment. Different mechanisms of plasma recombination relevant to the experimental conditions of the tokamak scrape-off layer (SOL) and divertor simulators are considered. The physics of Molecular Activated Recombination (MAR) involving vibrationally excited molecular hydrogen are discussed. Although conventional Electron{endash}Ion Recombination (EIR) alone can strongly alter the plasma parameters, MAR impact can be substantial for both tokamak SOL plasma and divertor simulators. Investigation of the effects of EIR on the plasma flow in divertor simulators shows that due to the balances of (a) energy transport and electron cooling, and (b) the plasma flow and recombination, that EIR extinguishes the simulator plasma at an electron temperature about 0.15 eV. {copyright} {ital 1997 American Institute of Physics.}
Energy Technology Data Exchange (ETDEWEB)
OAK B202 TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE. Several tokamak experiments have reported the development of a central region with vanishing currents (the current hole). Straightforward application of results from the work of Greene, Johnson and Weimer [Phys. Fluids, 3, 67 (1971)] on tokamak equilibrium to these plasmas leads to apparent singularities in several physical quantities including the Shafranov shift and casts doubts on the existence of this type of equilibria. In this paper, the above quoted equilibrium theory is re-examined and extended to include equilibria with a current hole. It is shown that singularities can be circumvented and that equilibria with a central current hole do satisfy the magnetohydrodynamic equilibrium condition with regular behavior for all the physical quantities and do not lead to infinitely large Shafranov shifts. Isolated equilibria with negative current in the central region could exist. But equilibria with negative currents in general do not have neighboring equilibria and thus cannot have experimental realization, i.e. no negative currents can be driven in the central region.
Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry
Energy Technology Data Exchange (ETDEWEB)
Long-Poe Ku and Allen H. Boozer
If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.
Dust-Particle Transport in Tokamak Edge Plasmas
Energy Technology Data Exchange (ETDEWEB)
Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D
Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.
Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows
Cohen, Bruce; Umansky, Maxim
Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.
Optimization of magnetic field system for glass spherical tokamak GLAST-III
Ahmad, Zahoor; Ahmad, S.; Naveed, M. A.; Deeba, F.; Aqib Javeed, M.; Batool, S.; Hussain, S.; Vorobyov, G. M.
GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA-1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils.
Cluster storage for COMPASS tokamak
Czech Academy of Sciences Publication Activity Database
Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír
Roč. 87, č. 12 (2012), s. 2238-2241 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006
(Injection of compact toroids for tokamak fueling and current drive)
Energy Technology Data Exchange (ETDEWEB)
Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.
The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.
Horton, Anita
Dallas artist Jon Flaming's deep love of Texas is evident in his paintings and sculpture. Although he has created one sculptural Texas theater, his work primarily showcases old Texas barbershops, vacant homes, and gas stations. In this article, the author describes how her students, inspired by Flaming's works, created three-dimensional historical…
Lower hybrid current drive in tokamak plasmas
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Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)
Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak
Energy Technology Data Exchange (ETDEWEB)
Fraboulet, D.
Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.
MHD stability limits in the TCV Tokamak
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Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)
Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The
Tangential and Vertical Compact Torus Injection Experiments on the STOR-M Tokamak
Xiao, Chijin; D, Liu; S, Livingstone; A, K. Singh; E, Zhang; A, Hirose
This paper describes the setup and results of compact torus (CT) injection experiments on the STOR-M tokamak. Tangential CT injection into STOR-M induced H-mode-like phenomena including doubling the electron density, reduction in the Hα radiation level, suppression of the floating potential fluctuations, suppression of the m = 2 Mirnov oscillations, and increase in the global energy confinement time. Experimental setup, bench-test results, and some preliminary injection data for vertical CT injection experiments on STOR-M will be shown. In addition, numerical simulations of the CT trajectories in tokamak discharges for both tangential and vertical injection geometries will be discussed.
Identification of the ubiquitous Coriolis momentum pinch in JET tokamak plasmas
Weisen, H.; Camenen, Y.; Salmi, A.; Versloot, T. W.; de Vries, P. C.; Maslov, M.; Tala, T.; Beurskens, M.; Giroud, C.
A broad survey of the experimental database of neutral beam heated plasmas in the JET tokamak has established the theoretically expected ubiquity, in rotating plasmas, of a convective transport mechanism which has its origin in the vertical particle drift resulting from the Coriolis force. This
Robust Sliding Mode Control for Tokamaks
Directory of Open Access Journals (Sweden)
I. Garrido
Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.
Macroscopic erosion of divertor and first wall armour in future tokamaks
Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.
Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.
The ETE spherical Tokamak project. IAEA report
Energy Technology Data Exchange (ETDEWEB)
Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: email@example.com
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Economic considerations of commercial tokamak options
Energy Technology Data Exchange (ETDEWEB)
Dabiri, A.E.
Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m/sup 2/, which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m/sup 2/, will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e).
Tokamak active laser pyrometry for tungsten deposited layer characterisation
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Semerok, A., E-mail: firstname.lastname@example.com [CEA Saclay, DEN/DENS/DPC/SCP/LILM, P.C. 56, 91191 Gif-sur-Yvette, Cedex (France); Jaubert, F.; Fomichev, S.V.; Thro, P.-Y. [CEA Saclay, DEN/DENS/DPC/SCP/LILM, P.C. 56, 91191 Gif-sur-Yvette, Cedex (France); Grisolia, C. [CEA Cadarache, IRFM, 13108, Saint Paul-lez-Durance, Cedex (France)
In modern fusion reactors, the erosion of plasma facing surface results in layers deposition on tokamak 'cold' surfaces. To provide efficient operation of tokamaks, it is essential to characterise the deposited layer with high tritium content. In situ rapid surface characterisation without reactor components disassembly is required. Active laser pyrometry together with a repetition rate Nd-YAG laser (1 Hz-1 kHz repetition rate frequency) applied for surface heating can be used to characterise some thermo-physical properties (thermal capacity, thermal contact, and conductivity) of a micrometric layer. The pyrometer system was developed and applied to characterise some properties of a W-layer (140 {mu}m) on a CFC-substrate. The numerical code developed for 3-D simulation of LH of a surface with the deposited layer was applied to simulate the experimental heating temperatures. The experimental and simulation results were compared. W-layer characterisation was performed by fitting the experimental and theoretical heating temperatures.
Impurity seeding in ASDEX upgrade tokamak modeled by COREDIV code
Energy Technology Data Exchange (ETDEWEB)
Galazka, K.; Ivanova-Stanik, I.; Czarnecka, A.; Zagoerski, R. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Bernert, M.; Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: ASDEX Upgrade Team
The self-consistent COREDIV code is used to simulate discharges in a tokamak plasma, especially the influence of impurities during nitrogen and argon seeding on the key plasma parameters. The calculations are performed with and without taking into account the W prompt redeposition in the divertor area and are compared to the experimental results acquired on ASDEX Upgrade tokamak (shots 29254 and 29257). For both impurities the modeling shows a better agreement with the experiment in the case without prompt redeposition. It is attributed to higher average tungsten concentration, which on the other hand seriously exceeds the experimental value. By turning the prompt redeposition process on, the W concentration is lowered, what, in turn, results in underestimation of the radiative power losses. By analyzing the influence of the transport coefficients on the radiative power loss and average W concentration it is concluded that the way to compromise the opposing tendencies is to include the edge-localized mode flushing mechanism into the code, which dominates the experimental particle and energy balance. Also performing the calculations with both anomalous and neoclassical diffusion transport mechanisms included is suggested. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)
Study of the electron heat transport in Tore-Supra tokamak; Etude du transport de la chaleur electronique dans le Tokamak Tore Supra
Energy Technology Data Exchange (ETDEWEB)
Harauchamps, E
This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)
Integrated tokamak modeling: when physics informs engineering and research planning
Poli, Francesca
Simulations that integrate virtually all the relevant engineering and physics aspects of a real tokamak experiment are a power tool for experimental interpretation, model validation and planning for both present and future devices. This tutorial will guide through the building blocks of an ``integrated'' tokamak simulation, such as magnetic flux diffusion, thermal, momentum and particle transport, external heating and current drive sources, wall particle sources and sinks. Emphasis is given to the connection and interplay between external actuators and plasma response, between the slow time scales of the current diffusion and the fast time scales of transport, and how reduced and high-fidelity models can contribute to simulate a whole device. To illustrate the potential and limitations of integrated tokamak modeling for discharge prediction, a helium plasma scenario for the ITER pre-nuclear phase is taken as an example. This scenario presents challenges because it requires core-edge integration and advanced models for interaction between waves and fast-ions, which are subject to a limited experimental database for validation and guidance. Starting from a scenario obtained by re-scaling parameters from the demonstration inductive ``ITER baseline'', it is shown how self-consistent simulations that encompass both core and edge plasma regions, as well as high-fidelity heating and current drive source models are needed to set constraints on the density, magnetic field and heating scheme. This tutorial aims at demonstrating how integrated modeling, when used with adequate level of criticism, can not only support design of operational scenarios, but also help to asses the limitations and gaps in the available models, thus indicating where improved modeling tools are required and how present experiments can help their validation and inform research planning. Work supported by DOE under DE-AC02-09CH1146.
Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code
Directory of Open Access Journals (Sweden)
Yutthapong Pinanroj
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.
Observation of MHD phenomenon for SST-1 superconducting tokamak
Bhandarkar, Manisha; Dhongde, Jasraj; Pradhan, Subrata
Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (major radius = 1.1 m, minor radius = 0.2 m) and is operational at the Institute for Plasma Research (IPR), India. In the last few experimental campaigns SST-1 has successfully achieved plasma current in order of 60-70kA and plasma duration in excess of ∼ 500 ms at a central magnetic field of 1.5T. An attempt has made to study the behavior of the magneto-hydrodynamic (MHD) activity during different phases of plasma pulse which leads to major/minor disruptions, its present modes (poloidal/toroidal mode number i.e. m = 2, n = 1) impact on plasma confinement and signature of lock mode and its frequency in the SST-1 plasma using experimental data from Mirnov signals. Observed MHD phenomenon has also been correlated with other diagnostics (i.e. ECE, Density, Soft X-Ray etc.) and heating system (ECRH) for the recent campaigns of SST-1.
Finite pressure effects on the tokamak sawtooth crash
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Nishimura, Yasutaro
The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed.
Electron thermal transport in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Konings, J.A.
The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).
Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows
Cohen, Bruce; Umansky, Maxim
Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.
Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks
Directory of Open Access Journals (Sweden)
Peysson Y.
Full Text Available The Lower Hybrid (LH wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum. Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model, for robust interpretative and predictive simulations.
Integrated modeling of temperature profiles in L-mode tokamak discharges
Energy Technology Data Exchange (ETDEWEB)
Rafiq, T.; Kritz, A. H.; Tangri, V. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)
Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.
Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN; Un sistema espectroscopico para medidas de impurezas en el Tokamak TJ-1 de la JEN
Energy Technology Data Exchange (ETDEWEB)
Navas, G.; Zurro, B.
we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs.
A charged fusion product diagnostic for a spherical tokamak
Perez, Ramona Leticia Valenzuela
Designs for future nuclear fusion power reactors rely on the ability to create a stable plasma (hot ionized gas of hydrogen isotopes) as a medium with which to sustain nuclear fusion reactions. My dissertation work involves designing, constructing, testing, installing, operating, and validating a new diagnostic for spherical tokamaks, a type of reactor test facility. Through detecting charged particles emitted from the plasma, this instrument can be used to study fusion reaction rates within the plasma and how they are affected by plasma perturbations. Quantitatively assessing nuclear fusion reaction rates at specific locations inside the plasma and as a function of time can provide valuable data that can be used to evaluate theory-based simulations related to energy transport and plasma stability. The Proton Detector (PD), installed in the Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy (CCFE) in Abingdon, England, was the first instrument to experimentally detect 3 MeV Protons and 1 MeV Tritons created from deuterium- deuterium (hydrogen isotopes) nuclear fusion reactions inside a spherical tokamak's plasma. The PD consists of an array of particle detectors with a protective housing and the necessary signal conditioning electronics and readout. After several years of designing (which included simulations for detector orientations), fabricating, and testing the PD, it was installed in MAST and data were collected over a period of two months in the summer of 2013. Proton and triton rates as high as 200 kHz were measured and an initial radial profile of these fusion reaction rates inside the plasma was extracted. These results will be compared to a complementary instrument at MAST as well as theory-based simulations and form the knowledge basis for developing a larger future instrument. The design and performance of all instrument components (electrical, computational, mechanical), and subsequent data analysis methods and results are
MHD Effects of a Ferritic Wall on Tokamak Plasmas
Hughes, Paul E.
It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency
Study of energy transport in Tore Supra Tokamak; Etude du transport de l`energie sur le Tokamak Tore Supra
Energy Technology Data Exchange (ETDEWEB)
Guiziou, L.
The goal of this thesis is to characterize the energy confinement and the heat transport in Tore Supra tokamak. The first chapter is an introduction to the different plasma confinement regimes: ohmic, low confinement and improved confinement regimes. The second chapter is devoted to the presentation of the different theoretical and empirical approaches about energy confinement and heat transport. In the third chapter an attempt of explanations for non-local transport phenomenons is given. A turbulence correlation length greater than the ionic Larmor radius seams to be a reasonable explanation. This theoretical study focusses on the possibility for modes coupling in a tokamak. This study tries to determine a radial correlation length considering the two principal coupling modes: toroidal and non-linear. Different transport regimes are discussed using an analytical model and considering the influence of one coupling with respect to the other. In chapter four, the measurements of current profiles and transport coefficients are presented. The codes used for the reconstruction of equilibrium and for the experimental determination of the diffusivity are briefly presented. In chapter five, experimental results of energy transport studies for Tore Supra plasmas are presented. The different modes are analysed in detail and the study focusses on the influence of magnetic shear in the improved confinement regime. Finally, the different parametric dependences of the electronic thermal diffusivity are compared to local transport models. 165 refs., 57 figs., 2 tabs., 2 appendix.
Ball, Justin
Bulk toroidal rotation has proven capable of stabilising both dangerous MHD modes and turbulence. In this thesis, we explore a method to drive rotation in large tokamaks: up-down asymmetry in the magnetic equilibrium. We seek to maximise this rotation by finding optimal up-down asymmetric flux surface shapes. First, we use the ideal MHD model to show that low order external shaping (e.g. elongation) is best for creating up-down asymmetric flux surfaces throughout the device. Then, we calculate realistic up-down asymmetric equilibria for input into nonlinear gyrokinetic turbulence analysis. Analytic gyrokinetics shows that, in the limit of fast shaping effects, a poloidal tilt of the flux surface shaping has little effect on turbulent transport. Since up-down symmetric surfaces do not transport momentum, this invariance to tilt implies that devices with mirror symmetry about any line in the poloidal plane will drive minimal rotation. Accordingly, further analytic investigation suggests that non-mirror symmetri...
Parasitic momentum flux in the tokamak core
Stoltzfus-Dueck, T.
Tokamak plasmas rotate spontaneously without applied torque. This intrinsic rotation is important for future low-torque devices such as ITER, since rotation stabilizes certain instabilities. In the mid-radius `gradient region,' which reaches from the sawtooth inversion radius out to the pedestal top, intrinsic rotation profiles may be either flat or hollow, and can transition suddenly between these two states, an unexplained phenomenon referred to as rotation reversal. Theoretical efforts to explain the mid-radius rotation shear have largely focused on quasilinear models, in which the phase relationships of some selected instability result in a nondiffusive momentum flux (``residual stress''). In contrast, the present work demonstrates the existence of a robust, fully nonlinear symmetry-breaking momentum flux that follows from the free-energy flow in phase space and does not depend on any assumed linear eigenmode structure. The physical origin is an often-neglected portion of the radial ExB drift, which is shown to drive a symmetry-breaking outward flux of co-current momentum whenever free energy is transferred from the electrostatic potential to ion parallel flows. The fully nonlinear derivation relies only on conservation properties and symmetry, thus retaining the important contribution of damped modes. The resulting rotation peaking is counter-current and scales as temperature over plasma current. As first demonstrated by Landau, this free-energy transfer (thus also the corresponding residual stress) becomes inactive when frequencies are much higher than the ion transit frequency, which allows sudden transitions between hollow and flat profiles. Simple estimates suggest that this mechanism may be consistent with experimental observations. This work was funded in part by the Max-Planck/Princeton Center for Plasma Physics and in part by the U.S. Dept. of Energy, Office of Science, Contract No. DE-AC02-09CH11466.
Magnetic flux reconstruction methods for shaped tokamaks
Energy Technology Data Exchange (ETDEWEB)
Tsui, Chi-Wa [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
The use of a variational method permits the Grad-Shafranov (GS) equation to be solved by reducing the problem of solving the 2D non-linear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a parameterization of the plasma boundary and the current profile (p` and FF` functions). The author treats the current profile parameters as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. Matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green`s function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing Principle provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. The performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package has been implemented which integrates a vacuum field solver using a filament model for the plasma, a multi-layer perception neural network as an interface, and the volume integration of plasma current density using Green`s functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data. The results are promising.
Texas motorcycle crash countermeasure workshop.
The Texas Department of Transportation (TxDOT) contracted with the Texas A&M : Transportation Institute (TTI) to develop a 5-year strategic plan for improving motorcycle safety : in the State of Texas. The Texas Strategic Action Plan for Motorcycl...
Tokamak power systems studies, FY 1985
Energy Technology Data Exchange (ETDEWEB)
Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.
The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.
Overview of spherical tokamak research in Japan
Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.
Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.
Electron cyclotron emission diagnostics on KSTAR tokamak.
Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M
A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.
A Fast Shutdown Technique for Large Tokamaks
Energy Technology Data Exchange (ETDEWEB)
E. Fredrickson; G.L. Schmidt; K. Hill; S.C. Jardin; et al
A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR.
Study of heat flux deposition in the Tore Supra Tokamak; Etude des depots de chaleur dans le tokamak Tore Supra
Energy Technology Data Exchange (ETDEWEB)
Carpentier, S.
Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length lambda{sub q} (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length lambda{sub q} in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)
Plasma recombination in runaway discharges in tokamak TCABR
Energy Technology Data Exchange (ETDEWEB)
Soboleva, T.K. [Universidad Nacional Autonoma de Mexico, Mexico City (Mexico). Inst. de Ciencias Nucleares; Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Krasheninnikov, S.I. [University of California, San Diego, CA (United States)
A new regime of runaway discharges has been observed in the TCABR tokamak. One of the most distinctive features of this regime is the effect of plasma detachment from the limiter. This experimental fact can only be explained by the volume recombination, which requires a low-temperature plasma. The analysis of the energy and particle balance in the system plasma-relativistic runaway beam in TCABR, which takes into account only the collisional mechanism of the heat transfer from runaways to thermal electrons, predicts electron temperatures T{sub e} = 0.1 - 2 eV; the temperature decreases with the neutral density increase. The recombination process with the rate constant around 10{sup -16} m3/s is required for the explanation of plasma density behavior in the experiment. At present, it is difficult to conclude about the mechanism of recombination. More reliable and detailed experimental data, mainly about the plasma temperature, are necessary. (author)
Magnetic field threshold for runaway generation in tokamak disruptions
F"Ul"Op, T.; Pokol, G.; Smith, H. M.; Helander, P.
Due to a sudden cooling of the plasma in tokamak disruptions a beam of relativistic runaway electrons is sometimes generated, which may cause damage on plasma facing components. Experimental observations on large tokamaks show that the number of runaway electrons produced in disruptions depends on the magnetic field strength. In this work, two possible reasons for this threshold are studied. The first possible explanation for these observations is that the runaway beam excites whistler waves that scatter the electrons in velocity space and prevents the beam from growing. The growth rates of the most unstable whistler waves are inversely proportional to the magnetic field strength and it is possible to derive a magnetic field threshold below which no runaways are expected. The second possible explanation is the magnetic field dependence of the criterion for substantial runaway production determined by the induced electric field available and by the efficiency of the generation mechanisms. It is shown, that even in rapidly cooling plasmas, where hot-tail generation is expected to give rise to substantial runaway population, the whistler waves can stop the runaway formation below a certain magnetic field unless the post-disruption temperature is very low.
Active cooling system for Tokamak in-vessel operation manipulator
Energy Technology Data Exchange (ETDEWEB)
Yuan, Jianjun, E-mail: email@example.com; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang
Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.
Overview of the EUROfusion Medium Size Tokamak program
Martin, Piero; Beurskens, Marc; Coda, Stefano; Eich, Thomas; Meyer, Hendrik; the EUROfusion MST1 Team
As a result of the new organization of the European fusion programme, now under the umbrella of the EUROfusion Consortium, the MST (Medium Size Tokamaks) task force is in charge of executing the European science programme in the ASDEX Upgrade, TCV and MAST-U tokamaks. This paper will present an overview of the main results obtained in the 2014 campaign-where only ASDEX upgrade was operating-and the preliminary achievements of the recently started 2015/16 campaign, where also TCV will contribute. The main subjects of the experimental campaigns are (i) the development of scenarios relevant for the ITER Q=10 goal, in an all metal wall device (ii) the understanding of ELM mitigation/suppression with pellets and resonant magnetic perturbations, and in particular the effect of density versus collisionality, (iii) the understanding and optimization of methods for disruption mitigation or avoidance and runaway electrons control and (iv) the exploration of ITER and DEMO relevant scenarios with high normalized separatrix power flux, Psep / R , (Psep is the power through the separatrix, R the major radius) and tolerable target heat loads. The overview of the future programs in MST will be given. http://www.euro-fusionscipub.org/mst1
Physics of GAM-initiated L-H transition in a tokamak
Askinazi, L. G.; Belokurov, A. A.; Bulanin, V. V.; Gurchenko, A. D.; Gusakov, E. Z.; Kiviniemi, T. P.; Lebedev, S. V.; Kornev, V. A.; Korpilo, T.; Krikunov, S. V.; Leerink, S.; Machielsen, M.; Niskala, P.; Petrov, A. V.; Tukachinsky, A. S.; Yashin, A. Yu; Zhubr, N. A.
Based on experimental observations using the TUMAN-3M and FT-2 tokamaks, and the results of gyrokinetic modeling of the interplay between turbulence and the geodesic acoustic mode (GAM) in these installations, a simple model is proposed for the analysis of the conditions required for L-H transition triggering by a burst of radial electric field oscillations in a tokamak. In the framework of this model, one-dimensional density evolution is considered to be governed by an anomalous diffusion coefficient dependent on radial electric field shear. The radial electric field is taken as the sum of the oscillating term and the quasi-stationary one determined by density and ion temperature gradients through a neoclassical formula. If the oscillating field parameters (amplitude, frequency, etc) are properly adjusted, a transport barrier forms at the plasma periphery and sustains after the oscillations are switched off, manifesting a transition into the high confinement mode with a strong inhomogeneous radial electric field and suppressed transport at the plasma edge. The electric field oscillation parameters required for L-H transition triggering are compared with the GAM parameters observed at the TUMAN-3M (in the discharges with ohmic L-H transition) and FT-2 tokamaks (where no clear L-H transition was observed). It is concluded based on this comparison that the GAM may act as a trigger for the L-H transition, provided that certain conditions for GAM oscillation and tokamak discharge are met.
Sylvatic trichinellosis in Texas
Directory of Open Access Journals (Sweden)
Pence D.B.
Full Text Available There are no published reports of domestic or sylvatic trichinellosis in Texas. The aim of the present survey was to determine the presence of Trichinella species in selected representative species of potential wildlife reservoirs in southern Texas. In 1998-99, tongues of 211 wild mammals were collected in southern Texas: 154 coyotes (Canis latrans, three bobcats (Lynx rufus, 32 racoons (Procyon lotor, 1 3 opossum (Didelphis marsupialis, four ocelots (Leopardus pardalis and five wild boars (Sus scrofa. Presence of Trichinella sp. larvae was investigated by artificial digestion and larvae of positive samples were identified at the species level by a multiple-polymerase chain reaction analysis. Nine (5.8 % coyotes had trichinellosis ; in the muscles of seven of these coyotes, the larvae were identified as Trichinella murrelli. This is the first report of sylvatic trichinellosis in Texas.
National Aeronautics and Space Administration — The Terra eXperiment 2001 was conducted from Kelly AFB San Antonio, Texas from March 14 to April 4 to improve calibration of the MODerate resolution Imaging...
Detachment evolution on the TCV tokamak
Harrison, J. R.; Vijvers, W. A. J.; Theiler, C.; Duval, B. P.; Elmore, S.; Labit, B.; Lipschultz, B.; van Limpt, S. H. M.; Lisgo, S. W.; Tsui, C. K.; Reimerdes, H.; Sheikh, U.; Verhaegh, K. H. A.; Wischmeier, M.
Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1]. Before the roll-over in the ion
Tokamak Plasmas: Electron temperature $(T_ {e}) $ measurements ...
Indian Academy of Sciences (India)
Thomson scattering technique based on high power laser has already proved its superoirity in measuring the electron temperature (e) and density (e) in fusion plasma devices like tokamaks. The method is a direct and unambiguous one, widely used for the localised and simultaneous measurements of the above ...
Tokamak Transport Studies Using Perturbation Analysis
Cardozo, N. J. L.; Dehaas, J. C. M.; Hogeweij, G. M. D.; Orourke, J.; Sips, A.C.C.; Tubbing, B. J. D.
Studies of the transport properties of tokamak plasmas using perturbation analysis are discussed. The focus is on experiments with not too large perturbations, such as sawtooth induced heat and density pulse propagation, power modulation and oscillatory gas-puff experiments. The approximations made
On dust in tokamak edge plasmas
Energy Technology Data Exchange (ETDEWEB)
Krasheninnikov, S.I. [Jacobs School of Engineering, Department of Mechanical and Aerospace Engineering, University of California at San Diego, Engineering Building II, room 474, 9500 Gilman Drive, La Jolla, CA 92093-0411 (United States)]. E-mail: email@example.com; Soboleva, T.K. [UNAM, Mexico, DF (Mexico); Kurchatov Institute, Moscow (Russian Federation); Tomita, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Smirnov, R.D. [Graduate University for Advanced Studies, Toki, Gifu 509-5292 (Japan); Janev, R.K. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)
We study the dust particle dynamics in tokamak edge plasmas, with special emphasis on dust particle transport in the sheath and plasma recycling regions. The characteristics of this transport have been examined for both smooth and corrugated wall surfaces. The implications of dust particle transport in the divertor region on the core plasma contamination with impurities have also been examined.
Tokamak startup with electron cyclotron heating
Energy Technology Data Exchange (ETDEWEB)
Holly, D J; Prager, S C; Shepard, D A; Sprott, J C
Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.
Centers for Disease Control (CDC) Podcasts
Dr. Kristy Murray, an associate professor in pediatrics and assistant dean of the National School of Tropical Medicine at Baylor College of Medicine and Texas Childrenâs Hospital, discusses increased cases of typhus in southern Texas. Created: 7/6/2017 by National Center for Emerging and Zoonotic Infectious Diseases (NCEZID). Date Released: 7/6/2017.
Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak; Determinacion de la densidad de un plasma por interferometria de microondas. El interferometro de 2 mm del Tokamak TJ-1
Energy Technology Data Exchange (ETDEWEB)
Martin, R.; Manero, F.
In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs.
Efficiency of the generation of impulsion by cyclotron waves currents of the electrons in an Axisymmetric Tokamak; Eficiencia de la generacion de corrientes de impulsion por ondas ciclotronicas de los electrones en un Tokamak axisimetrico
Energy Technology Data Exchange (ETDEWEB)
Gutierrez T, C.; Beltran P, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)
The neoclassical theory of transport is used to calculate the current efficiency of electronic cyclotron impulsion (ECCD) in an axisymmetric tokamak in the few collisions regime. The standard parameter of the tokamak is used to obtain a system of equations that describe the hydrodynamic of the plasma, where the ponderomotive force (PM) due to high power radio frequency waves is taken in account. The PM force is produced in the proximity of electron cyclotron resonance surface in a specific poloidal localization. The efficiency ECCD is analyzed in the cases of first and second harmonic (for different angles of injection of radio frequency waves) and it is validated using the experimental values of the TCV and T-10 tokamaks. The results are according to those obtained by means of the techniques of the Green functions. (Author)
Blob/hole formation and zonal-flow generation in the edge plasma of the JET tokamak
DEFF Research Database (Denmark)
Xu, G.S.; Naulin, Volker; Fundamenski, W.
The first experimental evidence showing the connection between blob/hole formation and zonal-flow generation was obtained in the edge plasma of the JET tokamak. Holes as well as blobs are observed to be born in the edge shear layer, where zonal-flows shear off meso-scale coherent structures, lead...
Effect of electron-to-ion mass ratio on radial electric field generation in tokamak
Li, Zhenqian; Dong, Jiaqi; Sheng, Zhengmao; Yu, M. Y.; Wang, Weixing
Generation of coherent radial electric fields in plasma by drift-wave turbulence driven by plasma inhomogeneities is ab initio studied using gyro-kinetic particle simulation for conditions of operational tokamaks. In particular, the effect of the electron-to-ion mass ratio ɛ on the entire evolution of the plasma is considered. It is found that the electric field can be increased, and the turbulence-induced particle transport reduced, by making ɛ smaller, in agreement with many existing experimental observations.
Signature of Turbulent Zonal Flows Observed in the DIII-D Tokamak
Energy Technology Data Exchange (ETDEWEB)
Coda, S.; Porkolab, M.; Burrell, K. H.
The spectrum of turbulent density fluctuations at long poloidal wavelengths in the edge plasma of the DIII-D tokamak peaks at nonzero radial wave number. The associated electric-potential fluctuations cause sheared {tilde E}xB flows primarily in the poloidal direction. These zonal flows have been predicted by theory and are believed to regulate the overall level of turbulence and anomalous transport. This study provides the first indirect experimental identification of zonal flows.
Signature of turbulent zonal flows observed in the DIII-D tokamak.
Coda, S; Porkolab, M; Burrell, K H
The spectrum of turbulent density fluctuations at long poloidal wavelengths in the edge plasma of the DIII-D tokamak peaks at nonzero radial wave number. The associated electric-potential fluctuations cause sheared E x B flows primarily in the poloidal direction. These zonal flows have been predicted by theory and are believed to regulate the overall level of turbulence and anomalous transport. This study provides the first indirect experimental identification of zonal flows.
Effect of the Shafranov shift and the gradient of $\\beta$ on intrinsic momentum transport in up-down asymmetric tokamaks
Ball, Justin; Lee, Jungpyo; Cerfon, Antoine J
Tokamaks with up-down asymmetric poloidal cross-sections spontaneously rotate due to turbulent transport of momentum. In this work, we investigate the effect of the Shafranov shift on this intrinsic rotation, primarily by analyzing tokamaks with tilted elliptical flux surfaces. By expanding the Grad-Shafranov equation in the large aspect ratio limit we calculate the magnitude and direction of the Shafranov shift in tilted elliptical tokamaks. The results show that, while the Shafranov shift becomes up-down asymmetric and depends strongly on the tilt angle of the flux surfaces, it is insensitive to the shape of the current and pressure profiles (when specific experimental parameters are kept fixed). Next, local nonlinear gyrokinetic simulations of these MHD equilibria are performed with GS2, which reveal that the Shafranov shift can significantly enhance the momentum transport. However, to be consistent, the effect of $\\beta'$ (i.e. the radial gradient of $\\beta$) on the magnetic equilibrium was also included....
Electron cyclotron emission imaging diagnostic system for Rijnhuizen Tokamak Project
Energy Technology Data Exchange (ETDEWEB)
Deng, B.H.; Hsia, R.P.; Domier, C.W.; Burns, S.R.; Hillyer, T.R.; Luhmann, N.C. Jr. [University of California at Davis, 228 Walker Hall, Davis, California 95616 (United States); Oyevaar, T.; Donne, A.J. [FOM-Inst. voor Plasmafysica Rijnhuizen, Association Euratom-FOM (International organizations without location); RTP team
A 16-channel electron cyclotron emission (ECE) imaging diagnostic system has been developed and installed on the Rijnhuizen Tokamak Project for measuring plasma electron cyclotron emission with a temporal resolution of 2 {mu}s. The high spatial resolution of the system is achieved by utilizing a low cost linear mixer/receiver array. Unlike conventional ECE diagnostics, the sample volumes of the ECE imaging system are aligned vertically, and can be shifted across the plasma cross-section by varying the local oscillator frequency, making possible 2D measurements of electron temperature profiles and fluctuations. The poloidal/radial wavenumber spectra and correlation lengths of T{sub e} fluctuations in the plasma core can also be obtained by properly positioning the focal plane of the imaging system. Due to these unique features, ECE imaging is an ideal tool for plasma transport study. Technical details of the system are described, together with preliminary experimental results. {copyright} {ital 1999 American Institute of Physics.}
Maintenance features of the Compact Ignition Tokamak fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Spampinato, P.T.; Hager, E.R.
The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs.
Gyrokinetic modelling of stationary electron and impurity profiles in tokamaks
Skyman, Andreas; Tegnered, Daniel
Particle transport due to Ion Temperature Gradient/Trapped Electron (ITG/TE) mode turbulence is investigated using the gyrokinetic code GENE. Both a reduced quasilinear (QL) treatment and nonlinear (NL) simulations are performed for typical tokamak parameters corresponding to ITG dominated turbulence. A selfconsistent treatment is used, where the stationary local profiles are calculated corresponding to zero particle flux simultaneously for electrons and trace impurities. The scaling of the stationary profiles with magnetic shear, safety factor, electron-to-ion temperature ratio, collisionality, toroidal sheared rotation, triangularity, and elongation is investigated. In addition, the effect of different main ion mass on the zero flux condition is discussed. The electron density gradient can significantly affect the stationary impurity profile scaling. It is therefore expected, that a selfconsistent treatment will yield results more comparable to experimental results for parameter scans where the stationary b...
Simulations of burn dynamics in tokamak fusion reactors
Energy Technology Data Exchange (ETDEWEB)
Mantsinen, M.J.; Salomaa, R.R.E.
The global dynamics of tokamak reactors is investigated with the time-dependent, volume-averaged (0D) particle and power balance code FRESCO (Fusion REactor Simulation COde). The main emphasis is on studies of reactivity transients during tokamak start-up and shut down, as well as after sudden changes in plasma and tokamak parameters. In particular, the plasma responses to changes in the confinement, fuelling rates and impurity concentrations are considered. 76 refs.
Conditioning of the vacuum chamber of the Tokamak Novillo; Acondicionamiento de la camara de vacio del Tokamak Novillo
Energy Technology Data Exchange (ETDEWEB)
Valencia A, R.; Lopez C, R.; Melendez L, L.; Chavez A, E.; Colunga S, S.; Gaytan G, E
The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10{sup -7} Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)
Module description of TOKAMAK equilibrium code MEUDAS
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)
Rapidly Moving Divertor Plates In A Tokamak
Energy Technology Data Exchange (ETDEWEB)
S. Zweben
It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.
MHD stable regime of the tokamak
Energy Technology Data Exchange (ETDEWEB)
Cheng, C.Z.; Furth, H.P.; Boozer, A.H.
A broad family of tokamak current profiles is found to be stable against ideal and resistive MHD kink modes for 1 less than or equal to q(0), with q(a) as low 2. For 0.5 less than or equal to q(0) < and q(a) > 1, current profiles can be found that are unstable only to the m = 1, n = 1 mode. A specific ''optimal'' tokamak profile can be selected from the range of stable solutions, by imposing a common upper limit on dj/dr - corresponding in ohmic equilibrium to a limitation of dT/sub e//dr by anomalous transport.
Boundary Plasma Turbulence Simulations for Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Xu, X; Umansky, M; Dudson, B; Snyder, P
The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.
Magnetic sensor for steady state tokamak
Energy Technology Data Exchange (ETDEWEB)
Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).
Transforming Developmental Education in Texas
Journal of Developmental Education, 2014
In recent years, with support from the Texas Legislature, the Texas Higher Education Coordinating Board has funded various developmental education initiatives, including research and evaluation efforts, to help Texas public institutions of higher education provide more effective programs and services to underprepared students. Based on evaluation…
Kerry J.W. Dooley
This resource update provides an overview of forest resources in east Texas based on an inventory conducted by the U.S. Forest Service, Forest Inventory and Analysis (FIA) program at the Southern Research Station (SRS) in cooperation with Texas A&M Forest Service. The 254 counties of Texas are consolidated into seven FIA survey unitsâSoutheast (unit 1),...
Tokamak turbulence with stochastic field lines
Energy Technology Data Exchange (ETDEWEB)
Beyer, P.; Garbet, X.; Ghendrih, Ph
Three-dimensional numerical simulations of ballooning turbulence in a tokamak plasma with stochastic magnetic field lines are presented. Three main features are observed. First, the level of pressure fluctuations decreases in the ergodic layer. Secondly, this is essentially due to a suppression of large scale structures. Finally, the turbulent heat diffusivity does not decrease in the stochastic are due to an increase of electric fluctuations. These observations are in agreement with turbulence measurements on Tore Supra. (author) 27 refs.
Microinstabilities in weak density gradient tokamak systems
Energy Technology Data Exchange (ETDEWEB)
Tang, W.M.; Rewoldt, G.; Chen, L.
A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient.
Initial DEMO tokamak design configuration studies
Energy Technology Data Exchange (ETDEWEB)
Bachmann, Christian, E-mail: firstname.lastname@example.com [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others
Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.
Energy Technology Data Exchange (ETDEWEB)
OAK-B135 The pressure profile and plasma shape, parameterized by elongation ({kappa}), triangularity ({delta}), and squareness ({zeta}), strongly influence stability. In this study, ideal stability of single null and symmetric, double-null, advanced tokamak (AT) configurations is examined. All the various shapes are bounded by a common envelope and can be realized in the DIII-D tokamak. The calculated AT equilibria are characterized by P{sub 0}/
{approx} 2.0-4.5, weak negative central shear, high q{sub min} (> 2.0), high bootstrap fraction, an H-mode pedestal, and varying shape parameters. The pressure profile is modeled by various polynomials together with a hyperbolic tangent pedestal, consistent with experimental observations. Stability is calculated with the DCON code and the resulting stability boundary is corroborated by GATO runs.
Effect of Profiles and Space on Ideal Stability of Advanced Tokamak Equilibria
Energy Technology Data Exchange (ETDEWEB)
Makowski, M A; Casper, T A; Ferron, J R; Taylor, T S; Turnbull, A D
The pressure profile and plasma shape, parameterized by elongation ({kappa}), triangularity ({delta}), and squareness ({zeta}), strongly influence stability. In this study, ideal stability of single null and symmetric, double-null, advanced tokamak (AT) configurations is examined. All the various shapes are bounded by a common envelope and can be realized in the DIII-D tokamak. The calculated AT equilibria are characterized by P{sub 0}/{l_angle}P{r_brace} {approx} 2.0-4.5, weak negative central shear, high q{sub min} (>2.0), high bootstrap fraction, an H-mode pedestal, and varying shape parameters. The pressure profile is modeled by various polynomials together with a hyperbolic tangent pedestal, consistent with experimental observations. Stability is calculated with the DCON code and the resulting stability boundary is corroborated by GATO runs.
Kinetic simulations of scrape-off layer physics in the DIII-D tokamak
Directory of Open Access Journals (Sweden)
R.M. Churchill
The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels (Mi ∼ 0.5, with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/Te∼3−4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. These two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.
Laser cleaning of diagnostic mirrors from tokamak-like carbon contaminants
Energy Technology Data Exchange (ETDEWEB)
Maffini, A., E-mail: email@example.com [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Uccello, A. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Dellasega, D. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, EURATOM-ENEA-CNR Association, Milan (Italy); Russo, V. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Perissinotto, S. [Center for Nano Science and Technology @ Polimi, Istituto Italiano di Tecnologia, Milan (Italy); Passoni, M. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, EURATOM-ENEA-CNR Association, Milan (Italy)
This paper presents a laboratory-scale experimental investigation of laser cleaning of diagnostic First Mirrors (FMs). Redeposition of contaminants sputtered from tokamak first wall onto FMs surface could dramatically decrease their reflectivity in an unacceptable way for the functioning of the plasma diagnostic systems. Laser cleaning is a promising solution to tackle this issue. In this work, pulsed laser deposition was exploited to produce rhodium films functional as FMs and to deposit onto them carbon contaminants with tailored features, resembling those found in tokamaks. The same laser system was also used to perform laser cleaning experiments by means of a sample handling procedure that allows to clean some cm{sup 2} in few minutes. The cleaning effectiveness was evaluated in terms of specular reflectivity recovery and mirror surface integrity. The effect of different laser wavelengths (λ = 1064, 266 nm) on the cleaning process is also addressed.
Thermodynamic approach to the interpretation of self-consistent pressure profiles in a tokamak
Dyabilin, K. S.; Razumova, K. A.
The phenomenon of invariable pressure profiles in tokamaks is interpreted in the framework of the thermodynamic approach suggesting that invariable self-consistent states correspond to the minimum of free energy. Solutions qualitatively consistent with the experiment are obtained under the assumption that the mechanism for the formation of self-consistent profiles is directly related to equilibrium diamagnetic currents. The dynamics of the system and specific transport phenomena, such as energy and particle pinching and a decrease in the local density under auxiliary electron cyclotron resonance heating (density pump-out), are analyzed in the vicinity of an equilibrium state characterized by a stable pressure profile. The scaling for the energy confinement time deduced from the transport model agrees qualitatively with the ITER scaling based on the analysis of experimental data obtained in many tokamaks. The possibility of using generalized Tsallis statistics to analyze pressure profiles is considered.
Study of electron beams within ISTTOK tokamak by means of a multi-channel Cherenkov detector; their correlation with hard X-rays
Energy Technology Data Exchange (ETDEWEB)
Jakubowski, L., E-mail: firstname.lastname@example.com [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Malinowski, K.; Sadowski, M.J.; Zebrowski, J. [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Plyusnin, V.V. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Rabinski, M. [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Jakubowski, M.J. [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland)
The paper describes experimental studies of electron beams emitted from a plasma torus within the ISTTOK tokamak, which were performed by means of a new four-channel detector of the Cherenkov type. A range of electron energy was estimated. There were also measured hard X-rays, and their correlation with the fast run-away electron beams was investigated experimentally.
Neutral particle dynamics in the Alcator C-Mod tokamak
Energy Technology Data Exchange (ETDEWEB)
Niemczewski, Artur P. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)
This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism.
Isotope effects on L-H threshold and confinement in tokamak plasmas
Maggi, C. F.; Weisen, H.; Hillesheim, J. C.; Chankin, A.; Delabie, E.; Horvath, L.; Auriemma, F.; Carvalho, I. S.; Corrigan, G.; Flanagan, J.; Garzotti, L.; Keeling, D.; King, D.; Lerche, E.; Lorenzini, R.; Maslov, M.; Menmuir, S.; Saarelma, S.; Sips, A. C. C.; Solano, E. R.; Belonohy, E.; Casson, F. J.; Challis, C.; Giroud, C.; Parail, V.; Silva, C.; Valisa, M.; Contributors, JET
The dependence of plasma transport and confinement on the main hydrogenic ion isotope mass is of fundamental importance for understanding turbulent transport and, therefore, for accurate extrapolations of confinement from present tokamak experiments, which typically use a single hydrogen isotope, to burning plasmas such as ITER, which will operate in deuterium–tritium mixtures. Knowledge of the dependence of plasma properties and edge transport barrier formation on main ion species is critical in view of the initial, low-activation phase of ITER operations in hydrogen or helium and of its implications on the subsequent operation in deuterium–tritium. The favourable scaling of global energy confinement time with isotope mass, which has been observed in many tokamak experiments, remains largely unexplained theoretically. Moreover, the mass scaling observed in experiments varies depending on the plasma edge conditions. In preparation for upcoming deuterium–tritium experiments in the JET tokamak with the ITER-like Be/W Wall (JET-ILW), a thorough experimental investigation of isotope effects in hydrogen, deuterium and tritium plasmas is being carried out, in order to provide stringent tests of plasma energy, particle and momentum transport models. Recent hydrogen and deuterium isotope experiments in JET-ILW on L-H power threshold, L-mode and H-mode confinement are reviewed and discussed in the context of past and more recent isotope experiments in tokamak plasmas, highlighting common elements as well as contrasting observations that have been reported. The experimental findings are discussed in the context of fundamental aspects of plasma transport models.
Study of heat and synchrotron radiation transport in fusion tokamak plasmas. Application to the modelling of steady state and fast burn termination scenarios for the international experimental fusion reactor ITER
Energy Technology Data Exchange (ETDEWEB)
Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)
The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment
Vail, Kathleen
Santa Fe School Superintendent Yvonne Gonzales, the "Texas Tornado," was hired to fix a 40% student-dropout rate and a white/Hispanic gap in achievement test scores. Gonzales is an avid integrationist; relies on humor, appeasement, and persuasion tactics; and has alienated some school employees by increasing central office…
Myers, Ray H.
In Dallas in 1970, high school outdoor education began as a cocurricular woods and waters boys' club sponsored by a community sportsman. Within one year, it grew into a fully accredited, coeducational, academic course with a curriculum devoted to the study of wildlife in Texas, ecology, conservation, hunting, firearm safety, fishing, boating and…
The electron cyclotron absorption diagnostic at the Rijnhuizen tokamak project
van Gelder, J. F. M.; Miedema, H. S.; Donne, A. J. H.; Oomens, A. A. M.; Schüller, F. C.
A new 20-channel electron cyclotron absorption diagnostic has been developed at the Rijnhuizen tokamak project. It is the first time the electron pressure profile in a tokamak plasma can be measured directly with a time resolution of 1 ms. The diagnostic measures simultaneously the emission and
Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas
J.W. Haverkort (Willem)
htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma
Soft-X-Ray Tomography Diagnostic at the Rtp Tokamak
Da Cruz, D. F.; Donne, A. J. H.
An 80-channel soft x-ray tomography system has been constructed for diagnosing the RTP (Rijnhuizen Tokamak Project) tokamak plasma. Five pinhole cameras, each with arrays of 16 detectors are distributed more or less homogeneously around a poloidal plasma cross section. The cameras are positioned
Stability-transport modeling of the SINP tokamak discharges
Indian Academy of Sciences (India)
Nov 27, 2015 ... The code has been applied to follow the evolution of tokamak plasma discharges obtained in the Saha Institute of Nuclear Physics (SINP) tokamak. From these simulations, we have been able to identify the possible models of thermal conductivity, diffusion and impurity contents in these discharges. Effects ...
Equilibrium and stability of tokamak plasmas and accretion disks
Blokland, J.W.S.
In both fusion research as well in astrophysics, plasmas are widely studied. These plasmas can be found in different geometric configurations, such as in a tokamak, stellarator or in astrophysical jets, accretion disks, etc. In this thesis we focus on plasmas found in tokamaks or accretion disks. In
Dean, David Allen
Abstract This catalogue lists 1,084 species of spiders (three identified to genus only) in 311 genera from 53 families currently recorded from Texas and is based on the "Bibliography of Texas Spiders" published by Bea Vogel in 1970. The online list of species can be found at http://pecanspiders.tamu.edu/spidersoftexas.htm. Many taxonomic revisions have since been published, particularly in the families Araneidae, Gnaphosidae and Leptonetidae. Many genera in other families have been revised. The Anyphaenidae, Ctenidae, Hahniidae, Nesticidae, Sicariidae and Tetragnathidae were also revised. Several families have been added and others split up. Several genera of Corinnidae were transferred to Phrurolithidae and Trachelidae. Two genera from Miturgidae were transferred to Eutichuridae. Zoridae was synonymized under Miturgidae. A single species formerly in Amaurobiidae is now in the Family Amphinectidae. Some trapdoor spiders in the family Ctenizidae have been transferred to Euctenizidae. Gertsch and Mulaik started a list of Texas spiders in 1940. In a letter from Willis J. Gertsch dated October 20, 1982, he stated "Years ago a first listing of the Texas fauna was published by me based largely on Stanley Mulaik material, but it had to be abandoned because of other tasks." This paper is a compendium of the spiders of Texas with distribution, habitat, collecting method and other data available from revisions and collections. This includes many records and unpublished data (including data from three unpublished studies). One of these studies included 16,000 adult spiders belonging to 177 species in 29 families. All specimens in that study were measured and results are in the appendix. Hidalgo County has 340 species recorded with Brazos County at 323 and Travis County at 314 species. These reflect the amount of collecting in the area. PMID:27103878
ICPP: Results from the MAST Spherical Tokamak
Sykes, Alan
The MAST (Mega-Amp Spherical Tokamak) experiment is now fully operational, producing 1MA plasmas with MW level auxiliary heating from Neutral Beam Injection and 60GHz Electron Cyclotron Resonance Heating. Central electron and ion temperatures are both of order 1keV (measured by 30-point Thomson Scattering, Neutral Particle Analyzer and Charge-Exchange spectroscopy respectively). Following boronisation, the Greenwald density limit has been exceeded in double-null divertor discharges by 50operation has been achieved in both Ohmic and NBI heated plasmas. In addition to conventional plasma induction, MAST can employ the `merging-compression' scheme (pioneered on START) producing initial spherical tokamak plasmas of up to 0.5MA without use of flux from the central solenoid. The central solenoid can then be applied to further increase the current at ramp rates of up to 13MA/s; plasma current of 1MA is reached at only one-half of the full solenoid swing. Studies of strike point power loading in both Ohmic and beam heated plasmas have confirmed the result from START that the fraction of power loading on the inboard strike point is lower than predicted from simple models. Comprehensive arrays of halo detectors indicate tolerable levels of halo currents with low asymmetries; an encouraging result for the ST concept, and providing key data to test models. Results from MAST will be used both to extend the conventional tokamak database, and to determine the potential of the ST as a route to fusion power in its own right. Acknowledgement: this work is funded jointly by the UK Department of Trade and Industry and EURATOM. The NBI equipment is on loan from ORNL, the NPA from PPPL.
Correlation of electron beams and hard x-ray emissions in ISTTOK Tokamak
Energy Technology Data Exchange (ETDEWEB)
Jakubowski, L.; Malinowski, K.; Sadowski, M.J.; Zebrowski, J.; Rabinski, M.; Jakubowski, M.J. [National Centre for Nuclear Research (NCBJ), Otwock (Poland); Plyusnin, V.V.; Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal)
The paper reports on experimental studies of electron beams in the ISTTOK tokamak, those were performed by means of an improved four-channel detector. The Cherenkov-type detector measuring head was equipped with four radiators made of two types of alumina-nitrate (AlN) poly-crystals: machinable and translucent ones, both of 10 mm in diameter and 2.5 mm in thickness. The movable support that enabled the whole detectors to be placed inside the tokamak vacuum chamber, at chosen positions along the ISTTOK minor radius. Since the electron energy distribution is one of the most important characteristics of tokamak plasmas, the main aim of the study was to perform estimations of an energy spectrum of the recorded electrons. For this purpose the radiators were coated with molybdenum (Mo) layers of different thickness. The technique based on the use of Cherenkov-type detectors enabled the detection of fast electrons (of energy above 66 keV) and determination of their spatial and temporal characteristics in the ISTTOK experiment. Measurements of hard X-rays (HXR), which were emitted during ISTTOK discharges, have also been performed. Particular attention was paid to the correlation measurements of HXR pulses with run-away electron beams. (copyright 2013 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Improved Confinement Induced by CT Injection in the STOR-M Tokamak
Xiao, C.; Hirose, A.; Zawalski, W.; Jin, T.; Ding, W.; White, D.; Raman, R.; Decoste, R.; Gregory, B. C.; Martin, F.
Compact torus (CT) injection as a means to centrally fuel a large tokamak is an attractive alternative. The University of Saskatchewan Compact Torus Injector (USCTI) is designed to study the interaction between the compact torus and tokamak plasmas. USCTI has been modified to include chromium coated copper outer electrodes and to allow baking at higher temperatures up to 100 ^circC. As a result of the modification and parameter optimization, CTs can be formed at a lower gas injection throughput (80% lower than before). The CT velocity has also been increased from 120 km/sec to 180 km/sec. Disruption-free tangential CT injection has been achieved. CT injection triggers Ohmic H-mode which lasts for 4 msec. The H-mode triggered by CT injection is characterized by an increase in the line averaged density, sudden drop in H_α emission, sudden increase in floating potential at the edge and SOL, decrease in floating potential and m=2 MHD fluctuations, and steepening of edge density profile. In the next experimental campaign, possibility of momentum transfer from energetic CT to tokamak discharge will be investigated.
REVIEW ARTICLE Tokamak equilibria with nearly zero central current: the current hole
Fujita, Takaaki
The observation of stable sustainment of the 'current hole', namely the nearly zero current density region in the central part of a tokamak plasma, has opened a new class of configurations in tokamak plasmas, and a variety of research from the viewpoints of equilibrium, magnetohydrodynamics (MHD) stability, particle orbits and radial transport has been generated. Some theories and codes have been tested and extended by being applied to extreme conditions in the current hole with very weak poloidal field. The current hole is generated due to a transient negative toroidal electric field established when a large off-axis non-inductive current is rapidly formed. It has been observed in high confinement plasmas with a large fraction of bootstrap current in advanced tokamak operation. The current hole is very stiff against current drive, which suggests that it is a saturated or self-organized system. Appearance of the current hole in ITER and DEMO would be expected in some of the operation scenarios, and its influence and its control methods have been studied. Results of experimental and theoretical studies on the current hole are reviewed.
Study on the key technologies of the Transfer Equipment Cask for Tokamak Equator Port Plug
Energy Technology Data Exchange (ETDEWEB)
Wang, Buyun, E-mail: email@example.com [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Gao, Lifu [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Cao, Huibin; Sun, Jian [Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Sun, Yuxiang; Song, Quanjun; Ma, Chengxue; Chang, Li; Shuang, Feng [Department of Automation, University of Science and Technology of China, Hefei, Anhui 230027 (China); Robot Sensors and Human-Machine Interaction Laboratory, Institute of Intelligent Machines, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)
Highlights: • Design on Intelligent Air Transfer System (IATS) for Transfer Equipment Cask (TECA). • A rhombic-like parallel robot for docking with minimum misalignment. • Design on electro-hydraulic servo system of the TECA for Tokamak Equator Port Plug (TEPP) manipulation. • A control architecture with several algorithms and information acquired from sensors could be used by the TECA for Remote Handling (RH). - Abstract: The Transfer Equipment Cask (TECA) is a key solution for Remote Handling (RH) in Tokamak Equator Port Plug (TEPP) operations. From the perspectives of both engineering and technical designs of effective experiments on the TEPP, key technologies on these topics covering the TECA are required. According to conditions in ITER (International Thermonuclear Experimental Reactor) and features of the TEPP, this paper introduces the design of an Intelligent Air Transfer System (IATS) with an adaptive attitude and high precision positioning that transports a cask system of more than 30 tons from the Tokamak Building (TB) to the Hot Cell Building (HCB). Additionally, different actuators are discussed, and the hydraulic power drive is eventually selected and designed. A rhombic-like parallel robot is capable of being used for docking with minimum misalignment. Practical mechanisms of the cask system are presented for hostile environments. A control architecture with several algorithms and information acquired from sensors could be used by the TECA. These designs yield realistic and extended applications for the RH of ITER.
Tungsten coating by ATC plasma spraying on CFC for WEST tokamak
Firdaouss, M.; Desgranges, C.; Hernandez, C.; Mateus, C.; Maier, H.; Böswirth, B.; Greuner, H.; Samaille, F.; Bucalossi, J.; Missirlian, M.
In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.
Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers
Energy Technology Data Exchange (ETDEWEB)
Maingi, Rajesh [North Carolina State Univ., Raleigh, NC (United States)
The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensional (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles.
Data acquisition and control system for the ECE imaging diagnostic on the EAST tokamak
Luo, C.; Lan, T.; Zhu, Y.; Xie, J.; Gao, B.; Liu, W.; Yu, C.; Milne, P. G.; Domier, C. W.; Luhmann, N. C.
A 384-channel electron cyclotron emission imaging (ECEI) system is installed on the experimental advanced superconducting tokamak (EAST) and 7-gigabyte data is produced for each regular discharge of a 10-second pulse. The data acquisition and control (DAC) system for the EAST ECEI diagnostics covers the large data production and embeds the ability to report the data quality instantly after the discharge. The symmetric routing design of the timing signal distributions among the 384 channels provides a low-cost solution to the synchronization of a large number of channels. The application of the load-balance bond service largely reduces the configuration difficulty and the cost in the high-speed data transferring tasks. Benefiting from the various kinds of hardware units with dedicated functionalities, an automated and user interactive DAC work flow is achieved, including the pre-selections of the automation scheme and the observation region, 384-channel data acquisition and local caching, post-discharge imaging data quality evaluation, remote system status monitoring, and inter-discharge imaging system event handling. The system configuration in a specific physics experiment is further optimized through the associated operating software which is enhanced by the input of the tokamak operation status and the region of interest (ROI) from other diagnostics. The DAC system is based on a modularized design and scalable to the long-pulse discharges in the EAST tokamak.
Development of a Cherenkov-type diagnostic system to study runaway electrons within the COMPASS tokamak
Rabinski, M.; Jakubowski, L.; Malinowski, K.; Sadowski, M. J.; Zebrowski, J.; Jakubowski, M. J.; Mirowski, R.; Weinzettl, V.; Ficker, O.; Mlynar, J.; Panek, R.; Paprok, R.; Vlainic, M.
Direct measurements of fast electrons, which are produced in high-temperature plasma and escape from tokamak-type facilities, are of particular interest for ITER and future fusion devices, where intense runaway electrons (RE) can significantly damage the first wall components. Therefore, the runaway control and mitigation based on credible measuring methods should be developed already in present devices. A team from the National Centre for Nuclear Research (NCBJ), Poland, developed special probes equipped with Cherenkov-type detectors for measurements of the fast electrons within edge plasmas of tokamaks. Studies of the fast runaway electrons were extensively carried out at the COMPASS tokamak at the Institute of Plasma Physics (IPP) in Prague during experimental campaigns in 2014–2016. In order to investigate an electron-beam energy distribution a three-channel probe equipped with the Cherenkov-type detectors sensitive to electrons of different energies has been constructed. The measurements performed by means of these detectors showed that the first fast electron peak appears usually in the current ramp-up phase, even before the hard X-rays (HXR) pulse. Some electron signals can also be observed during subsequent HXR emissions. However, the most distinct electron peaks in all energy channels appear mainly during the plasma disruption. A correlation of Cherenkov signals with the MHD activity was also studied.
Simulations of the operational control of a cryogenic plant for a superconducting burning-plasma tokamak
Mitchell, N
In recent proposals for next generation superconducting tokamaks, such as the ITER project, the nuclear burning plasma is confined by magnetic fields generated from a large set (up to 100 GJ stored energy) of superconducting magnets. These magnets suffer heat loads in operation from thermal and nuclear radiation from the surrounding components and plasma as well as eddy currents and AC losses generated within the magnets, together with the heat conduction through supports and resistive heat generated at the current lead transitions to room temperature. The initial cryoplant for such a tokamak is expected to have a steady state capacity of up to about 85 kW at 4.5 K, comparable to the system installed for LHC at CERN. Experimental tokamaks are expected to operate at least initially in a pulsed mode with 20-30 short plasma pulses and plasma burn periods each day. A conventional cryoplant, consisting of a cold box and a set of primary heat exchangers, is ill-suited to such a mode of operation as the instantaneou...
Upgrade of the TCV tokamak, first phase: Neutral beam heating system
Energy Technology Data Exchange (ETDEWEB)
Karpushov, Alexander N., E-mail: firstname.lastname@example.com [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Alberti, Stefano; Chavan, René [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Davydenko, Vladimir I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Duval, Basil P. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Ivanov, Alexander A. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Fasel, Damien; Fasoli, Ambrogio [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Gorbovsky, Aleksander I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Goodman, Timothy [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Kolmogorov, Vyacheslav V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Martin, Yves; Sauter, Olivier [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Sorokin, Aleksey V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); and others
Highlights: • Widening the parameter range of reactor relevant regimes on the TCV tokamak. • Installation of 1 MW, 30 keV neutral beam, direct ion heating, access to T{sub i}/T{sub e} ≥ 1. • ASTRA simulation of plasma response to NB and EC heating in different regimes. • Specific low divergency neutral beam injector with tunable beam power and energy. - Abstract: Experiments on TCV are designed to complement the work at large integrated tokamak facilities (such as JET) to provide a stepwise approach to extrapolation to ITER and DEMO in areas where medium-size tokamaks can often exploit their experimental capabilities and flexibility. Improving the understanding and control requirements of burning plasmas is a major scientific challenge, requiring access to plasma regimes and configurations with high normalized plasma pressure and a wide range of ion to electron temperature ratios, including T{sub e}/T{sub i} ∼ 1. These conditions will be explored by adding a 1 MW neutral heating beam to TCV's auxiliary for direct ion heating (2015) and increasing the ECH power injected in X-mode at the third harmonic (2 MW in 2015–2016). The manufacturing of the neutral beam injector was launched in 2014.
Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.
Energetic particles in spherical tokamak plasmas
McClements, K. G.; Fredrickson, E. D.
Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion-ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion
Differential and Integral Models of TOKAMAK
Directory of Open Access Journals (Sweden)
Ivo Dolezel
Full Text Available Modeling of 3D electromagnetic phenomena in TOKAMAK with typically distributed main and additional coils is not an easy business. Evaluated must be not only distribution of the magnetic field, but also forces acting in particular coils. Use of differential methods (such as FDM or FEM for this purpose may be complicated because of geometrical incommensurability of particular subregions in the investigated area or problems with the boundary conditions. That is why integral formulation of the problem may sometimes be an advantages. The theoretical analysis is illustrated on an example processed by both methods, whose results are compared and discussed.
[Injection of compact toroids for tokamak fueling and current drive]. Progress report, 1990--1991
Energy Technology Data Exchange (ETDEWEB)
Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.
The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.
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K.H. BURRELL; M.E. AUSTIN; D.P. BRENNAN; J.C. DeBOO; E.J. DOYLE; C. FENZI; C. FUCHS; P. GOHIL; R.J. GROEBNER; L.L. LAO; T.C. LUCE; M.A. MAKOWSKI; G.R. McKEE; R.A. MOYER; C.C. PETTY; M. PORKOLAB; C.L.RETTIG; T.L. RHODES; J.C. ROST; B.W. STALLARD; E.J. STRAIT; E.J. SYNAKOWSKI; M.R. WADE; J.G. WATKINS; W.P. WEST
High confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J.L. Luxon, et al., Plasma Phys. and Contr. Nucl. Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987) Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation which is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 seconds or >25 energy confinement times {tau}{sub E}), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved {beta}{sub N} H{sub 89} = 7 for up to 5 times the {tau}{sub E} of 150 ms. The {beta}{sub N} H{sub 89} values of 7 substantially exceed the value of 4 routinely achieved in standard ELMing H-mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H-mode is possible because of the presence of an edge magnetic hydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.
Quiescent Double Barrier H-Mode Plasmas in the DIII-D Tokamak
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Burrell, K H; Austin, M E; Brennan, D P; DeBoo, J C; Doyle, E J; Fenzi, C; Fuchs, C; Gohil, P; Greenfield, C M; Groebner, R J; Lao, L L; Luce, T C; Makowski, M A; McKee, G R; Moyer, R A; Petty, C C; Porkolab, M; Rettig, C L; Rhodes, T L; Rost, J C; Stallard, B W; Strait, E J; Synakowski, E J; Wade, M R; Watkins, J G; West, W P
High confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation which is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 seconds or >25 energy confinement times {tau}{sub E}), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved {beta}{sub N} H89 = 7 for up to 5 times the {tau}{sub E} of 150 ms. The {beta}{sub N} H89 values of 7 substantially exceed the value of 4 routinely achieved in standard ELMing H-mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H-mode is possible because of the presence of an edge magnetic hydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.
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Data for Texas abandoned oil fields were primarily derived from two sources: (1) Texas Railroad Commission (TRRC), and (2) Dwight's ENERGYDATA. For purposes of this report, abandoned oil fields are defined as those fields that had no production during 1977. The TRRC OILMASTER computer tapes were used to identify these abandoned oil fields. The tapes also provided data on formation depth, gravity of oil production, location (both district and county), discovery date, and the cumulative production of the field since its discovery. In all, the computer tapes identified 9211 abandoned fields, most of which had less than 250,000 barrel cumulative production. This report focuses on the 676 abandoned onshore Texas oil fields that had cumulative production of over 250,000 barrels. The Dwight's ENERGYDATA computer tapes provided production histories for approximately two-thirds of the larger fields abandoned in 1966 and thereafter. Fields which ceased production prior to 1966 will show no production history nor abandonment date in this report. The Department of Energy hopes the general availability of these data will catalyze the private sector recovery of this unproduced resource.
Tokamak experimental power reactor conceptual design. Volume II
Energy Technology Data Exchange (ETDEWEB)
Volume II contains the following appendices: (1) summary of EPR design parameters, (2) impurity control, (3) plasma computational models, (4) structural support system, (5) materials considerations for the primary energy conversion system, (6) magnetics, (7) neutronics penetration analysis, (8) first wall stress analysis, (9) enrichment of isotopes of hydrogen by cryogenic distillation, and (10) noncircular plasma considerations. (MOW)
Spherical tokamak Globus-M2: design, integration, construction
Minaev, V. B.; Gusev, V. K.; Sakharov, N. V.; Varfolomeev, V. I.; Bakharev, N. N.; Belyakov, V. A.; Bondarchuk, E. N.; Brunkov, P. N.; Chernyshev, F. V.; Davydenko, V. I.; Dyachenko, V. V.; Kavin, A. A.; Khitrov, S. A.; Khromov, N. A.; Kiselev, E. O.; Konovalov, A. N.; Kornev, V. A.; Kurskiev, G. S.; Labusov, A. N.; Melnik, A. D.; Mineev, A. B.; Mironov, M. I.; Miroshnikov, I. V.; Patrov, M. I.; Petrov, Yu. V.; Rozhansky, V. A.; Saveliev, A. N.; Senichenkov, I. Yu.; Shchegolev, P. B.; Shcherbinin, O. N.; Shikhovtsev, I. V.; Sladkomedova, A. D.; Solokha, V. V.; Tanchuk, V. N.; Telnova, A. Yu.; Tokarev, V. A.; Tolstyakov, S. Yu.; Zhilin, E. G.
The Globus-M spherical tokamak has demonstrated practically all of the project objectives during the 15-year period of operation. The main factor limiting further progress in plasma performance is a relatively low toroidal magnetic field. The maximum toroidal magnetic field achieved on Globus-M was 0.4 T with the exception of a limited number of shots with 0.55 T, which led to damage of the toroidal field coil in 2002. The increase of the magnetic field up to 1.0 T together with the plasma current up to 0.5 MA will result in the significant enhancement of the operating parameters in the upgraded Globus-M2 machine. The experimental program will be focused on plasma heating and non-inductive current drive and will contribute to the creation of a physical and technological base for the compact fusion neutron source development. In the article a brief overview of the physical background for the machine upgrade is outlined. The current status of the project implementation is described. First experimental results on moderate magnetic field increase from 0.4 T up to 0.5 T in the existing Globus-M machine are discussed. The improvement of plasma confinement as well as enhancement of efficiency of the beam driven current is observed.
Effects of electrode biasing in STOR-M Tokamak
Basu, Debjyoti; Nakajima, Masaru; Rohollahi, Akbar; McColl, David; Adegun, Joseph; Xiao, Chijin; Hirose, Akira
STOR-M is an iron-core, limiter based tokamak with major and minor radii of 46cm and 12 cm, respectively. Recently, electrode biasing experiments have been carried to study the improved confinement. For this purpose we have developed a DC power supply which can be gated by a high power SCR. The rectangular SS electrode has a height of 10 cm, a width of 2 cm and a thickness of 0.2 cm. The radial position of the electrode throughout the experiments is kept around 4mm inside the limiter in the plasma edge region. After application of positive bias with voltages between +90 V to +110 V during the plasma discharge current flat top with slightly higher edge-qa (nearly 5 to 6), noticeable increment of average plasma density and soft x-ray intensity along the central chord have been observed. No distinguishable change in H α emission has been measured. These phenomena may be attributed to improved confinement formed at the inner region but not at the edge. In the upcoming experimental campaign, Ion Doppler spectroscopy will be used to measure possible velocity shear inside the inner plasma region. Edge plasma pressure gradient will also be measured using Langmuir probes. Detailed experimental results will be presented.
Issues in tokamak/stellarator transport and confinement enhancement mechanisms
Energy Technology Data Exchange (ETDEWEB)
Perkins, F.W.
At present, the mechanism for anomalous energy transport in low-{beta} toroidal plasmas -- tokamaks and stellarators -- remains unclear, although transport by turbulent E {times} B velocities associated with nonlinear, fine-scale microinstabilities is a leading candidate. This article discusses basic theoretical concepts of various transport and confinement enhancement mechanisms as well as experimental ramifications which would enable one to distinguish among them and hence identify a dominant transport mechanism. While many of the predictions of fine-scale turbulence are born out by experiment, notable contradictions exist. Projections of ignition margin rest both on the scaling properties of the confinement mechanism and on the criteria for entering enhanced confinement regimes. At present, the greatest uncertainties lie with the basis for scaling confinement enhancement criteria. A series of questions, to be answered by new experimental/theoretical work, is posed to resolve these outstanding contradictions (or refute the fine-scale turbulence model) and to establish confinement enhancement criteria. 73 refs., 4 figs., 5 tabs.
Development of atomic beam probe for tokamaks
Energy Technology Data Exchange (ETDEWEB)
Berta, M., E-mail: email@example.com [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)
Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.
System studies of compact ignition tokamaks
Energy Technology Data Exchange (ETDEWEB)
Galambos, J.D.; Blackfield, D.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Selcow, E.
The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B/sup 2/a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations.
The Spherical Tokamak MEDUSA for Costa Rica
Ribeiro, Celso; Vargas, Ivan; Guadamuz, Saul; Mora, Jaime; Ansejo, Jose; Zamora, Esteban; Herrera, Julio; Chaves, Esteban; Romero, Carlos
The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, Rphysics /technical related issues which will help all tasks of the very low aspect ratio stellarator SCR-1(A≡R/>=3.6, under design[2]) and also the ongoing activities in low temperature plasmas. Courses in plasma physics at undergraduate and post-graduate joint programme levels are regularly conducted. The scientific programme is intend to clarify several issues in relevant physics for conventional and mainly STs, including transport, heating and current drive via Alfv'en wave, and natural divertor STs with ergodic magnetic limiter[3,4]. [1] G.D.Garstka, PhD thesis, University of Wisconsin at Madison, 1997 [2] L.Barillas et al., Proc. 19^th Int. Conf. Nucl. Eng., Japan, 2011 [3] C.Ribeiro et al., IEEJ Trans. Electrical and Electronic Eng., 2012(accepted) [4] C.Ribeiro et al., Proc. 39^th EPS Conf. Contr. Fusion and Plasma Phys., Sweden, 2012
Spontaneous Toroidal Rotation in Tokamaks
Haines, Malcolm
When two-fluid MHD theory of stability is employed the resulting growth rates are complex, and the perturbing magnetic fields move with a velocity that depends both on the components of the electron drift and heat flux perpendicular to the equilibrium magnetic field and on the diamagnetic velocity. On diffusing into a resistive wall a drag force is exerted on the wall which is proportional to the square-root of the velocity of the perturbing fields. The equal and opposite force or torque will be on the plasma, centred at the singular rational surface for each mode[1]. For typical experimental conditions this leads to a spontaneous, or intrinsic toroidal rotation of 20km/s occurring in a few milliseconds for perturbing magnetic fields of 0.0025tesla. The induced poloidal rotation by this mechanism is generally much larger, but there is considerable poloidal damping due to trapped particles on the ion-ion collision time- scale[2]. Furthermore poloidal angular momentum is in general not conserved for an isolated plasma, and any up-down asymmetry can act as a source or sink[3]; for example, Pfirsch-Schluter diffusion [3 damping by trapped particles[2] and the Ware pinch[4]. [1] J.B.Taylor, Phys.Rev.Lett. 91, 115002 (2003). [2] R.C.Morris, M.G.Haines and R.J.Hastie, Phys.Plasmas 3, 4513 (1996). [3] M.G.Haines, Phys.Rev.Lett. 25, 1480 (1970). [4] M.G.Haines and P.Martin, Phys.Plasmas 3, 4536 (1996).
Comparison of two models for the X-ray dispersion produced in a Novillo Tokamak with measurements make with thermoluminescent dosemeters; Comparacion de dos modelos para la dispersion de rayos X producidos en un Novillo Tokamak con mediciones efectuadas con dosimetros termoluminiscentes
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Flores O, A.; Castillo, A.; Barocio, S.R.; Melendez L, L.; Chavez A, E.; Cruz C, G.J.; Lopez, R.; Olayo, M.G.; Gonzalez M, P. [Instituto Nacional de Investigaciones Nucleares, 52045 Salazar, Estado de Mexico (Mexico); Azorin N, J. [Universidad Autonoma Metropolitana Iztapalapa, 09340 Mexico D.F. (Mexico)
It was presented the results to study about the X-ray dispersion produced in the Novillo Tokamak using thermoluminescent dosemeters (DTL). The measurements were make in the equatorial plane of Tokamak, along twelve radial directions. The dispersion is observed due to the radiation interaction with walls surrounding the machine. It was proposed two types of heuristic mathematical methods for describing the X-ray dispersion, comparing them with the experimental data obtained with Dtl. The predictions of both models are adjusted well to the experimental data. (Author)
Solar Hot Water for Motor Inn--Texas City, Texas
Final report describes solar domestic-hot-water heater installation at LaQuinta Motor Inn, Texas City, Texas which furnished 63% of total hot-water load of new 98-unit inn. Report presents a description of system, drawings and photographs of collectors, operations and maintenance instructions, manufacturers' specifications for pumps, and an engineer's report on performance.
Partnership for Edge Physics (EPSI), University of Texas Final Report
Energy Technology Data Exchange (ETDEWEB)
Moser, Robert [Univ. of Texas, Austin, TX (United States); Carey, Varis [Univ. of Texas, Austin, TX (United States); Michoski, Craig [Univ. of Texas, Austin, TX (United States); Faghihi, Danial [Univ. of Texas, Austin, TX (United States)
Simulations of tokamak plasmas require a number of inputs whose values are uncertain. The effects of these input uncertainties on the reliability of model predictions is of great importance when validating predictions by comparison to experimental observations, and when using the predictions for design and operation of devices. However, high fidelity simulation of tokamak plasmas, particular those aimed at characterization of the edge plasma physics, are computationally expensive, so lower cost surrogates are required to enable practical uncertainty estimates. Two surrogate modeling techniques have been explored in the context of tokamak plasma simulations using the XGC family of plasma simulation codes. The first is a response surface surrogate, and the second is an augmented surrogate relying on scenario extrapolation. In addition, to reduce the costs of the XGC simulations, a particle resampling algorithm was developed, which allows marker particle distributions to be adjusted to maintain optimal importance sampling. This means that the total number of particles in and therefore the cost of a simulation can be reduced while maintaining the same accuracy.
Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR
Energy Technology Data Exchange (ETDEWEB)
Sa, Wanderley Pires de
The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.
Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors
Energy Technology Data Exchange (ETDEWEB)
Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others
The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.
Thomas J. Brandeis
This resource update provides an overview of forest resources in east Texas derived from an inventory conducted by the U.S. Forest Service, Forest Inventory and Analysis (FIA) Program at the Southern Research Station in cooperation with the Texas A&M Forest Service. These estimates are based on field data collected using the FIA annualized sample design and are...
Study of the L–I–H transition with a new dual gas puff imaging system in the EAST superconducting tokamak
DEFF Research Database (Denmark)
Xu, G.S.; Shao, L.M.; Liu, S.C.
The intermediate oscillatory phase during the L–H transition, termed the I-phase, is studied in the EAST superconducting tokamak using a newly developed dual gas puff imaging (GPI) system near the L–H transition power threshold. The experimental observations suggest that the oscillatory behaviour...
Neutral beam injector performance on the PLT and PDX tokamaks
Energy Technology Data Exchange (ETDEWEB)
Schilling, G.; Ashcroft, D.L.; Eubank, H.P.; Grisham, L.R.; Kozub, T.A.; Kugel, H.W.; Rossmassler, J.; Williams, M.D.
An overall injector system description is presented first, and this will be followed by a detailed discussion of those problems unique to multiple injector operation on the tokamaks, i.e., power transmission, conditioning, reliability, and failures.
Tokamak reactor cost model based on STARFIRE/WILDCAT costing
Energy Technology Data Exchange (ETDEWEB)
Evans, K. Jr.
A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately.
Three-Dimensional Analysis of Tokamaks and Stellarators
National Research Council Canada - National Science Library
Paul R. Garabedian
The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments...
Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1
Energy Technology Data Exchange (ETDEWEB)
This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided.
Plasma Turbulence in the Scrape-off Layer of the ISTTOK Tokamak
Jorge, Rogerio; Halpern, Federico D; Loureiro, Nuno F; Silva, Carlos
The properties of plasma turbulence in a poloidally limited scrape-off layer (SOL) are addressed, with focus on ISTTOK, a large aspect ratio tokamak with a circular cross section. Theoretical investigations based on the drift-reduced Braginskii equations are carried out through linear calculations and non-linear simulations, in two- and three-dimensional geometries. The linear instabilities driving turbulence and the mechanisms that set the amplitude of turbulence as well as the SOL width are identified. A clear asymmetry is shown to exist between the low-field and the high-field sides of the machine. A comparison between experimental measurements and simulation results is presented.
Contribution to the multi-machine pedestal scaling from the COMPASS tokamak
Komm, M.; Bílková, P.; Aftanas, M.; Berta, M.; Böhm, P.; Bogár, O.; Frassinetti, L.; Grover, O.; Háček, P.; Havlicek, J.; Hron, M.; Imríšek, M.; Krbec, J.; Mitošínková, K.; Naydenkova, D.; Pánek, R.; Peterka, M.; Snyder, P. B.; Stefanikova, E.; Stöckel, J.; Sos, M.; Urban, J.; Varju, J.; Vondráček, P.; Weinzettl, V.; the COMPASS Team
First systematic measurements of pedestal structure during Ohmic and NBI-assisted Type I ELMy H-modes were performed on the COMPASS tokamak in two dedicated experimental campaigns during 2015 and 2016. By adjusting the NBI heating and a toroidal magnetic field, the electron pedestal temperature was increased from 200 eV up to 300 eV, which allowed reaching pedestal collisionality ν \\text{ped}\\ast text{ped}\\ast . The pedestal pressure was successfully reproduced by the EPED model. The dependence of pedestal pressure width on ν \\text{ped}\\ast and β \\text{ped ~ }\\text{pol} is discussed.
3D passive stabilization of n = 0 MHD modes in EAST tokamak.
Chen, S L; Villone, F; Xiao, B J; Barbato, L; Luo, Z P; Liu, L; Mastrostefano, S; Xing, Z
Evidence is shown of the capability of non-axisymmetrical conducting structures in the Experimental Advanced Superconducting Tokamak (EAST) to guarantee the passive stabilization of the n = 0 MHD unstable mode. Suitable numerical modeling of the experiments allows a clear interpretation of the phenomenon. This demonstration and the availability of computational tools able to describe the effect of 3D conductors will have a huge impact on the design of future fusion devices, in which the conducting structures closest to plasma will be highly segmented.
Collisional damping of the fast magnetosonic wave in the tokamak edge plasma
Energy Technology Data Exchange (ETDEWEB)
Porkolab, M.; Bonoli, P.T. [MIT Plasma Fusion Center, Cambridge, Massachusetts 02139 (United States); Chiu, S.C. [General Atomics, San Diego, California 92186 (United States)
The collisional absorption of the fast magnetosonic wave in the tokamak edge region is re-examined. This is of concern in either fast wave current drive (FWCD) experiments with weak central absorption (i.e., DIII-D) or in high density minority heating experiments in compact, high field devices (i.e., Alcator C-Mod). Using a simple Krook-type of collision model, the present calculations indicate negligible (i.e., less than 0.1{percent}) single-pass absorption due to collisions under typical experimental conditions. {copyright} {ital 1996 American Institute of Physics.}
Observation of Edge Instability Limiting the Pedestal Growth in Tokamak Plasmas
Diallo, A.; Hughes, J. W.; Greenwald, M.; LaBombard, B.; Davis, E.; Baek, S.-G.; Theiler, C.; Snyder, P.; Canik, J.; Walk, J.; Golfinopoulos, T.; Terry, J.; Churchill, M.; Hubbard, A.; Porkolab, M.; Delgado-Aparicio, L.; Reinke, M. L.; White, A.; Alcator C-Mod Team
With fusion device performance hinging on the edge pedestal pressure, it is imperative to experimentally understand the physical mechanism dictating the pedestal characteristics and to validate and improve pedestal predictive models. This Letter reports direct evidence of density and magnetic fluctuations showing the stiff onset of an edge instability leading to the saturation of the pedestal on the Alcator C-Mod tokamak. Edge stability analyses indicate that the pedestal is unstable to both ballooning mode and kinetic ballooning mode in agreement with observations.
Geodesic acoustic modes in tokamak plasmas with a radial equilibrium electric field
Zhou, Deng
The dispersion relation of geodesic acoustic modes in the tokamak plasma with an equilibrium radial electric field is derived and analyzed. Multiple branches of eigenmodes have been found, similar to the result given by the fluid model with a poloidal mass flow. Frequencies and damping rates of both the geodesic acoustic mode and the sound wave increase with respect to the strength of radial electric field, while the frequency and the damping rate of the lower frequency branch slightly decrease. Possible connection to the experimental observation is discussed.
Plasma Current Start-up in a Spherical Tokamak
Mitarai, Osamu; Kessel, Charles; Hirose, Akira
The various plasma current start-up techniques and related topics in a spherical tokamak (ST) device are described. The Ohmic heating coil current clamp experiments in NSTX are described and discussed, and the plasma current start-up experiments in the STOR-M tokamak with iron core and the outer vertical field coil is presented as one of technique for a plasma current start-up in a ST.
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Green, David L [ORNL; Jaeger, E. F. [XCEL; Berry, Lee A [ORNL; Chen, Guangye [ORNL; Ryan, Philip Michael [ORNL; Canik, John [ORNL
Observations of improved radio frequency (RF) heating efficiency in high-confinement (H-) mode plasmas on the National Spherical Tokamak Experiment (NSTX) are investigated by whole-device linear simulation. We present the first full-wave simulation to couple kinetic physics of the well confined core plasma to the poorly confined scrape-off plasma. The new simulation is used to scan the launched fast-wave spectrum and examine the steady-state electric wave field structure for experimental scenarios corresponding to both reduced, and improved RF heating efficiency. We find that launching toroidal wave-numbers that required for fast-wave propagation excites large amplitude (kVm 1 ) coaxial standing modes in the wave electric field between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggest these modes are a probable cause of degraded heating efficiency. Also, the H-mode density pedestal and fast-wave cutoff within the confined plasma allow for the excitation of whispering gallery type eigenmodes localised to the plasma edge.
Understanding L-H transition in tokamak fusion plasmas
Xu, Guosheng; Wu, Xingquan
This paper reviews the current state of understanding of the L-H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L-H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L-H transition. We uncover a comprehensive physical picture of the L-H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L-H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.
Physical meaning of one-machine and multimachine tokamak scalings
Dnestrovskij, Yu. N.; Danilov, A. V.; Dnestrovskij, A. Yu.; Lysenko, S. E.; Ongena, J.
Specific features of energy confinement scalings constructed using different experimental databases for tokamak plasmas are considered. In the multimachine database, some pairs of engineering variables are collinear; e.g., the current I and the input power P both increase with increasing minor radius a. As a result, scalings derived from this database are reliable only for discharges in which such ratios as I/ a 2 or P/ a 2 are close to their values averaged over the database. The collinearity of variables allows one to exclude the normalized Debye radius d* from the scaling expressed in a nondimensional form. In one-machine databases, the dimensionless variables are functionally dependent, which allow one to cast a scaling without d*. In a database combined from two devices, the collinearity may be absent, so the Debye radius cannot generally be excluded from the scaling. It is shown that the experiments performed in support of the absence of d* in the two-machine scaling are unconvincing. Transformation expressions are given that allow one to compare experiments for the determination of scaling in any set of independent variables.
Modelling of electron transport and of sawtooth activity in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Angioni, C
PRETOR. This code has been further extended and applied to the simulation of electron transport in TGV. In simulating the electron temperature profile of Ohmic sawtoothing plasmas, the proper description of the current density profile and the sawtooth activity play the dominant role and not the specific transport model, provided that a single parameter in the model is adjusted to match the global plasma performance. In TGV discharges with electron cyclotron heating (ECH), the behaviour of the electron temperature exhibits some characteristics which have been recently observed to be common to several tokamaks. In particular, with central heating the electron temperature profile is stiff outside the power deposition region, that is the gradient scale length is independent of the heating power and essentially constant along the minor radius. With off-axis heating, transport is strongly reduced in the central region of the plasma, whereas a steep increase of the heat conductivity is observed at the power deposition location. Although the semi-empirical Rebut-Lallia-Watkins (RLW) transport model does not involve a critical gradient scale length, as the experimental observations would suggest, rather a critical electron temperature gradient, we have shown that it allows simulations which reproduce the described experimental features with very good agreement. Due to the relatively low toroidal magnetic field of TCV, the experimental temperature gradient with ECH exceeds by far the threshold included in the model. It can thus be stated that the parametric dependence of the electron heat conductivity of this transport model is adequate to reproduce the electron transport for plasma parameters in the operation domain of TCV. PRETOR, interfaced with the experimental data and the code TORAY-GA for the computation of the ECH source, has hence been used as a reliable tool for transport analysis and planning of new experiments. This has contributed to the identification of an improved
Diamagnetic loop measurement in Korea Superconducting Tokamak Advanced Research machine.
Bak, J G; Lee, S G; Kim, H S
Diamagnetic loop (DL), which consists of two poloidal loops inside the vacuum vessel, is used to measure the diamagnetic flux during a plasma discharge in the Korea Superconducting Tokamak Advanced Research (KSTAR) machine. The vacuum fluxes in the DL signal can be compensated up to 0.1 mWb by using the coefficients, which are obtained from experimental investigations, in the vacuum flux measurements during vacuum shots under same operational conditions of magnetic coils for plasma experiment in the KSTAR machine. The maximum error in the diamagnetic flux measurement due to the errors of the coefficients was estimated as ∼0.22 mWb. From the diamagnetic flux measurements for the ohmically heated circular plasmas in the KSTAR machine, the stored energy agrees well with the estimated kinetic energy within the discrepancy of 25%. When the electron cyclotron heating, the neutral beam injection, and the ion cyclotron resonance heating are added to the ohmically heated limiter plasmas, the additional heating effects can be clearly observed from the increase of the stored energy evaluated in the DL measurement. © 2011 American Institute of Physics
Photo-neutron Production on HT-7 Superconducting Tokamak
Zhu, Yubao
Experimental studies of photo-neutron production on HT-7 superconducting tokamak are presented. Time-resolved and spatial-distributed neutron fluxes are obtained using several polyethylene moderated BF3 and ^3He proportional counters as well as ZnS(Ag) scintillator. Comparisons of neutron production between helium and deuterium discharges are performed. Beside the commonly observed photo-neutron at the early times of plasma start-up and the late disruption stage, remarkable photo-neutrons are also observed on the discharges plateau period under low plasma density regime and non-inductively current driven conditions. The magnitude and time-evolution of neutron flux correlate very well with hard X-ray and γ emissions. Photo-neutron flux distribution has a characteristic of toroidal asymmetry, which implies the localization of photonuclear reactions. The analyses confirm that photo-neutron productions are closely related to plasma density, loop voltage, MHD instability, energetic particles, impurity population and plasma-wall interactions.
Neoclassical offset toroidal velocity and auxiliary ion heating in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Lazzaro, E., E-mail: email@example.com [Istituto di Fisica del Plasma CNR (Italy)
In conditions of ideal axisymmetry, for a magnetized plasma in a generic bounded domain, necessarily toroidal, the uniform absorption of external energy (e.g., RF or any isotropic auxiliary heating) cannot give rise to net forces or torques. Experimental evidence on contemporary tokamaks shows that the near central absorption of RF heating power (ICH and ECH) and current drive in presence of MHD activity drives a bulk plasma rotation in the co-I{sub p} direction, opposite to the initial one. Also the appearance of classical or neoclassical tearing modes provides a nonlinear magnetic braking that tends to clamp the rotation profile at the q-rational surfaces. The physical origin of the torque associated with P{sub RF} absorption could be due the effects of asymmetry in the equilibrium configuration or in power deposition, but here we point out also an effect of the response of the so-called neoclassical offset velocity to the power dependent heat flow increment. The neoclassical toroidal viscosity due to internal magnetic kink or tearing modes tends to relax the plasma rotation to this asymptotic speed, which in absence of auxiliary heating is of the order of the ion diamagnetic velocity. It can be shown by kinetic and fluid calculations, that the absorption of auxiliary power by ions modifies this offset proportionally to the injected power thereby forcing the plasma rotation in a direction opposite to the initial, to large values. The problem is discussed in the frame of the theoretical models of neoclassical toroidal viscosity.
Cooldown of the Compact Ignition Tokamak
Energy Technology Data Exchange (ETDEWEB)
Keeton, D.C.
Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.
Safety factor profile control in a tokamak
Bribiesca Argomedo, Federico; Prieur, Christophe
Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...
Decommissioning of the Tokamak Fusion Test Reactor
Energy Technology Data Exchange (ETDEWEB)
E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola
The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.
Self-organized stationary states of tokamaks
Jardin, Stephen
We report here on a nonlinear mechanism that forms and maintains a self-organized stationary (sawtooth free) state in tokamaks. This process was discovered by way of extensive long-time simulations using the M3D-C1 3D extended MHD code in which new physics diagnostics have been added. It is well known that most high-performance modes of tokamak operation undergo ``sawtooth'' cycles, in which the peaking of the toroidal current density triggers a periodic core instability which redistributes the current density. However, certain modes of operation are known, such as the ``hybrid'' mode in DIII-D, ASDEX-U, JT-60U and JET, and the long-lived modes in NSTX and MAST, which do not experience this cycle of instability. Empirically, it is observed that these modes maintain a non-axisymmetric equilibrium which somehow limits the peaking of the toroidal current density. The physical mechanism responsible for this has not previously been understood, but is often referred to as ``flux-pumping,'' in which poloidal flux is redistributed in order to maintain q0 >1. In this talk, we show that in long-time simulations of inductively driven plasmas, a steady-state magnetic equilibrium may be obtained in which the condition q0 >1 is maintained by a dynamo driven by a stationary marginal core interchange mode. This interchange mode, unstable because of the pressure gradient in the ultra-low shear region in the center region, causes a (1,1) perturbation in both the electrostatic potential and the magnetic field, which nonlinearly cause a (0,0) component in the loop voltage that acts to sustain the configuration. This hybrid mode may be a preferred mode of operation for ITER. We present parameter scans that indicate when this sawtooth-free operation can be expected.
Transport Barriers in Bootstrap Driven Tokamaks
Staebler, Gary
Maximizing the bootstrap current in a tokamak, so that it drives a high fraction of the total current, reduces the external power required to drive current by other means. Improved energy confinement, relative to empirical scaling laws, enables a reactor to more fully take advantage of the bootstrap driven tokamak. Experiments have demonstrated improved energy confinement due to the spontaneous formation of an internal transport barrier in high bootstrap fraction discharges. Gyrokinetic analysis, and quasilinear predictive modeling, demonstrates that the observed transport barrier is due to the suppression of turbulence primarily due to the large Shafranov shift. ExB velocity shear does not play a significant role in the transport barrier due to the high safety factor. It will be shown, that the Shafranov shift can produce a bifurcation to improved confinement in regions of positive magnetic shear or a continuous reduction in transport for weak or negative magnetic shear. Operation at high safety factor lowers the pressure gradient threshold for the Shafranov shift driven barrier formation. The ion energy transport is reduced to neoclassical and electron energy and particle transport is reduced, but still turbulent, within the barrier. Deeper into the plasma, very large levels of electron transport are observed. The observed electron temperature profile is shown to be close to the threshold for the electron temperature gradient (ETG) mode. A large ETG driven energy transport is qualitatively consistent with recent multi-scale gyrokinetic simulations showing that reducing the ion scale turbulence can lead to large increase in the electron scale transport. A new saturation model for the quasilinear TGLF transport code, that fits these multi-scale gyrokinetic simulations, can match the data if the impact of zonal flow mixing on the ETG modes is reduced at high safety factor. This work was supported by the U.S. Department of Energy under DE-FG02-95ER54309 and DE-FC02
Plasma Shape and Current Density Profile Control in Advanced Tokamak Operating Scenarios
Shi, Wenyu
The need for new sources of energy is expected to become a critical problem within the next few decades. Nuclear fusion has sufficient energy density to potentially supply the world population with its increasing energy demands. The tokamak is a magnetic confinement device used to achieve controlled fusion reactions. Experimental fusion technology has now reached a level where tokamaks are able to produce about as much energy as is expended in heating the fusion fuel. The next step towards the realization of a nuclear fusion tokamak power plant is ITER, which will be capable of exploring advanced tokamak (AT) modes, characterized by a high fusion gain and plasma stability. The extreme requirements of the advanced modes motivates researchers to improve the modeling of the plasma response as well as the design of feedback controllers. This dissertation focuses on several magnetic and kinetic control problems, including the plasma current, position and shape control, and data-driven and first-principles-driven modeling and control of plasma current density profile and the normalized plasma pressure ratio betaN. The plasma is confined within the vacuum vessel by an external electromagnetic field, produced primarily by toroidal and poloidal field coils. The outermost closed plasma surface or plasma boundary is referred to as the shape of the plasma. A central characteristic of AT plasma regimes is an extreme elongated shape. The equilibrium among the electromagnetic forces acting on an elongated plasma is unstable. Moreover, the tokamak performance is improved if the plasma is located in close proximity to the torus wall, which guarantees an efficient use of available volume. As a consequence, feedback control of the plasma position and shape is necessary. In this dissertation, an Hinfinity-based, multi-input-multi-output (MIMO) controller for the National Spherical Torus Experiment (NSTX) is developed, which is used to control the plasma position, shape, and X
Texas Affordable Baccalaureate Program: A Collaboration between the Texas Higher Education Coordinating Board, South Texas College, and Texas A&M University-Commerce. CBE Case Study
Klein-Collins, Rebecca; Glancey, Kathleen
This case study is part of a series on newer competency-based degree programs that have been emerging in recent years. In January 2014, the Texas Higher Education Coordinating Board (THECB), South Texas College (STC), and Texas A&M University-Commerce (A&M Commerce) launched the Texas Affordable Baccalaureate Program, the state's first…
Water supply and needs for West Texas
This presentation focused on the water supplies and needs of West Texas, Texas High Plains. Groundwater is the most commonly used water resources on the Texas High Plains, with withdrawals from the Ogallala Aquifer dominating. The saturation thickness of the Ogallala Aquifer in Texas is such that t...
Breeding, Seth D.
Floods occurred in Texas during, June, July, and November 1940 that exceeded known stages on many small streams and at a few places on the larger streams. Stages at several stream-gaging stations exceeded the maximum known at those places since the collection of daily records began. A storm, haying its axis generally on a north-south line from Cameron to Victoria and extending across the Brazos, Colorado, Lavaca, and Guadalupe River Basins, caused heavy rainfall over a large part of south-central Texas. The maximum recorded rain of 22.7 inches for the 2-day period June 29-30 occurred at Engle. Of this amount, 17.5 inches fell in the 12-hour period between 8 p.m. June 29, and 8 a.m. June 30. Light rains fell at a number of places on June 28, and additional light rains fell at many places within the area from July 1 to 4. During the period June 28 to July 4 more than 20 inches of rain fell over an area of 300 square miles, more than 15 inches over 1,920 square miles, and more than 10 inches over 5,100 square miles. The average annual rainfall for the area experiencing the heaviest rainfall during this storm is about 35 inches. Farming is largely confined to the fertile flood plains in much of the area subjected to the record-breaking floods in June and July. Therefore these floods, coming at the height of the growing season, caused severe losses to crops. Much damage was done also to highways and railways. The city of Hallettsville suffered the greatest damage of any urban area. The Lavaca River at that place reached a stage 8 feet higher than ever known before, drowned several people, destroyed many homes, and submerged almost the entire business district. The maximum discharge there was 93,100 second-feet from a drainage area of 101 square miles. Dry Creek near Smithville produced a maximum discharge of 1,879 second-feet from an area of 1.48 square miles and a runoff of 11.3 inches in a 2-day period from a rainfall of 19.5 inches. The area in the Colorado River
The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor
Energy Technology Data Exchange (ETDEWEB)
Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL
Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.
Evolution of plasma rotation, radial electric field, MHD activity and plasma confinement in the STOR M tokamak
Trembach, Dallas; Dreval, Mykola
Experimental results from the STOR-M tokamak detailing simultaneous behavior of plasma SOL rotation, radial electric field, main plasma column parameters, and MHD activity are presented. In the STOR-M tokamak, fast (˜ 1 ms), well correlated changes in the radial electric field, plasma rotation, and floating potential fluctuations in the periphery are observed. During the correlated phase, the radial electric field changes its sign from positive to negative, the Mach number of toroidal plasma rotation, which is co-current, decreases from M||= 0.4 to nearly 0. MHD activity in STOR-M tends to be suppressed if the radial electric field is negative. When the electric field is negative, MHD frequency decreases and increases in the average electron density and poloidal beta are observed.
Technical Training seminar: Texas Instruments
Monday 6 November TECHNICAL TRAINING SEMINAR 14:00 to 17:30 - Training Centre Auditorium (bldg. 593) Texas Instruments Technical Seminar Michael Scholtholt, Field Application Engineer / TEXAS INSTRUMENTS (US, D, CH) POWER - A short approach to Texas Instruments power products Voltage mode vs. current mode control Differentiating DC/DC converters by analyzing control and compensation schemes: line / load regulation, transient response, BOM, board space, ease-of-use Introduction to the SWIFT software FPGA + CPLD power solutions WIRELESS / CHIPCON Decision criteria when choosing a RF platform Introduction to Texas Instruments wireless products: standardized platforms proprietary platforms ( 2.4 GHz / sub 1 GHz) development tools Antenna design: example for 2.4 GHz questions, discussion Industrial partners: Robert Medioni, François Caloz / Spoerle Electronic, CH-1440 Montagny (VD), Switzerland Phone: +41 24 447 0137, email: email@example.com, http://www.spoerle.com Language: English. Free s...
Regional Finance Forum: Financing Resilient and Sustainable Water Infrastructure, held in Addison, Texas, September 10-11, 2015.Co-sponsored by EPA's Water Infrastructure and Resiliency Finance Center and the Environmental Finance Center Network.
2001 Harris County, Texas Lidar
National Oceanic and Atmospheric Administration, Department of Commerce — This data set was received by the NOAA Coastal Services Center from the Texas Natural Resources Information System. The data was collected in October of 2001 by...
Power supplies and quench protection for the Tokamak Physics Experiment
Energy Technology Data Exchange (ETDEWEB)
Neumeyer, C.L. [Raytheon Engineers & Constructors, Princeton, NJ (United States). EBASCO Div.
The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems.
Deflection of a liquid metal jet/drop in a tokamak environment
Energy Technology Data Exchange (ETDEWEB)
Pelekasis, Nikos, E-mail: email@example.com [Department of Mechanical Engineering, University of Thessaly, Volos 38334 (Greece); Benos, Lefteris [Department of Mechanical Engineering, University of Thessaly, Volos 38334 (Greece); Gomes, Rui [Associação EURATOM/IST, Centro de Fusão Nuclear, Av. Rovisco Pais, 1049-001 Lisboa (Portugal)
Highlights: • We model steady flow of a liquid metal jet inside an electromagnetic field in the presence of inertia and capillary forces. • Similar analysis is performed for the motion of a liquid metal spherical drop. • The deflection of the trajectory is predicted as a function of the intensity of the externally imposed magnetic and electric fields. • The analysis is used as a proof of principle study in reference to experimental observations of jet/drop deflection due to j{sup →}×B{sup →} effects in the ISTTOK tokamak. • We discuss the possibility of using liquid metal flows as an alternative approach toward enhancing power exhaust in tokamak facilities. - Abstract: The interaction of a liquid gallium jet with plasma has been investigated in the ISTTOK tokamak. The jet was observed to remain intact during its interaction with plasma, within a certain length beyond which drop formation was observed. Significant deflection of the jet was detected as soon as plasma production was started. Furthermore, a strong dependency of the deflection magnitude on plasma position was observed that could be correlated with plasma potential gradients. As a means to capture and, possibly, quantify this effect, a preliminary magnetohydrodynamic analysis was performed in order to predict the trajectory of a jet that is traveling inside an electromagnetic field. The effect of Lorentz forces, gravity and pressure drop are accounted for in a unidirectional model that assumes a small jet radius in comparison with the trajectory length. The effect of external electric potential gradients on jet deflection was ascertained in conjunction with the importance of electric stresses in modulating the jet speed and radius. Analysis of the results reported in the ISTTOK experiments identifies the process of jet break-up as a capillary instability. The trajectory of the ensuing droplets is modeled and intensification of the deflection process is predicted in the presence of Lorentz
Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks
Energy Technology Data Exchange (ETDEWEB)
R. Majeski
At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700oC. However, at a sufficiently high operating temperature (700 - 1000 oC), tungsten is selfannealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment. The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium. Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands1 and fusion experiments2,3 to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 oC), it has been now been used as a PFC in several confinement experiments (TFTR, T11- M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.
Aspects of Tokamak toroidal magnet protection
Energy Technology Data Exchange (ETDEWEB)
Green, R.W.; Kazimi, M.S.
Simple but conservative geometric models are used to estimate the potential for damage to a Tokamak reactor inner wall and blanket due to a toroidal magnet field collapse. The only potential hazard found to exist is due to the MHD pressure rise in a lithium blanket. A survey is made of proposed protection methods for superconducting toroidal magnets. It is found that the two general classifications of protection methods are thermal and electrical. Computer programs were developed which allow the toroidal magnet set to be modeled as a set of circular filaments. A simple thermal model of the conductor was used which allows heat transfer to the magnet structure and which includes the effect of temperature dependent properties. To be effective in large magnets an electrical protection system should remove at least 50% of the stored energy in the protection circuit assuming that all of the superconductor in the circuit quenches when the circuit is activated. A protection system design procedure based on this criterion was developed.
Tokamak blanket design study, final report
Energy Technology Data Exchange (ETDEWEB)
A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.
Fast scanning probe for tokamak plasmas
Boedo, J.; Gray, D.; Chousal, L.; Conn, R.; Hiller, B.; Finken, K. H.
We describe a fast reciprocating probe drive, which has three main new features: (1) a detachable and modular probe head for easy maintenance, (2) a combination of high heat flux capability, high bandwidth, and low-Z materials construction, and (3) low weight, compact, inexpensive construction. The probe is mounted in a fast pneumatic drive in order to reach plasma regions of interest and remain inserted long enough to obtain good statistics while minimizing the heat flux to the tips and head. The drive is pneumatic and has been designed to be compact and reliable to comply with space and maintenance requirements of tokamaks. The probe described here has five tips which obtain a full spectrum of plasma parameters: electron temperature profile Te(r), electron density profile ne(r), floating potential profile Vf(r), poloidal electric field profile Eθ(r), saturation current profile Isat(r), and their fluctuations up to 3 MHz. We describe the probe show radial profiles of various parameters. We compare the density and temperature data to that obtained with a helium beam. We also discuss the techniques to process the data optimally, particularly double probe data and profile fits.
Exploration of turbulent optimization in stellarators & tokamaks
Mynick, H.; Pomphrey, N.; Xanthopoulos, P.; Lucia, M.
A methodfootnotetextH.E. Mynick, N. Pomphrey, P. Xanthopoulos, Phys. Rev. Letters, 105, 095004 (2010).^,footnotetextH.E. Mynick, N. Pomphrey, P. Xanthopoulos, Phys. Plasmas, 18, 056101 (2011). recently developed for evolving toroidal configurations to ones with reduced turbulent transport, using the STELLOPT optimization codes and the GENE gyrokinetic code, is being applied and extended. The growing body of results has found that the effectiveness of the current proxy measure Qprox used by STELLOPT to estimate transport levels depends on the class of toroidal device considered. The present proxy works well for quasi-axisymmetric stellarators and tokamaks, modestly for quasi-helically symmetric designs, but not for the W7X quasi-omnigenous/quasi-isodynamic design. We are exploring the origin of this variation, and improving the dependence of the proxy on key geometric factors, extending the proxy to apply to transport channels other than the ITG turbulence it was originally developed for, and are also examining the relative effectiveness of different search algorithms. To help in these efforts, we have adapted STELLOPT to provide a new capability for mapping the topography of the cost function in the search space.
Detachment evolution on the TCV tokamak
Directory of Open Access Journals (Sweden)
J.R. Harrison
Full Text Available Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1]. Before the roll-over in the ion current to the outer strike point, C III and Dα emission from the outer leg recede slowly from the strike point toward the X-point, at a rate of ∼2.0 × 10−19m/m−3 along the magnetic field as the electron temperature along the leg reduces with increasing density. Around the onset of detachment, the upstream density profile and outer target Dα profiles broaden, possibly leading to an increase in radiation in the SOL by increased interaction between the SOL and the carbon tiles lining the outer wall. The plasma conditions upstream and at various locations along the detached outer divertor leg have been characterised, and the consistency of this data has been checked with the interpretive OSM-EIRENE-DIVIMP suite of codes [2] and are broadly found to be consistent with measured Dγ/Dα emissivity profiles along the detached outer divertor leg.
Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator
Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.
A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.
Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak.
Seo, Seong-Heon; Park, Jinhyung; Wi, H M; Lee, W R; Kim, H S; Lee, T G; Kim, Y S; Kang, Jin-Seob; Bog, M G; Yokota, Y; Mase, A
Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 10(19) m(-3) when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.
Preliminary conceptual design of a medium sized tokamak (IST-1)
Bagerpour, M.; Alinejad, N.; Sobhanian, S.
In this paper an attempt is made to estimate the main parameters of the Iranian superconducting tokamak as a medium sized tokamak. In the first stage, the production and confinement of ohmically heated plasma is considered. Considering the aim of the design and the kink stability limit, three main parameters are assumed to be known. Using the known theoretical, empirical scale laws and numerical solution of Grad-Shafranov equation for a D-shaped plasmas and also considering the correction terms due to triangularity of the torus cross section, other physical and geometrical parameters have been estimated. The magnetic flux surfaces, plasma pressure and toroidal current density profiles are found by solving of Grad-Shafranov equation as an eigenvalue problem using finite element method. The preliminary results are compared with some recent tokamaks now in operation in different research centers.
Sarkar, Sahotra; Strutz, Stavana E; Frank, David M; Rivaldi, Chissa-Louise; Sissel, Blake; Sánchez-Cordero, Victor
Chagas disease, caused by Trypanosoma cruzi, remains a serious public health concern in many areas of Latin America, including México. It is also endemic in Texas with an autochthonous canine cycle, abundant vectors (Triatoma species) in many counties, and established domestic and peridomestic cycles which make competent reservoirs available throughout the state. Yet, Chagas disease is not reportable in Texas, blood donor screening is not mandatory, and the serological profiles of human and canine populations remain unknown. The purpose of this analysis was to provide a formal risk assessment, including risk maps, which recommends the removal of these lacunae. The spatial relative risk of the establishment of autochthonous Chagas disease cycles in Texas was assessed using a five-stage analysis. 1. Ecological risk for Chagas disease was established at a fine spatial resolution using a maximum entropy algorithm that takes as input occurrence points of vectors and environmental layers. The analysis was restricted to triatomine vector species for which new data were generated through field collection and through collation of post-1960 museum records in both México and the United States with sufficiently low georeferenced error to be admissible given the spatial resolution of the analysis (1 arc-minute). The new data extended the distribution of vector species to 10 new Texas counties. The models predicted that Triatoma gerstaeckeri has a large region of contiguous suitable habitat in the southern United States and México, T. lecticularia has a diffuse suitable habitat distribution along both coasts of the same region, and T. sanguisuga has a disjoint suitable habitat distribution along the coasts of the United States. The ecological risk is highest in south Texas. 2. Incidence-based relative risk was computed at the county level using the Bayesian Besag-York-Mollié model and post-1960 T. cruzi incidence data. This risk is concentrated in south Texas. 3. The
Directory of Open Access Journals (Sweden)
Sahotra Sarkar
Full Text Available BACKGROUND: Chagas disease, caused by Trypanosoma cruzi, remains a serious public health concern in many areas of Latin America, including México. It is also endemic in Texas with an autochthonous canine cycle, abundant vectors (Triatoma species in many counties, and established domestic and peridomestic cycles which make competent reservoirs available throughout the state. Yet, Chagas disease is not reportable in Texas, blood donor screening is not mandatory, and the serological profiles of human and canine populations remain unknown. The purpose of this analysis was to provide a formal risk assessment, including risk maps, which recommends the removal of these lacunae. METHODS AND FINDINGS: The spatial relative risk of the establishment of autochthonous Chagas disease cycles in Texas was assessed using a five-stage analysis. 1. Ecological risk for Chagas disease was established at a fine spatial resolution using a maximum entropy algorithm that takes as input occurrence points of vectors and environmental layers. The analysis was restricted to triatomine vector species for which new data were generated through field collection and through collation of post-1960 museum records in both México and the United States with sufficiently low georeferenced error to be admissible given the spatial resolution of the analysis (1 arc-minute. The new data extended the distribution of vector species to 10 new Texas counties. The models predicted that Triatoma gerstaeckeri has a large region of contiguous suitable habitat in the southern United States and México, T. lecticularia has a diffuse suitable habitat distribution along both coasts of the same region, and T. sanguisuga has a disjoint suitable habitat distribution along the coasts of the United States. The ecological risk is highest in south Texas. 2. Incidence-based relative risk was computed at the county level using the Bayesian Besag-York-Mollié model and post-1960 T. cruzi incidence data. This
Nonlinear effects of energetic particle driven instabilities in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Bruedgam, Michael
In a tokamak plasma, a population of superthermal particles generated by heating methods can lead to a destabilization of various MHD modes. Due to nonlinear wave-particle interactions, a consequential fast particle redistribution reduces the plasma heating and can cause severe damages to the wall of the fusion device. In order to describe the wave-particle interaction, the drift-kinetic perturbative HAGIS code is applied which evolves the particle trajectories and the waves nonlinearly. For a simulation speed-up, the 6-d particle phase-space is reduced by the guiding centre approach to a 5-d description. The eigenfunction of the wave is assumed to be invariant, but its amplitude and phase is altered in time. A sophisticated {delta}/f-method is employed to model the change in the fast particle distribution so that numerical noise and the excessive number of simulated Monte-Carlo points are reduced significantly. The original code can only calculate the particle redistribution inside the plasma region. Therefore, a code extension has been developed during this thesis which enlarges the simulation region up to the vessel wall. By means of numerical simulations, this thesis addresses the problem of nonlinear waveparticle interactions in the presence of multiple MHD modes with significantly different eigenfrequencies and the corresponding fast particle transport inside the plasma. In this context, a new coupling mechanism between resonant particles and waves has been identified that leads to enhanced mode amplitudes and fast particle losses. The extension of the code provides for the first time the possibility of a quantitative and qualitative comparison between simulation results and recent measurements in the experiment. The findings of the comparison serve as a validation of both the theoretical model and the interpretation of the experimental results. Thus, a powerful interface tool has been developed for a deeper insight of nonlinear wave-particle interaction
Generalised ballooning theory of two-dimensional tokamak modes
Abdoul, P. A.; Dickinson, D.; Roach, C. M.; Wilson, H. R.
In this work, using solutions from a local gyrokinetic flux-tube code combined with higher order ballooning theory, a new analytical approach is developed to reconstruct the global linear mode structure with associated global mode frequency. In addition to the isolated mode (IM), which usually peaks on the outboard mid-plane, the higher order ballooning theory has also captured other types of less unstable global modes: (a) the weakly asymmetric ballooning theory (WABT) predicts a mixed mode (MM) that undergoes a small poloidal shift away from the outboard mid-plane, (b) a relatively more stable general mode (GM) balloons on the top (or bottom) of the tokamak plasma. In this paper, an analytic approach is developed to combine these disconnected analytical limits into a single generalised ballooning theory. This is used to investigate how an IM behaves under the effect of sheared toroidal flow. For small values of flow an IM initially converts into a MM where the results of WABT are recaptured, and eventually, as the flow increases, the mode asymptotically becomes a GM on the top (or bottom) of the plasma. This may be an ingredient in models for understanding why in some experimental scenarios, instead of large edge localised modes (ELMs), small ELMs are observed. Finally, our theory can have other important consequences, especially for calculations involving Reynolds stress driven intrinsic rotation through the radial asymmetry in the global mode structures. Understanding the intrinsic rotation is significant because external torque in a plasma the size of ITER is expected to be relatively low.
Impurity control in near-term tokamak reactors
Energy Technology Data Exchange (ETDEWEB)
Stacey, W.M. Jr.; Smith, D.L.; Brooks, J.N.
Several methods for reducing impurity contamination in near-term tokamak reactors by modifying the first-wall surface with a low-Z or low-sputter material are examined. A review of the sputtering data and an assessment of the technological feasibility of various wall modification schemes are presented. The power performance of a near-term tokamak reactor is simulated for various first-wall surface materials, with and without a divertor, in order to evaluate the likely effect of plasma contamination associated with these surface materials.
Geodesic acoustic modes in noncircular cross section tokamaks
Energy Technology Data Exchange (ETDEWEB)
Sorokina, E. A., E-mail: firstname.lastname@example.com; Lakhin, V. P. [National Research Center "Kurchatov Institute," (Russian Federation); Konovaltseva, L. V. [People's Friendship University of Russia (Russian Federation); Ilgisonis, V. I. [National Research Center "Kurchatov Institute," (Russian Federation)
The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.
Advanced tokamak physics scenarios in Alcator C-Mod
Energy Technology Data Exchange (ETDEWEB)
Porkolab, M.; Bonoli, P.T.; Golovato, S.; Ramos, J.; Sugiyama, L.; Takase, Y. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Kessel, C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Nevins, W.M. [LLNL, Livermore, California 94550 (United States)
Several advanced tokamak modes of operation have been identified in the Alcator C-Mod tokamak. Of particular interest are (i) Reversed shear mode with high bootstrap fraction using on-axis FW current drive and off-axis mode-conversion current drive and/or lower hybrid current drive; (ii) High performance plasmas ({ital Q}{approximately}0.1{endash}1) which may be accessed by the PEP (pellet enhanced performance) mode of operation with intense ICRF heating. {copyright} {ital 1996 American Institute of Physics.}
Electromagneto-mechanical coupling analysis of a test module in J-TEXT Tokamak during plasma disruption
Energy Technology Data Exchange (ETDEWEB)
Dong, Haijie; Yuan, Zhensheng; Yuan, Hongwei; Pei, Cuixiang [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi'an Jiaotong University, Xi'an 710049 (China); Chen, Zhenmao, E-mail: firstname.lastname@example.com [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi'an Jiaotong University, Xi'an 710049 (China); Yang, Jinhong; Wang, Weihua [Institute of Applied Physics of AOA, Hefei 230031 (China)
In this paper, the dynamic response during plasma disruption of a test blanket module in vacuum vessel (VV) of the Joint TEXT (J-TEXT), which is an experimental Tokamak device with iron core, was simulated by applying a program developed by authors on the ANSYS platform using its parametric design language (APDL). The moving coordinate method as well as the load transfer and sequential coupling strategy were adopted to cope with the electromagneto-mechanical coupling effect. To establish the numerical model, the influence of the iron core on the eddy current and electromagnetic (EM) force during disruption was numerically investigated at first and the influence was found not significant. Together with the geometrical features of the J-TEXT Tokamak structure, 180° sector models without magnetic core were finally established for the EM field and the structural response simulations. To obtain the source plasma current, the plasma current evolution during disruption was simulated by using the Tokamak Simulation Code (TSC). With the numerical models and the source plasma current, the dynamic response of both the VV structure and the test module were calculated. The numerical results show that the maximum stress of the test module is in safe range, and the magnetic damping effect can weaken vibration of the test module. In addition, simulation without considering the coupling effect was carried out, which shows that the influence of coupling effect is not significant for the peak stress of the J-TEXT disruption problem.
Modeling of 3D magnetic equilibrium effects on edge turbulence stability during RMP ELM suppression in tokamaks
Wilcox, R. S.; Wingen, A.; Cianciosa, M. R.; Ferraro, N. M.; Hirshman, S. P.; Paz-Soldan, C.; Seal, S. K.; Shafer, M. W.; Unterberg, E. A.
Recent experimental observations have found turbulent fluctuation structures that are non-axisymmetric in a tokamak with applied 3D fields. In this paper, two fluid resistive effects are shown to produce changes relevant to turbulent transport in the modeled 3D magnetohydrodynamic (MHD) equilibrium of tokamak pedestals with these 3D fields applied. Ideal MHD models are insufficient to reproduce the relevant effects. By calculating the ideal 3D equilibrium using the VMEC code, the geometric shaping parameters that determine linear turbulence stability, including the normal curvature and local magnetic shear, are shown to be only weakly modified by applied 3D fields in the DIII-D tokamak. These ideal MHD effects are therefore not sufficient to explain the observed changes to fluctuations and transport. Using the M3D-C1 code to model the 3D equilibrium, density is shown to be redistributed on flux surfaces in the pedestal when resistive two fluid effects are included, while islands are screened by rotation in this region. The redistribution of density results in density and pressure gradient scale lengths that vary within pedestal flux surfaces between different helically localized flux tubes. This would produce different drive terms for trapped electron mode and kinetic ballooning mode turbulence, the latter of which is expected to be the limiting factor for pedestal pressure gradients in DIII-D.
Simultaneous evolution of plasma rotation, radial electric field, MHD activity and plasma confinement in the STOR-M tokamak
Dreval, M.; Xiao, C.; Trembach, D.; Hirose, A.; Elgriw, S.; Pant, A.; Rohraff, D.; Niu, T.
Radial electric field shear and poloidal plasma rotation are important factors affecting transport and confinement in tokamaks. Alteration of the electric field and plasma rotation in the vicinity of magnetic islands is also an important factor in tokamak plasma confinement. In the STOR-M tokamak, fast (~1 ms) simultaneous alterations of the radial electric field, plasma rotation (M|| = 0-0.4 in the plasma current direction), floating potential fluctuations in the periphery and MHD activity generated by rotating islands have been observed experimentally during normal ohmic discharges. The observed time and magnitude of the changes depend on the average electron density and poloidal beta at the beginning of the discharge. In discharges with high initial poloidal beta these changes are accompanied by a reduction in Hα emission and an increase in the line averaged density. Drastic decreases in Hα and increases in line averaged electron density and estimation of poloidal beta suggest that STOR-M confinement is significantly affected in ohmic discharges without an external additional energy input or biasing. MHD activity in STOR-M is damped when a negative electric field is observed at the limiter region of the plasma edge. MHD frequency is observed to decrease with the negative electric field.
Calculations of Energy Losses due to Atomic Processes in Tokamaks with Applications to the ITER Divertor
Post, D; Clark, R E H; Putvinskaya, N
Reduction of the peak heat loads on the plasma facing components is essential for the success of the next generation of high fusion power tokamaks such as the International Thermonuclear Experimental Reactor (ITER) 1 . Many present concepts for accomplishing this involve the use of atomic processes to transfer the heat from the plasma to the main chamber and divertor chamber walls and much of the experimental and theoretical physics research in the fusion program is directed toward this issue. The results of these experiments and calculations are the result of a complex interplay of many processes. In order to identify the key features of these experiments and calculations and the relative role of the primary atomic processes, simple quasi-analytic models and the latest atomic physics rate coefficients and cross sections have been used to assess the relative roles of central radiation losses through bremsstrahlung, impurity radiation losses from the plasma edge, charge exchange and hydrogen radiation losses f...
DNA contents in Texas bluegrass (Poa arachnifera) selected in Texas and Oklahoma determined by flow cytometry
Texas bluegrass (Poa arachnifera Torr.) is a dioecious, perennial, cool-season grass native to southern Kansas, Oklahoma, western Arkansas and most of Texas. Its major use has been for forage on rangelands in Texas and Oklahoma. More recently, interspecific hybrids between Texas bluegrass and Kentuc...
Maximum entropy reconstruction of poloidal magnetic field and radial electric field profiles in tokamaks
Chen, Yihang; Xiao, Chijie; Yang, Xiaoyi; Wang, Tianbo; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang
The Laser-driven Ion beam trace probe (LITP) is a new diagnostic method for measuring poloidal magnetic field (Bp) and radial electric field (Er) in tokamaks. LITP injects a laser-driven ion beam into the tokamak, and Bp and Er profiles can be reconstructed using tomography methods. A reconstruction code has been developed to validate the LITP theory, and both 2D reconstruction of Bp and simultaneous reconstruction of Bp and Er have been attained. To reconstruct from experimental data with noise, Maximum Entropy and Gaussian-Bayesian tomography methods were applied and improved according to the characteristics of the LITP problem. With these improved methods, a reconstruction error level below 15% has been attained with a data noise level of 10%. These methods will be further tested and applied in the following LITP experiments. Supported by the ITER-CHINA program 2015GB120001, CHINA MOST under 2012YQ030142 and National Natural Science Foundation Abstract of China under 11575014 and 11375053.
Robust Control of the Spatial Current Profile in the DIII-D Tokamak
Barton, J.; Schuster, E.; Walker, M. L.; Humphreys, D. A.
Advanced tokamak operating scenarios, characterized by large noninductively driven plasma currents, typically require active regulation of a specific current density profile. Non-model-based control of the q profile has been tested at DIII-D. However, some present limitations of the controller motivate the design of a model-based controller that accounts for the dynamics of the whole q profile in response to the control actuators. A control-oriented model of the current profile evolution in DIII-D was recently developed and used to design feedforward control schemes. In order to reject the effects of external disturbances to the system, a feedback control input needs to be added to the feedforward input. In this work, we report on the design of a robust feedback controller, on the implementation of the combined model-based feedforward + feedback controller in the DIII-D Plasma Control System, and on the experimental validation of the combined controller in the DIII-D tokamak. Supported by the NSF CAREER award program ECCS-0645086 and the US DOE under DE-FG02-09ER55064 and DE-FC02-04ER54698.
ICRF fast wave current drive and mode conversion current drive in EAST tokamak
Yin, L.; Yang, C.; Gong, X. Y.; Lu, X. Q.; Du, D.; Chen, Y.
Fast wave in the ion-cyclotron resonance frequency (ICRF) range is a promising candidate for non-inductive current drive (CD), which is essential for long pulse and high performance operation of tokamaks. A numerical study on the ICRF fast wave current drive (FWCD) and mode-conversion current drive (MCCD) in the Experimental Advanced Superconducting Tokamak (EAST) is carried out by means of the coupled full wave and Ehst-Karney parameterization methods. The results show that FWCD efficiency is notable in two frequency regimes, i.e., f ≥ 85 MHz and f = 50-65 MHz, where ion cyclotron absorption is effectively avoided, and the maximum on-axis driven current per unit power can reach 120 kA/MW. The sensitivity of the CD efficiency to the minority ion concentration is confirmed, owing to fast wave mode conversion, and the peak MCCD efficiency is reached for 22% minority-ion concentration. The effects of the wave-launch position and the toroidal wavenumber on the efficiency of current drive are also investigated.
Data Acquisition and Automation for Plasma Rotation Diagnostic in the TCABR Tokamak
Ronchi, G.; Severo, J. H. F.; de Sá, W. P.; Galvão, R. M. O.
In this work we describe the implementation of a full modular system of data acquisition and processing for the plasma rotation diagnostic in the TCABR tokamak. The experimental setup uses a single monochromator and six photomultipliers (PMT), in which pair of PMTs measures the light at slightly different wavelengths. Thus, it can measure the time evolution of the Doppler shift of the impurities emission lines coming from three spatial positions (one for toroidal rotation and two for poloidal rotation). The data acquisition and preanalysis program were written with LabVIEW software and is capable of controlling the spectrometer wavelength, PMTs power supplies, data acquisition, and storage. All data are recorded in MDSplus trees that easily allow data visualization and post-processing analysis (both locally and remotely) via MATLAB, Python, Java and others programming languages. This system can run independently from other diagnostics and machine systems and can be integrated with the main tokamak control system by means of TCP/IP messages.
Regional Haze Plan for Texas and Oklahoma
EPA partially approved and partially disapproved the Texas regional haze plan. EPA also finalized a plan to limit sulfur dioxide emissions from eight Texas coal-fired electricity generating facilities
Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo
Energy Technology Data Exchange (ETDEWEB)
Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)
Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)
Electrostatic turbulence intermittence driven by biasing in Texas Helimak
Energy Technology Data Exchange (ETDEWEB)
Toufen, D. L. [Federal Institute of Education, Science and Technology of São Paulo - IFSP, 07115-000 Guarulhos, São Paulo (Brazil); Institute of Physics, University of São Paulo, 05315-970 São Paulo, São Paulo (Brazil); Pereira, F. A. C.; Guimarães-Filho, Z. O.; Caldas, I. L. [Institute of Physics, University of São Paulo, 05315-970 São Paulo, São Paulo (Brazil); Gentle, K. W. [Department of Physics and Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)
We investigate changes in the intermittent sequence of bursts in the electrostatic turbulence due to imposed positive bias voltage applied to control the plasma radial electric field in Texas Helimak [K. W. Gentle and H. He, Plasma Sci. Technol. 10, 284 (2008)]—a toroidal plasma device with a one-dimensional equilibrium, magnetic curvature, and shear. We identify the burst characteristics by analyzing ion saturation current fluctuations collected in a large set of Langmuir probes. The number of bursts increase with positive biasing, giving rise to a long tailed skewed turbulence probability distribution function. The burst shape does not change much with the applied bias voltage, while their vertical velocity increases monotonically. For high values of bias voltage, the bursts propagate mainly in the vertical direction which is perpendicular to the radial density gradient and the toroidal magnetic field. Moreover, in contrast with the bursts in tokamaks, the burst velocity agrees with the phase velocity of the overall turbulence in both vertical and radial directions. For a fixed bias voltage, the time interval between bursts and their amplitudes follows exponential distributions. Altogether, these burst characteristics indicate that their production can be modelled by a stochastic process.
Texas Real Estate Curriculum Workshop Summary Report.
Lyon, Robert
The Texas Real Estate Research Center-Texas Education Agency (TRERC-TEA) curriculum workshop was attended by over 40 participants representing 26 Texas community colleges. These participants divided into eight small groups by real estate specialty area and developed curriculum outlines and learning objectives for the following real estate courses:…
Analysis of tokamak plasma confinement modes using the fast ...
Indian Academy of Sciences (India)
Oct 20, 2016 ... absence of the outer field, and then compared with each other. The number of plasma modes and the safety factor q were determined using the FFT method in the presence and absence of the outer field. The safety factor q plays a significant role in determin- ing the stability of tokamak plasma and seems to.
Electron cyclotron emission imaging diagnostic system for Rijnhuizen Tokamak Project
Deng, B.H.; Hsia, R. P.; Domier, C.W.; Burns, S. R.; Hillyer, T. R.; N C Luhmann Jr.,; Oyevaar, T.; Donne, A. J. H.; R. T. P. Team,
A 16-channel electron cyclotron emission (ECE) imaging diagnostic system has been developed and installed on the Rijnhuizen Tokamak Project for measuring plasma electron cyclotron emission with a temporal resolution of 2 mu s. The high spatial resolution of the system is achieved by utilizing a low
Bulk Ion Heating with ICRF Waves in Tokamaks
DEFF Research Database (Denmark)
Mantsinen, M. J.; Bilato, R.; Bobkov, V. V.
Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER a...
Conceptual design of Remote Control System for EAST tokamak
Energy Technology Data Exchange (ETDEWEB)
Sun, X.Y., E-mail: email@example.com; Wang, F.; Wang, Y.; Li, S.
Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.
Internal magnetic field measurement in tokamak plasmas using a ...
Indian Academy of Sciences (India)
There is a growing interest in developing a reliable method for the measurement of the in- ternal magnetic field in high ... This information is essential for understanding confinement, stability and energy balance of the tokamak plasma. .... The instrument measures the difference between the left-hand and right-hand circularly ...
Evidence of Inward Toroidal Momentum Convection in the JET Tokamak
DEFF Research Database (Denmark)
Tala, T.; Zastrow, K.-D.; Ferreira, J.
Experiments have been carried out on the Joint European Torus tokamak to determine the diffusive and convective momentum transport. Torque, injected by neutral beams, was modulated to create a periodic perturbation in the toroidal rotation velocity. Novel transport analysis shows the magnitude...
Sensitivity of transient synchrotron radiation to tokamak plasma parameters
Energy Technology Data Exchange (ETDEWEB)
Fisch, N.J.; Kritz, A.H.
Synchrotron radiation from a hot plasma can inform on certain plasma parameters. The dependence on plasma parameters is particularly sensitive for the transient radiation response to a brief, deliberate, perturbation of hot plasma electrons. We investigate how such a radiation response can be used to diagnose a variety of plasma parameters in a tokamak. 18 refs., 13 figs.
Loop-voltage tomography in tokamaks using transient synchrotron radiation
Energy Technology Data Exchange (ETDEWEB)
Fisch, N.J.; Kritz, A.H. (Princeton Univ., NJ (USA). Plasma Physics Lab.; Hunter Coll., New York, NY (USA). Dept. of Physics)
The loop voltage in tokamaks is particularly difficult to measure anywhere but at the plasma periphery. A brief, deliberate, perturbation of hot plasma electrons, however, produces a transient radiation response that is sensitive to this voltage. We investigate how such a radiation response can be used to diagnose the loop voltage. 24 refs., 6 figs.
Recording non-local temperature rise in the RTP tokamak
Hogeweij, G. M. D.; Mantica, P.; Gorini, G.; de Kloe, J.; Cardozo, N. J. L.; R. T. P. Team,
In the Rijnhuizen Tokamak Project (RTP) plasmas with electron cyclotron heating (ECH), a transient rise of the core T-e is observed when hydrogen pellets are injected tangentially to induce fast cooling of the peripheral region. The core T-e rise is a sharp function of the normalized power
Stability of localized modes in rotating tokamak plasmas
J.W. Haverkort (Willem); H.J. de Blank
textabstractThe ideal magnetohydrodynamic stability is investigated of localized interchange modes in a large-aspect ratio tokamak plasma. The resulting stability criterion includes the effects of toroidal rotation and rotation shear and contains various well-known limiting cases. The analysis
Structural effects of plasma instabilities on the JET tokamak
Energy Technology Data Exchange (ETDEWEB)
Buzio, M
The subject of this work is a novel approach to the analysis of the mechanical response of the main structural components of the JET Tokamak to the JxB forces generated by MHD plasma instabilities. The proposed method is based on the creation of simplified lumped-parameter models representing the essential mechanical and electromagnetic characteristics of the interacting components and consists basically in fitting their output to experimental measurements in order to infer information on induced currents and related forces by means of statistical parameter estimation techniques. First of all, the general time history and space distribution of disruption loads are described and the path of the reaction forces throughout the machine is analysed in detail, with particular reference to the recently observed non-axisymmetric cases. For this purpose, a magneto-static model of the interaction between a kinked plasma and the coil system has been developed. This is based on semi-analytical integration of Biot-Savart's law and makes use of a representation of the conductors in terms of Bezier curves. The method implemented, which pen-nits efficient and detailed calculation of magnetic load distributions in complex geometries, is used to analyse non-axisymmetric events and to point out the most critical components under this kind of loading. The attention is focused next on lumped-parameter models which have been created to represent the basic response modes of the Vessel. These models are represented by a combination of concentrated masses, springs and dampers and include additional parameters describing the magnetic loads. The output of these models is fitted to experimental measurements of displacements and support forces in order to obtain estimates for the magnitude, position and timing of induced currents and related forces. A general procedure for model-based maximum likelihood parameters estimation has been implemented as an interactive PC-Windows program in
Residential Energy Efficiency Potential: Texas
Energy Technology Data Exchange (ETDEWEB)
Wilson, Eric J [National Renewable Energy Laboratory (NREL), Golden, CO (United States)
Energy used by Texas single-family homes that can be saved through cost-effective improvements. Prepared by Eric Wilson and Noel Merket, NREL, and Erin Boyd, U.S. Department of Energy Office of Energy Policy and Systems Analysis.
South Texas Maquiladora Suppliers Project.
Patrick, J. Michael
This project was undertaken to assist South Texas industries in improving export to nearby Mexican maquiladoras (factories). The maquiladora program is based on co-production by two plants under a single management, one on each side of the border. Activities addressed four objectives: (1) to determine the dollar value, quantity, and source of the…
Red tides (algae) bloomed late this summer along a 300-mile stretch of Texas' Gulf Coast, killing millions of fish and shellfish as well as making some people sick. State officials are calling this the worst red tide bloom in 14 years. The algae produces a poison that paralyzes fish and prevents them from breathing. There is concern that the deadly algae could impact or even wipe out this year's oyster harvest in Texas, which usually peaks during the Thanksgiving and Christmas holidays. The red tides were first observed off the Texas coast in mid-August and have been growing steadily in size ever since. Red tides tend to bloom and subside rapidly, depending upon changes in wind speed and direction, water temperature, salinity, and rainfall patterns (as the algae doesn't do as well in fresher water). This true-color image of the Texas Gulf Coast was acquired on September 29, 2000, by the Moderate-resolution Imaging Spectroradiometer (MODIS) flying aboard NASA's Terra spacecraft. The red tide can be seen as the dark reddish discoloration in the ocean running southwest to northeast along the coast. In this scene, the bloom appears to be concentrated north and east of Corpus Christi, just off Matagorda Island. The image was made at 500-meter resolution using a combination of MODIS' visible bands 1 (red), 4 (green), and 3 (blue). The city of Houston can be seen clearly as the large, greyish cluster of pixels to the north and west of Galveston Bay, which is about mid-way up the coastline in this image. Also visible in this image are plumes of smoke, perhaps wildfires, both to the north and northeast of Houston. For more information about red tides, refer to the Texas Red Tide Web site. Image courtesy Andrey Savtchenko, MODIS Data Support Team, and the MODIS Ocean Team, NASA's Goddard Space Flight Center
Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak
Energy Technology Data Exchange (ETDEWEB)
Maurya, Gulab Singh; Kumar, Rohit; Rai, Awadhesh Kumar, E-mail: email@example.com [Laser Spectroscopy Research Laboratory, Department of Physics, University of Allahabad, UP 211002 (India); Kumar, Ajai [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat 382428 (India)
In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known as "back collection method" to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.
Resistive reduced MHD modeling of multi-edge-localized-mode cycles in Tokamak X-point plasmas.
Orain, F; Bécoulet, M; Huijsmans, G T A; Dif-Pradalier, G; Hoelzl, M; Morales, J; Garbet, X; Nardon, E; Pamela, S; Passeron, C; Latu, G; Fil, A; Cahyna, P
The full dynamics of a multi-edge-localized-mode (ELM) cycle is modeled for the first time in realistic tokamak X-point geometry with the nonlinear reduced MHD code jorek. The diamagnetic rotation is found to be instrumental to stabilize the plasma after an ELM crash and to model the cyclic reconstruction and collapse of the plasma pressure profile. ELM relaxations are cyclically initiated each time the pedestal gradient crosses a triggering threshold. Diamagnetic drifts are also found to yield a near-symmetric ELM power deposition on the inner and outer divertor target plates, consistent with experimental measurements.
Thermal instability theory analysis of multifaceted asymmetric radiation from the edge (MARFE) in Tokamak Experiment for Technology Oriented Research (TEXTOR)
Energy Technology Data Exchange (ETDEWEB)
Kelly, F. A.; Stacey, W. M.; Rapp, J.; Brix, M.
The density limits for a series of shots in TEXTOR [Tokamak Experiment for Technology Oriented Research, E. Hintz, P. Bogen, H. A. Claa{ss}en , in Contributions to High-Temperature Plasma Physics, edited by K. H. Spatschek and J. Uhlenbusch (Akademie Verlag, Berlin, 1994, p. 373)], over a range of heating powers, that ended in multifaceted asymmetric radiation from the edge (MARFE) have been analyzed within the context of thermal instability theory. The prediction of MARFE onset agrees with observation to within the experimental uncertainty.
Effects of Density and Impurity on Edge Localized Modes in Tokamaks
Zhu, Ping
Plasma density and impurity concentration are believed to be two of the key elements governing the edge tokamak plasma conditions. Optimal levels of plasma density and impurity concentration in the edge region have been searched for in order to achieve the desired fusion gain and divertor heat/particle load mitigation. However, how plasma density or impurity would affect the edge pedestal stability may have not been well known. Our recent MHD theory modeling and simulations using the NIMROD code have found novel effects of density and impurity on the dynamics of edge-localized modes (ELMs) in tokamaks. First, previous MHD analyses often predict merely a weak stabilizing effect of toroidal flow on ELMs in experimentally relevant regimes. We find that the stabilizing effects on the high- n ELMs from toroidal flow can be significantly enhanced with the increased edge plasma density. Here n denotes the toroidal mode number. Second, the stabilizing effects of the enhanced edge resistivity due to lithium-conditioning on the low- n ELMs in the high confinement (H-mode) discharges in NSTX have been identified. Linear stability analysis of the experimentally constrained equilibrium suggests that the change in the equilibrium plasma density and pressure profiles alone due to lithium-conditioning may not be sufficient for a complete suppression of the low- n ELMs. The enhanced resistivity due to the increased effective electric charge number Zeff after lithium-conditioning provides additional stabilization of the low- n ELMs. These new effects revealed in our theory analyses may help further understand recent ELM experiments and suggest new control schemes for ELM suppression and mitigation in future experiments. They may also pose additional constraints on the optimal levels of plasma density and impurity concentration in the edge region for H-mode tokamak operation. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB
Improvement of plasma energy confinement in tokamak under radiative cooling of the edge plasma
Razumova, K. A.; Borschegovskiy, A. A.; Gorbunov, E. P.; Dremin, M. M.; Kasyanova, N. V.; Kirneva, N. A.; Kislov, A. Ya.; Klyuchnikov, L. A.; Krupin, V. A.; Krylov, S. V.; Lysenko, S. E.; Melnikov, A. V.; Myalton, T. B.; Nemets, A. R.; Notkin, G. E.; Nurgaliev, M. R.; Sarychev, D. V.; Sushkov, A. V.; Chistyakov, V. V.; Ongena, J.; Messiaen, A.
Improvement of plasma energy confinement in the T-10 tokamak by injection of impurity gases was studied experimentally. Injection of Ne and He in the ohmic and ECR heating regimes allows one to separate the dependences of energy confinement on the plasma density and on the edge plasma cooling rate. It is shown that the well-known dependence of the energy confinement time on the plasma density is, in fact, the dependence on the radiative loss power. This phenomenon can be explained by plasma self-organization. The experiments are described by a thermodynamic model for self-organized plasma in which the transport coefficient depends on the difference between the actual and self-consistent pressure profiles. The reduction in the heat flux at the plasma edge due to radiative cooling leads to a decrease in the transport coefficient in this region and, accordingly, improves energy confinement. Results of approximate model calculations for experiments with Ne injection are presented.
Exposure of CFC-materials to high transient heat loads in the TEXTOR tokamak
Scholz, T.; Boedo, J.; Bolt, H.; Duwe, R.; Finken, K. H.; Gray, D.; Hassanein, A.
Transient high heat flux events like ELMs, vertical displacement events and disruptions can cause the thermal ablation of plasma facing material. Until now experimental work in this field had been carried out by exposing material specimens to heat loads by electron or laser beam or by tests in pulsed plasma accelerators. In the present work carbon specimens were directly exposed to intense plasma fluxes in the TEXTOR tokamak. The exposure was performed with a fast probe allowing the insertion of the material over a distance of 9 cm into the edge plasma for a duration of 80 ms. The results of in-situ diagnostic measurements and of the post-experiment examination of the specimens are compared with a reference experiment by electron beam and with numerical analyses. Results indicated that the heat flux to the probe surfaces and the probe erosion is much lower than expected.
A new soft x-ray pulse height analysis array in the HL-2A tokamak.
Zhang, Y P; Liu, Yi; Yang, J W; Song, X Y; Liao, M; Li, X; Yuan, G L; Yang, Q W; Duan, X R; Pan, C H
A new soft x-ray pulse height analysis (PHA) array including nine independent subsystems, on basis of a nonconventional software multichannel analysis system and a silicon drift detector (SDD) linear array consisting of nine high performance SDD detectors, has been developed in the HL-2A tokamak. The use of SDD has greatly improved the measurement accuracy and the spatiotemporal resolutions of the soft x-ray PHA system. Since the ratio of peak to background counts obtained from the SDD PHA system is very high, p/b > or = 3000, the soft x-ray spectra measured by the SDD PHA system can approximatively be regarded as electron velocity distribution. The electron velocity distribution can be well derived in the pure ohmic and auxiliary heating discharges. The performance of the new soft x-ray PHA array and the first experimental results with some discussions are presented.
L-H power threshold studies with tungsten/carbon divertor on the EAST tokamak
DEFF Research Database (Denmark)
Chen, L.; Xu, G. S.; Gao, W.
The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full...... carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L-H power threshold. When operating in the upper single null...... configuration, with the ion grad-B drift direction away from the primary X-point, a lower normalized power threshold is observed in EAST with the tungsten/carbon divertor, compared to the carbon divertor after intensive lithium wall coating. A newly installed cryopump increasing the pumping efficiency also...
Statistical Properties of Turbulent Transport During L-H Transition in the STOR-M Tokamak
White, D.; Ding, W.; Xiao, C.; Hirose, A.
An algorithm to determine the turbulence correlation direction, as well as correlation strength between spatially (radially) separated positions, has been developed and applied to edge density, potential, and temperature fluctuations in the STOR-M tokamak before and after the L-H transition induced by electrode biasing. The radial correlation length revealed from the nonlinear correlation is significantly longer than that estimated from the conventional correlation technique. The ``master-slave" correlation direction is reversed after L-H transition from radially outward during L-mode to radially inward during H-mode. The thermal diffusivity estimated from the nonlinear correlation length and ``entropy" exhibits a better agreement with experimental value than the conventional estimate.
Temperature measurement of plasma-facing surfaces in tokamaks by active pyrometry
Energy Technology Data Exchange (ETDEWEB)
Grigorova, V.; Semerok, A.; Farcage, D.; Weulersse, J.M. [CEA Saclay, DEN/DPC/SCP/LILM, Bat. 467, 91191 Gif-sur-Yvette (France); Thro, P.Y., E-mail: email@example.com [CEA Saclay, DEN/DPC/SCP/LILM, Bat. 467, 91191 Gif-sur-Yvette (France); Gauthier, E.; Roche, H.; Loarer, Th.; Grisolia, Ch. [CEA Cadarache, DSM/ IRFM/SIPP, 13108 Saint Paul Lez Durance (France)
This paper discusses feasibility and tests of a new method for in situ temperature measurement of tokamak plasma-facing metallic surfaces under plasma presence. In such conditions, the other temperature-measurement methods are not applicable due to the perturbing thermal radiation reflected by the walls. Our approach overcomes this limitation by looking with two pyrometers to the measured surface while thermally perturbed. Because of the thermal perturbation each pyrometer records a signal modulation. The temperature, deduced by the ratio between the two signal modulations is dependent neither on the environmental reflecting fluxes nor on the surface emissivity. Originally, the measured temperature is linked to the signals ratio via the experimental set-up parameters. Here, we proposed an alternative way to deduce it from the pyrometers calibration data only. With this method we obtained temperature measurements with accuracy better than 90%.
Two-dimensional vacuum ultraviolet images in different MHD events on the EAST tokamak
Zhijun, WANG; Xiang, GAO; Tingfeng, MING; Yumin, WANG; Fan, ZHOU; Feifei, LONG; Qing, ZHUANG; EAST Team
A high-speed vacuum ultraviolet (VUV) imaging telescope system has been developed to measure the edge plasma emission (including the pedestal region) in the Experimental Advanced Superconducting Tokamak (EAST). The key optics of the high-speed VUV imaging system consists of three parts: an inverse Schwarzschild-type telescope, a micro-channel plate (MCP) and a visible imaging high-speed camera. The VUV imaging system has been operated routinely in the 2016 EAST experiment campaign. The dynamics of the two-dimensional (2D) images of magnetohydrodynamic (MHD) instabilities, such as edge localized modes (ELMs), tearing-like modes and disruptions, have been observed using this system. The related VUV images are presented in this paper, and it indicates the VUV imaging system is a potential tool which can be applied successfully in various plasma conditions.
Observation of ion-cyclotron-frequency mode-conversion flow drive in tokamak plasmas.
Lin, Y; Rice, J E; Wukitch, S J; Greenwald, M J; Hubbard, A E; Ince-Cushman, A; Lin, L; Porkolab, M; Reinke, M L; Tsujii, N
Strong toroidal flow (Vphi) and poloidal flow (Vtheta) have been observed in D-3He plasmas with ion cyclotron range of frequencies (ICRF) mode-conversion (MC) heating on the Alcator C-Mod tokamak. The toroidal flow scales with the rf power Prf (up to 30 km/s per MW), and is significantly larger than that in ICRF minority heated plasmas at the same rf power or stored energy. The central Vphi responds to Prf faster than the outer regions, and the Vphi(r) profile is broadly peaked for r/a or =1.5 MW and increases with power (up to 0.7 km/s per MW). The experimental evidence together with numerical wave modeling suggests a local flow drive source due to the interaction between the MC ion cyclotron wave and 3He ions.
3-D Monte Carlo analyses of the shielding system in a tokamak fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Gallina, M.; Petrizzi, L.; Rado, V. (ENEA, Frascati (Italy). Centro Ricerche Energia)
As part of the ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations have been carried out to assess the shielding system performance in the basic machine configuration by means of the Monte Carlo Neutron Photon (MCNP) transport code (3-B version). The main issue is the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. ''Self generated weight windows'' and source biasing technique have been used to treat deep penetration through the bulk shield and streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted. (author).
3-D Monte Carlo analyses of shielding system in tokamak fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Gallina, M.; Petrizzi, L.; Rado, V.
Within the framework of the ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations were carried out to assess system shielding performances in the basic machine configuration by means of the Monte Carlo Neutron Photon (MCNP) code (3-B version). The main issue concerns the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. 'Self generated weight windows' (w.w.) and source biasing techniques were used to treat the deep penetration through the bulk shield and streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted.
Penetration of filamentary structures in the x-point region of spherical tokamaks
Baver, D. A.; Myra, J. R.; Scotti, F.; Zweben, S. J.; Militello, F.; Walkden, N.
ArbiTER is a flexible eigenvalue code designed for plasma physics applications. It is used here to gain insight into the spatial dependence of filamentary structures in the scrape-off layer of spherical tokamaks. In particular, observations on MAST reveal the presence of a quiescent x-point region. Observations in NSTX similarly reveal a reduction in divertor fluctuations near the separatrix and a loss of midplane correlation. We will report on the penetration of filamentary structures into the vicinity of the x-point, as well as growth rate trends, for a variety of profiles and toroidal mode numbers. This will determine whether linear properties of these structures can explain experimental observations. Work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-02ER54678.
A snowflake divertor: a possible solution to the power exhaust problem for tokamaks
Energy Technology Data Exchange (ETDEWEB)
Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Umansky, M. V.
This paper summarizes recent progress in the theory of a snowflake divertor, a possible path to reduce both steady-state and intermittent heat loads on the divertor plates to an acceptable level. The most important feature of a SF divertor is the presence of a large zone of a very weak poloidal magnetic field around the poloidal field (PF) null. Qualitative explanation of a variety of new features characteristic of a SF divertor is provided based on simple scaling relations. The main part of the paper is focused on the concept of spreading of the heat flux by curvature-driven convection near the PF null. References to experimental results from the NSTX and TCV tokamaks are provided.
Simulation of W dust transport in the KSTAR tokamak, comparison with fast camera data
Directory of Open Access Journals (Sweden)
A. Autricque
Full Text Available In this paper, dust transport in tokamak plasmas is studied through both experimental and modeling aspects. Image processing routines allowing dust tracking on CCD camera videos are presented. The DUMPRO (DUst Movie PROcessing code features a dust detection method and a trajectory reconstruction algorithm. In addition, a dust transport code named DUMBO (DUst Migration in a plasma BOundary is briefly described. It has been developed at CEA in order to simulate dust grains transport in tokamaks and to evaluate the contribution of dust to the impurity inventory of the plasma. Like other dust transport codes, DUMBO integrates the Orbital Motion Limited (OML approach for dust/plasma interactions modeling. OML gives direct expressions for plasma ions and electrons currents, forces and heat fluxes on a dust grain. The equation of motion is solved, giving access to the dust trajectory. An attempt of model validation is made through comparison of simulated and measured trajectories on the 2015 KSTAR dust injection experiment, where W dust grains were successfully injected in the plasma using a gun-type injector. The trajectories of the injected particles, estimated using the DUMPRO routines applied on videos from the fast CCD camera in KSTAR, show two distinct general dust behaviors, due to different dust sizes. Simulations were made with DUMBO to match the measurements. Plasma parameters were estimated using different diagnostics during the dust injection experiment plasma discharge. The experimental trajectories show longer lifetimes than the simulated ones. This can be due to the substitution of a boiling/sublimation point to the usual vaporization/sublimation cooling, OML limitations (eventual potential barriers in the vicinity of a dust grain are neglected and/or to the lack of a vapor shielding model in DUMBO.
Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD
Energy Technology Data Exchange (ETDEWEB)
Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)
A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.
Stability analysis of tokamak plasmas; Analyse de stabilite de plasmas de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bourdelle, C
In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)
Statistical study of density fluctuations in the tore supra tokamak
Energy Technology Data Exchange (ETDEWEB)
Devynck, P.; Fenzi, C.; Garbet, X.; Laviron, C. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antar, G.; Gervais, F.; Hennequin, P.; Quemeneur, A.; Sabot, R.; Truc, A. [LPMI, CNRS UPR-287, Ecole Polytechnique, 91 - Palaiseau (France)
It is believed that the radial anomalous transport in tokamaks is caused by plasma turbulence. Using infra-red laser scattering technique on the Tore Supra tokamak, statistical properties of the density fluctuations are studied as a function of the scales in ohmic as well as additional heating regimes using the lower hybrid or the ion cyclotron frequencies. The probability distributions are compared to a Gaussian in order to estimate the role of intermittency which is found to be negligible. The temporal behaviour of the three-dimensional spectrum is thoroughly discussed; its multifractal character is reflected in the singularity spectrum. The autocorrelation coefficient as well as their long-time incoherence and statistical independence. We also put forward the existence of fluctuations transfer between two distinct but close wavenumbers. A rather clearer image is thus obtained about the way energy is transferred through the turbulent scales. (author) 28 refs.
Abel inversion of asymmetric plasma density profile at Aditya tokamak
Joshi, N. Y.; Atrey, P. K.; Pathak, S. K.
In Aditya tokamak, at Institute for Plasma Research, till now, multi-channel microwave interferometer system is used to measure the cord averaged plasma density at predefined radial position. An inversion code is developed to determine the local density profile from the chord average density measurement of radially asymmetric plasma. The radial density profile is interpolated using Spline interpolation analytical technique for symmetric plasma density profile. Code implements the Slice and Stack method to determine localized density from asymmetric averaged plasma density measurement from interferometer. Inverted results are tested with various monotonically varying asymmetric radial density profiles of the plasma shots. It also provides the poloidal picture of plasma density distribution with circular constant density surfaces. Localized density measurements, which is very important for successful operation of tokamak, is in agreement with observation of other diagnostics.
The simple map for a single-null divertor tokamak
Energy Technology Data Exchange (ETDEWEB)
Punjabi, A.; Verma, A.; Boozer, A. [Hampton Univ. (Vatican City State, Holy See). Center for Fusion Research and Training
We present the simple map for a single-null divertor tokamak. The simple map is an area-preserving map based on the idea that magnetic field lines are a single-degree-of-freedom time-dependent Hamiltonian system, and that the basic features of such systems near the X-point are generic. We obtain the properties of this map and the resulting footprints of field lines on the divertor plate. These include the width of the stochastic layer, the edge safety factor, the area of the footprint and the amount of magnetic flux diverted. We give the safety factor profile, the average and median values of strike angles, lengths and the Liapunov exponents. We describe how the effects of magnetic perturbations can be included in the simple map. We show how the map can be applied to the problem of the determination of heat flux on the divertor plate in tokamaks. (Author).
Phase Contrast Imaging on the HL-2A Tokamak
Yu, Yi; Gong, Shaobo; Xu, Min; Jiang, Wei; Zhong, Wulv; Shi, Zhongbin; Wang, Huajie; Wu, Yifan; Yuan, Boda; Lan, Tao; Ye, Minyou; Duan, Xuru; HL-2A Team
In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wavenumber of 66 cm-1. The designed wavenumber resolution is 2.09cm-1, and the time resolution is higher than 0.2 μs. The broad kρs ranging from 0.34 to 13.37 makes it suitable for turbulence measurement. An upgraded PCI system is also discussed, which is designed for the HL-2M tokamak. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No. 2015GB120002), the National Natural Science Foundation of China (Grant No. 11375053, 11105144, 10905057, 11535013).
Remote operation of the GOLEM tokamak for Fusion Education
Energy Technology Data Exchange (ETDEWEB)
Grover, O.; Kocman, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Odstrcil, M. [University of Southampton, Southampton SO17 1BJ (United Kingdom); Odstrcil, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Matusu, M. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, Prague CZ-182 21 (Czech Republic); Svoboda, V., E-mail: firstname.lastname@example.com [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Vondrasek, G. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Zara, J. [Faculty of Electrical Engineering CTU Prague, CZ-166 27 (Czech Republic)
Highlights: • The remote operation of the tokamak GOLEM for educational purposes. - Abstract: Practically oriented education in the field of thermonuclear fusion is highly requested. However, the high complexity of appropriate experiments makes it difficult to develop and maintain laboratories where students can take part in hands-on experiments in this field of study. One possible solution is to establish centres with specific high temperature plasma experiments where students can visit such a laboratory and perform their experiments in-situ. With the advancements of IT technologies it naturally follows to make a step forward and connect these with necessary plasma physics technologies and thus allow to access even sophisticated experiments remotely. Tokamak GOLEM is a small, modest device with its infrastructure linked to web technologies allowing students to set-up necessary discharge parameters, submit them into a queue and within minutes obtain the results in the form of a discharge homepage.
Gyrokinetic Simulation of Global Turbulent Transport Properties in Tokamak Experiments
Energy Technology Data Exchange (ETDEWEB)
Wang, W.X.; Lin, Z.; Tang, W.M.; Lee, W.W.; Ethier, S.; Lewandowski, J.L.V.; Rewoldt, G.; Hahm, T.S.; Manickam, J.
A general geometry gyro-kinetic model for particle simulation of plasma turbulence in tokamak experiments is described. It incorporates the comprehensive influence of noncircular cross section, realistic plasma profiles, plasma rotation, neoclassical (equilibrium) electric fields, and Coulomb collisions. An interesting result of global turbulence development in a shaped tokamak plasma is presented with regard to nonlinear turbulence spreading into the linearly stable region. The mutual interaction between turbulence and zonal flows in collisionless plasmas is studied with a focus on identifying possible nonlinear saturation mechanisms for zonal flows. A bursting temporal behavior with a period longer than the geodesic acoustic oscillation period is observed even in a collisionless system. Our simulation results suggest that the zonal flows can drive turbulence. However, this process is too weak to be an effective zonal flow saturation mechanism.
Ageing of structural materials in tokamaks: TEXTOR liner study
Weckmann, A.; Petersson, P.; Rubel, M.; Fortuna-Zaleśna, E.; Zielinski, W.; Romelczyk-Baishya, B.; Grigore, E.; Ruset, C.; Kreter, A.
After the final shut-down of the tokamak TEXTOR, all of its machine parts became accessible for comprehensive studies. This unique opportunity enabled the study of the Inconel 625 liner by a wide range of methods. The aim was to evaluate eventual alteration of surface and bulk characteristics from recessed wall elements that may influence the machine performance. The surface was covered with stratified layers consisting mainly of boron, carbon, oxygen, and in some cases also silicon. Wall conditioning and limiter materials hence predominantly define deposition on the liner. Deposited layers on recessed wall elements reach micrometre thickness within decades, peel off and may contribute to the dust inventory in tokamaks. Deuterium content was about 4,7 at% on average most probably due to wall conditioning with deuterated gas, and very low concentration in the Inconel substrate. Inconel 625 retained its mechanical strength despite 26 years of cyclic heating, stresses and particle bombardment.
Analytical solutions for Tokamak equilibria with reversed toroidal current
Energy Technology Data Exchange (ETDEWEB)
Martins, Caroline G. L.; Roberto, M.; Braga, F. L. [Departamento de Fisica, Instituto Tecnologico de Aeronautica, Sao Jose dos Campos, Sao Paulo 12228-900 (Brazil); Caldas, I. L. [Instituto de Fisica, Universidade de Sao Paulo, 05315-970 Sao Paulo, SP (Brazil)
In tokamaks, an advanced plasma confinement regime has been investigated with a central hollow electric current with negative density which gives rise to non-nested magnetic surfaces. We present analytical solutions for the magnetohydrodynamic equilibria of this regime in terms of non-orthogonal toroidal polar coordinates. These solutions are obtained for large aspect ratio tokamaks and they are valid for any kind of reversed hollow current density profiles. The zero order solution of the poloidal magnetic flux function describes nested toroidal magnetic surfaces with a magnetic axis displaced due to the toroidal geometry. The first order correction introduces a poloidal field asymmetry and, consequently, magnetic islands arise around the zero order surface with null poloidal magnetic flux gradient. An analytic expression for the magnetic island width is deduced in terms of the equilibrium parameters. We give examples of the equilibrium plasma profiles and islands obtained for a class of current density profile.
Progress in application of high temperature superconductor in tokamak magnets
Czech Academy of Sciences Publication Activity Database
Gryaznevich, M.; Svoboda, V.; Stöckel, Jan; Sykes, A.; Sykes, N.; Kingham, D.; Hammond, G.; Apte, P.; Todd, T.N.; Ball, S.; Chappell, S.; Melhem, D.; Ďuran, Ivan; Kovařík, Karel; Grover, O.; Markovič, T.; Odstrčil, M.; Odstrčil, T.; Šindlery, A.; Vondrášek, G.; Kocman, J.; Lilley, M.K.; de Grouchy, P.; Kim, H.-T.
Roč. 88, 9-10 (2013), s. 1593-1596 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] Institutional support: RVO:61389021 Keywords : tokamaks * HTS * magnet s Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001117#
Advanced Tokamak Plasmas in the Fusion Ignition Research Experiment
Energy Technology Data Exchange (ETDEWEB)
C.E. Kessel; D. Meade; D.W. Swain; P. Titus; M.A. Ulrickson
The Advanced Tokamak (AT) capability of the Fusion Ignition Research Experiment (FIRE) burning plasma experiment is examined with 0-D systems analysis, equilibrium and ideal-MHD stability, radio-frequency current-drive analysis, and full discharge dynamic simulations. These analyses have identified the required parameters for attractive burning AT plasmas, and indicate that these are feasible within the engineering constraints of the device.
Structural materials for large superconducting magnets for tokamaks
Energy Technology Data Exchange (ETDEWEB)
Long, C.J.
The selection of structural materials for large superconducting magnets for tokamak-type fusion reactors is considered. The important criteria are working stress, radiation resistance, electromagnetic interaction, and general feasibility. The most advantageous materials appear to be face-centered-cubic alloys in the Fe-Ni-Cr system, but high-modulus composites may be necessary where severe pulsed magnetic fields are present. Special-purpose structural materials are considered briefly.
A moving finite element model of the tokamak scrapeoff layer
Energy Technology Data Exchange (ETDEWEB)
Glasser, A.H.; Kuprat, A.P.
Most numerical simulations of the tokamak scrapeoff layer use a mapping to flux coordinates and a piecewise equidistributed grid in those coordinates to resolve the multiple length scales and anisotropy characteristic of this problem. We have developed an alternative numerical method using simple cylindrical coordinates with a complex adaptive grid scheme. It is based on an understructured grid of traingles which move adaptively, aligning themselves with the magnetic field and concentrating in regions of sharp gradients.
Application of advanced composites in tokamak magnet systems
Energy Technology Data Exchange (ETDEWEB)
Long, C. J.
The use of advanced (high-modulus) composites in superconducting magnets for tokamak fusion reactors is discussed. The most prominent potential application is as the structure in the pulsed poloidal-field coil system, where a significant reduction in eddy currents could be achieved. Present low-temperature data on the advanced composites are reviewed briefly; they are too meager to do more than suggest a broad class of composites for a particular application.
Implementation of rapid imaging system on the COMPASS tokamak.
Czech Academy of Sciences Publication Activity Database
Havránek, Aleš; Weinzettl, Vladimír; Fridrich, David; Cavalier, Jordan; Urban, Jakub; Komm, Michael
Roč. 123, November (2017), s. 857-860 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Camera * Data acquisition * Video processing * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S092037961730354X
Remote operation of the GOLEM tokamak for Fusion Education
Czech Academy of Sciences Publication Activity Database
Grover, O.; Kocman, J.; Odstrčil, M.; Odstrčil, T.; Matušů, M.; Stöckel, Jan; Svoboda, V.; Vondrášek, G.; Žára, J.
Roč. 112, November (2016), s. 1038-1044 ISSN 0920-3796. [Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research IAEA /10./. Ahmedabad, 20.04.2015-24.04.2015] Institutional support: RVO:61389021 Keywords : Tokamak technology * Remote participation * Education * Nuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379616303441
Automation of Aditya tokamak plasma position control DC power supply
Energy Technology Data Exchange (ETDEWEB)
Arambhadiya, Bharat, E-mail: email@example.com; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.
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.
Tokamak startup using point-source dc helicity injection.
Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C
Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.
Viscous damping of toroidal angular momentum in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Stacey, W. M. [Georgia Tech Fusion Research Center, Atlanta, Georgia 30332 (United States)
The Braginskii viscous stress tensor formalism was generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry in order to provide a representation for the viscous damping of toroidal rotation in tokamaks arising from various "neoclassical toroidal viscosity" mechanisms. In the process, it was verified that the parallel viscosity contribution to damping toroidal angular momentum still vanishes even in the presence of toroidal asymmetries, unless there are 3D radial magnetic fields.
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
Energy Technology Data Exchange (ETDEWEB)
Castracane, J.
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.
Multi-field plasma sandpile model in tokamaks and applications
Peng, X. D.; Xu, J. Q.
A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.
Quantification of chemical erosion in the divertor of the DIII-D tokamak
McLean, Adam Gordon
The International Thermonuclear Experimental Reactor (ITER) is currently designed to use graphite targets in the divertor for power handling and impurity control. Understanding and quantifying chemical sputtering is therefore key to the success of fusion as a clean energy source. The principal goal of this thesis is to design and carry out experiments, then analyze and interpret the results in order to elucidate the role of chemical sputtering in carbon sources in the DIII-D tokamak. A self-contained gas puff system has been designed, constructed, and employed for in-situ study of chemical erosion. The porous plug injector (PPI) releases methane through a porous graphite surface into the divertor plasma at a precisely calibrated rate, minimizing perturbation to local plasma while replicating the immediate environment of methane molecules released from a solid graphite surface more accurately than done previously. For the first time in a tokamak environment, the methane flow rate used in a puffing experiment was the same order of magnitude as that expected from laboratory experiments for intrinsic chemical sputtering. Effective photon efficiencies for CH4 injection are reported; results are found to have significant dependencies on surface conditions and the divertor operating regime. The contribution of sputtering processes to sources of C0 and C+ are assessed through measurement of background and incremental spectroscopic emissions of both physically and chemically-released sputtering products and by CI, 910 nm line profile fitting. Comparison of background and incremental emissions of chemically-released products demonstrate a dramatic drop in production of CH in cold and detached conditions. Finally, the chemical erosion yield is calculated in both attached and cold-divertor conditions and found to be much closer to that measured ex-situ in ion beam experiments than previously determined in DII-D. These observations represent a positive result for ITER which
ADX: a high field, high power density, advanced divertor and RF tokamak
LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.
The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an 'X-point target divertor' concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept
20 years of research on the Alcator C-Mod tokamak
Energy Technology Data Exchange (ETDEWEB)
Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); and others
The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental
Tokamak operation with safety factor q95 MHD stability.
Piovesan, P; Hanson, J M; Martin, P; Navratil, G A; Turco, F; Bialek, J; Ferraro, N M; La Haye, R J; Lanctot, M J; Okabayashi, M; Paz-Soldan, C; Strait, E J; Turnbull, A D; Zanca, P; Baruzzo, M; Bolzonella, T; Hyatt, A W; Jackson, G L; Marrelli, L; Piron, L; Shiraki, D
Magnetic feedback control of the resistive-wall mode has enabled the DIII-D tokamak to access stable operation at safety factor q(95) = 1.9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at a given toroidal magnetic field. In tokamaks with a divertor, the limit occurs at q(95) = 2, as confirmed in DIII-D. Since the energy confinement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a whole new high-current regime not accessible before. This result brings significant possible benefits in terms of fusion performance, but it also extends resistive-wall mode physics and its control to conditions never explored before. In present experiments, the q(95) < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.
Optimization study of normal conductor tokamak for commercial neutron source
Fujita, T.; Sakai, R.; Okamoto, A.
The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.
Toroidal flow measurement in CT injected STOR-M tokamak
Asai, Tomohiko; Morelli, Jordan; Singh, Ajay; Xiao, Chijin; Hirose, Akira; Nagata, Masayoshi; Uyama, Tadao
Compact Torus (CT) injection is a technology being developed for fueling of large tokamak reactors. It has been demonstrated in the STOR-M tokamak that tangential CT injection is capable of inducing an improved confinement mode (H-mode). It has been conjectured that tangential CT injection may enhance the toroidal rotation of the bulk tokamak plasma which is responsible for the H-mode by preventing or reducing microinstabilities[1]. In order to investigate the mechanisms of the L-H transition induced by enhanced toroidal flow (particularly that caused by CT injection), an Ion Doppler Spectroscope (IDS) has been developed. The IDS employs a 0.75 m focal length Czerny-Turner spectrometer with a resolution of 0.1 Åand a 16-channel PMT array. Data of plasma flow measurements will be presented with and without CT injection. Also, the results will be compared with toroidal flow measurement obtained using a 4-sided Mach probe in the plasma edge region. [1] S. Sen et al., Phys. Rev. Lett. 88, 185001 (2002).
Compact Torus Fueling of the STOR-M Tokamak
Xiao, C.; Hirose, A.; Zawalski, W.; White, D.; Raman, R.; Decoste, R.; Gregory, B. C.; Martin, F.
Tangential injection of accelerated compact torus (CT) has been performed on the STOR-M tokamak (R/a=46/12 cm, B_t<1 T, I_p<= 50 kA, barn_e=(0.5 - 1)×10^13 cm-3) using the University of Saskatchewan Compact Torus Injector (USCTI). The CT parameters are: m~=1 μg, v=120 km/sec, B=0.1 T and n=(2 - 4)×10^15 cm-3. After CT injection, the electron density in tokamak doubles and the poloidal β-value increases. Indications of reduction in the loop voltage and H_α emission level have also been observed. Currently, following efforts are being made: (a) to coat chromium on the electrode surface, (b) to increase the on-line baking temperature, and (c) to reduce the neutral gas load which follows the CT plasma. In addition, numerical calculation of CT motion in a tokamak magnetic field has been carried out. For horizontal injection, the initial CT magnetic dipole direction should be aligned with the CT velocity for deeper penetration. In the case of vertical injection, the CT trajectory is independent of the initial magnetic dipole direction and central penetration is facilitated by off-axis injection.
Disruption-induced poloidal currents in the tokamak wall
Energy Technology Data Exchange (ETDEWEB)
Pustovitov, V.D., E-mail: email@example.com [National Research Centre 'Kurchatov Institute', Pl. Kurchatova 1, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow 115409, Russia (Russian Federation)
Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.
Intrinsic momentum transport in up-down asymmetric tokamaks
Ball, Justin; Barnes, Michael; Dorland, William; Hammett, Gregory W; Rodrigues, Paulo; Loureiro, Nuno F
Recent work demonstrated that breaking the up-down symmetry of tokamak flux surfaces removes a constraint that limits intrinsic momentum transport, and hence toroidal rotation, to be small. We show, through MHD analysis, that ellipticity is most effective at introducing up-down asymmetry throughout the plasma. We detail an extension to GS2, a local $\\delta f$ gyrokinetic code that self-consistently calculates momentum transport, to permit up-down asymmetric configurations. Tokamaks with tilted elliptical poloidal cross-sections were simulated to determine nonlinear momentum transport. The results, which are consistent with experiment in magnitude, suggest that a toroidal velocity gradient, $\\left( \\partial u_{\\zeta i} / \\partial \\rho \\right) / v_{th i}$, of 5% of the temperature gradient, $\\left(\\partial T_{i} / \\partial \\rho \\right) / T_{i}$, is sustainable. Here $v_{th i}$ is the ion thermal speed, $u_{\\zeta i}$ is the ion toroidal mean flow, $\\rho$ is the minor radial coordinate normalized to the tokamak m...
On the computation of the disruption forces in tokamaks
Pustovitov, V. D.; Rubinacci, G.; Villone, F.
The currents and forces induced in the tokamak vacuum vessel (wall) during the disruption are calculated for different values of wall resistivity. Several consequences and new developments are derived from the general result that the global disruption force acting on the perfectly conducting wall must be exactly opposite to the similar force acting on the plasma, which is inherently small in tokamaks. This theoretical prediction is tested and confirmed here for the ITER tokamak with disruption modelled as the fast thermal quench followed by slower current quench that develops into the vertical displacement event. The plasma is simulated by the evolutionary equilibrium code CarMa0NL. One of the results is that the computed integral force on a perfectly conducting wall is zero at each instant during a disruption. This in turn highlights the importance of having good models for the plasma (in which the equilibrium constraint is explicitly imposed) and for the structures (able to correctly describe the induced currents and the resistive effects). The dependence of the disruption force on the magnetic field penetration through the wall is demonstrated. Also the concept of a disruption force damper is proposed, able to 'absorb' a significant part of the force that would arise on a resistive wall during a disruption.
Core fueling to produce peaked density profiles in large tokamaks
Energy Technology Data Exchange (ETDEWEB)
Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Attenberger, S.E.; Houlberg, W.A.; Milora, S.L. [Oak Ridge National Lab., TN (United States)
Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. We show that with radially ``hollow`` diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles which are peaked off-axis. The fueling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fueling does not require MeV particle energy. Even with beam voltages of {approximately}200 keV, however, exceptionally good particle confinement, {tau}{sub p} {much_gt} {tau}{sub E} is required to achieve net electrical power generation. In system with no power production requirement (e.g., neutron sources) neutral beam fueling should be capable of producing peaked density profiles in devices as large as ITER. Fueling systems with low energy cost per particle (such as cryogenic pellet injection) must be used in power producing tokamaks when {tau}{sub p} {approximately} {tau}{sub E}. Simulations with pellet injection speeds of 7 km/sec show the peaking factor, n{sub eo}/{l_angle}n{sub e}{r_angle}, approaching 2.
Effect of Magnetic Islands on Divertors in Tokamaks and Stellarators
Punjabi, Alkesh; Boozer, Allen
Divertors are required for handling the plasma particle and heat exhausts on the walls in fusion plasmas. Relatively simple methods, models, and maps from field line Hamiltonian are developed to better understand the interaction of strong plasma shaping and magnetic islands on the size and behavior of the magnetic flux tubes that go from the plasma edge to the wall in non-axisymmetric system. This approach is applicable not only in tokamaks but also in stellarators. Stellarator diverters in which magnetic islands are dominant are called resonant and when shaping is dominant are called non-resonant. Optimized stellarators generally have sharp edges on their surface, but unlike the case for tokamaks these edges do not encircle the entire plasma, so they do not define an edge value for the rotational transform. The approach is used in the DIII-D tokamak. Computation results are consistent with the predictions of the models. Further simulations are being done to understand why the transition from an effective cubic to a linear increase in loss time and area of footprint occurs and whether this increase is discontinuous or not. This work is supported by the US DOE Grants DE-FG02-01ER54624 and DE-FG02-04ER54793 to Hampton University and DE-FG02-95ER54333 to Columbia University. This research used resources of the NERSC, supported by the Office of Science, US DOE, under Contract No. DE-AC02-05CH11231.
Energy and particle core transport in tokamaks and stellarators compared
Energy Technology Data Exchange (ETDEWEB)
Beurskens, Marc; Angioni, Clemente; Beidler, Craig; Dinklage, Andreas; Fuchert, Golo; Hirsch, Matthias; Puetterich, Thomas; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald/Garching (Germany)
The paper discusses expectations for core transport in the Wendelstein 7-X stellarator (W7-X) and presents a comparison to tokamaks. In tokamaks, the neoclassical trapped-particle-driven losses are small and turbulence dominates the energy and particle transport. At reactor relevant low collisionality, the heat transport is limited by ion temperature gradient limited turbulence, clamping the temperature gradient. The particle transport is set by an anomalous inward pinch, yielding peaked profiles. A strong edge pedestal adds to the good confinement properties. In traditional stellarators the 3D geometry cause increased trapped orbit losses. At reactor relevant low collisionality and high temperatures, these neoclassical losses would be well above the turbulent transport losses. The W7-X design minimizes neoclassical losses and turbulent transport can become dominant. Moreover, the separation of regions of bad curvature and that of trapped particle orbits in W7-X may have favourable implications on the turbulent electron heat transport. The neoclassical particle thermodiffusion is outward. Without core particle sources the density profile is flat or even hollow. The presence of a turbulence driven inward anomalous particle pinch in W7-X (like in tokamaks) is an open topic of research.
Oceanographic measurements from the Texas Automated Buoy System (TABS)
National Oceanic and Atmospheric Administration, Department of Commerce — The Texas Automated Buoy System contains daily oceanographic measurements from seven buoys off the Texas coast from Brownsville to Sabine. The Texas General Land...
Determination of radial location of rotating magnetic islands by use of poloidal soft x-ray detector arrays in the STOR-M tokamak
Dreval, M.; Xiao, C.; Elgriw, S.; Trembach, D.; Wolfe, S.; Hirose, A.
A technique is presented for determining the radial location of the rotating magnetic islands in the STOR-M tokamak by use of soft x-ray (SXR) detector arrays. The location is determined by examining the difference in the integrated SXR emission intensities through two adjacent lines of sight. A model for calculating dependence of the line integrated SXR emission intensity on the radius, the mode numbers and the magnetic island geometry, has been developed. The SXR difference signal shows phase inversion when the impact parameter of the line of sight sweeps across the magnetic islands. Experimentally, the difference SXR signals significantly reduce noise and suppress the influence of background plasma fluctuations through common mode rejection when a dominant mode exists in the STOR-M tokamak. The radial locations of the m = 2 magnetic islands have been determined under several experimental conditions in the STOR-M discharges. With the decrease in the tokamak discharge current and thus the increase of the safety factor at the edge, the radial location of the m = 2 magnetic islands has been found to move radially inward.
Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors
Energy Technology Data Exchange (ETDEWEB)
Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: email@example.com [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)
Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.
The Texas A&M Radioisotope Production and Radiochemistry Program
Energy Technology Data Exchange (ETDEWEB)
Akabani, Gamal [Texas A & M Univ., College Station, TX (United States). Dept. of Nuclear Engineering. Dept. of Veterinary Integrative Biosciences
The main motivation of the project at Texas A&M University was to carry out the production of critically needed radioisotopes used in medicine for diagnostics and therapy, and to establish an academic program in radionuclide production and separation methods. After a lengthy battle with the Texas A&M University Radiation Safety Office, the Texas Department of State Health Services granted us a license for the production of radionuclides in July 2015, allowing us to work in earnest in our project objectives. Experiments began immediately after licensing, and we started the assembly and testing of our target systems. There were four analytical/theoretical projects and two experimental target systems. These were for At-211 production and for Zn-62/Cu-62 production. The theoretical projects were related to the production of Mo-99/Tc-99m using a) a subcritical aqueous target system and b) production of Tc-99m from accelerator-generated Mo-99 utilizing a photon-neutron interaction with enriched Mo-100 targets. The two experimental projects were the development of targetry systems and production of At-211 and Zn-62/Cu-62 generator. The targetry system for At-211 has been tested and production of At-211 is chronic depending of availability of beam time at the cyclotron. The installation and testing of the targetry system for the production of Zn-62/Cu-62 has not been finalized. A description of the systems is described. The academic program in radionuclide production and separation methods was initiated in the fall of 2011 and, due to the lack of a radiochemistry laboratory, it was suspended. We expect to re-start the academic program at the Texas A&M Institute for Preclinical Studies under the Molecular Imaging Program.
The Texas A&M Radioisotope Production and Radiochemistry Program
Energy Technology Data Exchange (ETDEWEB)
Akabani, Gamal [Texas A & M Univ., College Station, TX (United States)
The main motivation of the project at Texas A&M University was to carry out the production of critically needed radioisotopes used in medicine for diagnostic and therapy, and to establish an academic program in radionuclide production and separation methods. After a lengthy battle with the Texas A&M University Radiation Safety Office, the Texas Department of State Health Services granted us a license for the production of radionuclides in July 2015, allowing us to work in earnest in our project objectives. Experiments began immediately after licensing, and we started the assembly and testing of our target systems. There were four analytical/theoretical projects and two experimental target systems. These were for At-211 production and for Zn- 62/Cu-62 production. The theoretical projects were related to the production of Mo-99/Tc-99m using (a) a subcritical aqueous target system and (b) production of Tc-99m from accelerator-generated Mo-99 utilizing a photon-neutron interaction with enriched Mo-100 targets. The two experimental projects were the development of targetry systems and production of At-211 and Zn-62/Cu-62 generator. The targetry system for At-211 has been tested and production of At-211 is chronic depending of availability of beam time at the cyclotron. The installation and testing of the targetry system for the production of Zn-62/Cu-62 has not been finalized. A description of the systems is described. The academic program in radionuclide production and separation methods was initiated in the fall of 2011; due to the lack of a radiochemistry laboratory, it was suspended. We expect to re-start the academic program at the Texas A&M Institute for Preclinical Studies under the Molecular Imaging Program.
Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I; Calculo sobre una modificacion al campo magnetico toroidal del Tokamak Novillo. Parte I
Energy Technology Data Exchange (ETDEWEB)
Chavez A, E.; Melendez L, L.; Colunga S, S.; Valencia A, R.; Lopez C, R.; Gaytan G, E
The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)
Transport studies in TJ-I tokamak from steady and perturbative methods
Energy Technology Data Exchange (ETDEWEB)
Pardo, C.; Rodriguez-Yunta, A.; Vega, J.; Branas, B.; Estrada, T.; Ochando, M.A.; Tabares, F.L.; Zurro, B. (Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain))
Transport understanding is an essential task for the development of a future tokamak fusion reactor. In a general formulation, the dependence of particle and electron energy fluxes on density and temperature gradients, may be written as: [Gamma]=-D[nabla]n - D[sub T]n [nabla]T/T - nV, q=Q-5/2[Gamma]T=-[chi][sub n]T[nabla]n-[chi]n[nabla]T-nTU where both fluxes are related to both gradients, and the particle and energy pinches V and U may depend on any other force as could be the parallel electric field. The transport coefficients must be expected to be functions of local plasma parameters such as B, q, n, T, [nabla]n, [nabla]T, ... etc. This means that the fluxes may be non-linear functions of the gradients. Transport analysis in the steady state gives values, from experimental data, for fluxes and gradients. This is not enough to determine the values of the six transport coefficients. A perturbative experiment, such as the simultaneous measurement of density and temperature pulses induced by a sawtooth collapse, give us the incremental transport coefficients or the derivatives of the fluxes with respect to the gradients. By making a coupled analysis of both pulses, we can obtain values for the four derivatives: [partial derivative][Gamma][partial derivative][nabla]n, [partial derivative][Gamma]/[partial derivative][nabla]T, [partial derivative]q/[partial derivative][nabla]n and [partial derivative]q/[partial derivative][nabla]T. The combination of both steady and perturbative studies in discharges with different plasma parameters could give us a better picture of transport processes in a tokamak. (author) 6 refs., 5 figs.
Transport simulation of ELM pacing by pellet injection in tokamak plasmas
Kim, Ki Min; Na, Yong-Su; Hong, Sang Hee; Lang, P. T.; Alper, B.; contributors, JET-EFDA
This paper deals mainly with the numerical simulation on edge localized mode (ELM) pacing by pellet injection that is useful for fuelling and control of plasma profiles to achieve enhanced tokamak operations. The fuelling and pellet-induced ELMs are simulated with a 1.5-dimensional core transport code, which includes a neutral gas shielding model and a grad-B drift model for pellet deposition in H-mode tokamak plasmas. Fuelling and ELM pacing experiments by pellet injections at JET are introduced as a current experimental approach. For the description of ELM triggering by pellet injection based on ideal ballooning mode criteria, three possible models are suggested and discussed on their ELM characteristics, respectively: (i) the density enhanced ELMs in the post-pellet phase, (ii) the modification of the surface averaged pressure profiles in a transport time scale and (iii) the local increase in the pressure (density and/or temperature) gradients perturbed by pellets. Among them, the pellet-induced density perturbation model is adopted, in practice, to carry out an ELM pacing simulation in preparation for future experiments in KSTAR. The numerical simulation shows that the artificially induced ELM by pellets releases the reduced energy bursts, compared with spontaneous ELMs. The energy loss per burst by the pellet-induced ELM turns out to be much smaller than that by the spontaneous ELM as the pellet injection frequency becomes higher in ELM pacing. Based on the simulation results showing good agreement with the general ELM characteristics observed in pellet pacing experiments, the ELM pacing by pellet injection is very promising for mitigating the ELM energy bursts to the divertor by controlling the injection frequency.
Twenty Years of Research on the Alcator C-Mod Tokamak
Greenwald, Martin
Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE
Including collisions in gyrokinetic tokamak and stellarator simulations
Energy Technology Data Exchange (ETDEWEB)
Kauffmann, Karla
Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal
Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak
Hennen, B.A.; Westerhof, E.; Nuij, Pwjm; M.R. de Baar,; Steinbuch, M.
Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron
Tokamak plasma self-organization and the possibility to have the peaked density profile in ITER
Razumova, K. A.; Andreev, V. F.; Kislov, A. Y.; Kirneva, N. A.; Lysenko, S. E.; Pavlov, Y. D.; Shafranov, T. V.; Donne, A. J. H.; Hogeweij, G. M. D.; Spakman, G. W.; R. Jaspers,; Kantor, M.; Walsh, M.
The self-organization of a tokamak plasma is a fundamental turbulent plasma phenomenon, which leads to the formation of a self-consistent pressure profile. This phenomenon has been investigated in several tokamaks with different methods of heating. It is shown that the normalized pressure profile
Indication of GAM and electrode biasing effect on GAM in STOR-M Tokamak
Nakajima, Masaru; Basu, Debjyoti; Rohollahi, Akbar; McColl, David; Adegun, Joseph; Xiao, Chijin; Hirose, Akira
STOR-M is a small, iron-core, limiter based tokamak with major and minor radii of 46 cm and 12 cm, respectively. Recent experimental studies have been carried out to detect GAM in this machine. Four Langmuir probe sets have been inserted into the plasma. The first three Langmuir probe sets are located in the same toroidal plane, inserted from top, bottom and outboard of the mid-plane. The fourth set is inserted from the outboard of the mid-plane, but toroidally separated from the others by 90°. Each probe set consists of three Langmuir probe tips for Isat, floating potential and I-V curve measurements. Preliminary experimental results with slightly higher edge-qa (within 5 to 6) clearly indicate a 180° phase difference between the up and down density fluctuation signals near 20 kHz. The floating potential fluctuation signals from the same locations at the same frequency showed no observable phase shift. Preliminary data indicate the presence of conventional GAM in STOR-M. In the near future, magnetic fluctuation properties of GAM oscillations in STOR-M as well as the responses of the GAM properties to electrode biasing will be studied. Detailed experimental results will be presented.
Transport and turbulence in a magnetized plasma (application to tokamak plasmas); Transport et turbulence dans un plasma magnetise (application aux plasmas de tokamaks)
Energy Technology Data Exchange (ETDEWEB)
Sarazin, Y
This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.
Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions
Energy Technology Data Exchange (ETDEWEB)
Wuerz, H. [Kernforschungszentrum Karlsruhe, INR, Postfach 36 40, D-76021 Karlsruhe (Germany); Arkhipov, N.I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Bakhtin, V.P. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Konkashbaev, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Landman, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Safronov, V.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Toporkov, D.A. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Zhitlukhin, A.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation)
The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.)).
Angiostrongylus cantonensis Meningitis and Myelitis, Texas, USA.
Al Hammoud, Roukaya; Nayes, Stacy L; Murphy, James R; Heresi, Gloria P; Butler, Ian J; Pérez, Norma
Infection with Angiostrongylus cantonensis roundworms is endemic in Southeast Asia and the Pacific Basin. A. cantonensis meningitis and myelitis occurred in summer 2013 in a child with no history of travel outside of Texas, USA. Angiostrongyliasis is an emerging neurotropic helminthic disease in Texas and warrants increased awareness among healthcare providers.
Plasma heating systems planned for the Argonne experimental power reactor
Energy Technology Data Exchange (ETDEWEB)
Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.
A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.
Parasitic Momentum Flux in the Tokamak Core
Stoltzfus-Dueck, T
A geometrical correction to the E x B drift causes an outward flux of cocurrent momentum whenever electrostatic potential energy is transferred to ion parallel flows. The robust symmetry breaking follows from the free energy flow in phase space and does not depend on any assumed linear eigenmode structure, acting both for axisymmetric fluctuations (such as geodesic acoustic modes) as well as more general nonaxisymmetric fluctuations. The resulting rotation peaking is countercurrent and scales as electron temperature over plasma current. This peaking mechanism can only act when fluctuations are low-frequency enough to excite ion parallel flows, which may explain some recent experimental observations related to rotation reversals.
Characterisation, modelling and control of advanced scenarios in the european tokamak jet; Caracterisation, modelisation et controle des scenarios avances dans le tokamak europeen jet
Energy Technology Data Exchange (ETDEWEB)
Tresset, G
The advanced scenarios, developed for less than ten years with the internal transport barriers and the control of current profile, give rise to a 'new deal' for the tokamak as a future thermonuclear controlled fusion reactor. The Joint European Torus (JET) in United Kingdom is presently the most powerful device in terms of fusion power and it has allowed to acquire a great experience in these improved confinement regimes. The reduction of turbulent transport, considered now as closely linked to the shape of current profile optimised for instance by lower hybrid current drive or the self-generated bootstrap current, can be characterised by a dimensionless criterion. Most of useful information related to the transport barriers are thus available. Large database analysis and real time plasma control are envisaged as attractive applications. The so-called 'S'-shaped transport models exhibit some interesting properties in fair agreement with the experiments, while the non-linear multivariate dependencies of thermal diffusivity can be approximated by a neural network, suggesting a new approach for transport investigation and modelling. Finally, the first experimental demonstrations of real time control of internal transport barriers and current profile have been performed on JET. Sophisticated feedback algorithms have been proposed and are being numerically tested to achieve steady-state and efficient plasmas. (author)
Joint Services Electronics Program: Basic Research in Electronics (JSEP) at the University of Texas at Austin. Appendix.
at the 7th tnt Conf. on A nalv-sis and Optimnization of Syvstems. Antibes, France . GRIZZLE,. W., and NuIJiER, H., 1985. Zeros at infinity for...crystal metal surfaces Marshall Onellion and J. L. Erskine Department of Ph.isLcs, ,niversin of Texas, Austin, Texas (Received 30 May 19851...coverage and null result obtained by Erskine ") under similar experimental spatial distributions. Adatom concentration was accurately conditions for Xe on W
Monitoring of density in tokamaks: pumping and gas injection; Controle de la densite dans les tokamaks: pompage et injection de matiere
Energy Technology Data Exchange (ETDEWEB)
Dejarnac, R
In thermonuclear fusion devices, controlling the Deuterium-Tritium fuel density and exhausting the Helium ashes is a crucial point. This is achieved by fuelling the discharges by different methods (gas puffing and pellet injection are the most commonly used) and by implementing pumping devices at the plasma periphery. These two issues are treated in this work, both from an experimental and a modelling point of view, using the neutral transport code EIRENE as main tool for our studies. As far as pumping is concerned, we have modelled the outboard pump limiter of the Tore Supra tokamak with the EIRENE code to which we coupled a plasma module specially developed to simulate the neutrals and the plasma in a coherent way. This allowed to validate the code against experimental data. As far as plasma fuelling is concerned, we present here an original method: the supersonic pulsed gas injection (SPGI). This intermediate method between conventional gas puff (GP) and pellet injection was designed and tested at Tore Supra. It consists of injecting very dense and short gas puffs at high speed into the plasma. Experimentally, SPGI was found to have a better fuelling efficiency than GP and to lead to a strong plasma cooling. The mechanisms responsible for this improved efficiency are analysed by modelling, using the EIRENE code to determine the ionisation source and a 1 D transport model to reproduce the plasma density response. At last, an extrapolation of the present injector is presented, discussing the possibility to obtain a radial drift of the injected matter as observed in the case of high field side pellet injection. (author)
Development of coupling systems at the hybrid frequency for the non-inductive current generation inside a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak
Energy Technology Data Exchange (ETDEWEB)
Berio, S. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Aix-Marseille-1 Univ., 13 - Marseille (France)
Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead to the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a on-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (authors) 53 refs.
Development of hybrid frequency couplers for non-inductive current drive in a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak
Energy Technology Data Exchange (ETDEWEB)
Berio, St.
Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead of the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a non-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (author). 53 refs.
Magneto-hydro-dynamic limits in spherical tokamaks
Hender, T. C.; Allfrey, S. J.; Akers, R.; Appel, L. C.; Bevir, M. K.; Buttery, R. J.; Gryaznevich, M.; Jenkins, I.; Kwon, O. J.; McClements, K. G.; Martin, R.; Medvedev, S.; Nightingale, M. P. S.; Ribeiro, C.; Roach, C. M.; Robinson, D. C.; Sharapov, S. E.; Sykes, A.; Villard, L.; Walsh, M. J.
The operational limits observed in spherical tokamaks, notably the small tight aspect ratio tokamak (START) device [A. Sykes et al., Nucl. Fusion 32, 694 (1992)], are consistent with those found in conventional aspect ratio tokamaks. In particular the highest β achieved (˜40%) is consistent with an ideal magneto-hydro-dynamic (MHD) Troyon type limit, the upper limit on density is well described by the Greenwald density (πa2n¯e/Ip˜1) and the normalized current (Ip/aBt) is limited such that q95≳2. Stability calculations indicate scope for increasing both normalized β and normalized current beyond the values so far achieved, although wall stabilization is generally needed for low-n modes. In double null configurations current terminating disruptions occur at each of the operational boundaries, though the current quench tends to be slow at the density limit and disruptions at high β may be due to the low q. In early limiter START discharges, before the divertor coils were installed, disruptions rarely occurred. Instead internal reconnection events which have all the characteristics of a disruption except the current quench occurred. These various disruptive behaviors are explained in terms of a model in which helicity is conserved during the disruption. Due to the low toroidal field beam ions in START, and α particles in a ST power plant, are super-Alfvénic. This gives the possibility for toroidal Alfvén eigenmodes (TAEs) to occur and such modes are frequently observed in START neutral beam injection (NBI) discharges, but seem to be benign. The features of these observed TAEs are shown to be in agreement with MHD calculations.
On the minimum circulating power of steady state tokamaks
Energy Technology Data Exchange (ETDEWEB)
Itoh, K.; Itoh, S.; Fukuyama, A.; Yagi, M.
Circulating power for the sustenance and profile control of the steady state tokamak plasmas is discussed. The simultaneous fulfillment of the MHD stability at high beta value, the improved confinement and the stationary equilibrium requires the rotation drive as well as the current drive. In addition to the current drive efficiency, the efficiency for the rotation drive is investigated. The direct rotation drive by the external torque, such as the case of beam injection, is not efficient enough. The mechanism and the magnitude of the spontaneous plasma rotation are studied. (author).
Hall effect on tearing mode instabilities in tokamak
Zhang, W.; Ma, Z. W.; Wang, S.
The tearing mode instability is one of the most important dynamic processes in space and laboratory plasmas. Hall effects, resulting from the decoupling of electron and ion motions, can cause fast development and rotation of the perturbation structure of the tearing mode. A high-accuracy nonlinear magnetohydrodynamics code is developed to study Hall effects on the evolution of tearing modes in the Tokamak geometry. It is found that the linear growth rate increases with the increase in the ion skin depth and the self-consistently generated rotation can greatly alter the dynamic behavior of the double tearing mode.
System design of toroidal field power supply of CDD tokamak
Energy Technology Data Exchange (ETDEWEB)
Liu, Zheng Zhi
This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs.
Dynamic modeling of transport and positional control of tokamaks
Energy Technology Data Exchange (ETDEWEB)
Jardin, S.C.; Pomphrey, N.; DeLucia, J.
We describe here a numerical model of a free boundary axisymmetric tokamak plasma and its associated control systems. The plasma is modeled with a hybrid method using two-dimensional velocity and flux functions with surface-averaged MHD equations describing the evolution of the adiabatic invariants. Equations are solved for the external circuits and for the effects of eddy currents in nearby conductors. The method is verified by application to several test problems and used to simulate the formation of a bean-shaped plasma in the PBX experiment.
Plasma facing components design of KT-2 tokamak
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In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Xu, Chao Yin [China Univ. of Science and Technology, Hefei, AH (China)
The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs.
Spectral measurements of runway electrons in the TEXTOR tokamak
Energy Technology Data Exchange (ETDEWEB)
Kudyakov, Timur
The generation of multi-MeV runaway electrons is a well known effect related to the plasma disruptions in tokamaks. The runaway electrons can substantially reduce the lifetime of the future tokamak ITER. In this thesis physical properties of runaway electrons and their possible negative effects on ITER have been studied in the TEXTOR tokamak. A new diagnostic, a scanning probe, has been developed to provide direct measurements of the absolute number of runaway electrons coming from the plasma, its energy distribution and the related energy load in the material during low density (runaway) discharges and during disruptions. The basic elements of the probe are YSO crystals which transform the energy of runaway electrons into visible light which is guided via optical fibres to photomultipliers. In order to obtain the energy distribution of runaways, the crystals are covered with layers of stainless steel (or tungsten in two earlier test versions) of different thicknesses. The final probe design has 9 crystals and can temporally and spectrally resolve electrons with energies between 4 MeV and 30 MeV. The probe is tested and absolutely calibrated at the linear electron accelerator ELBE in Rossendorf. The measurements are in good agreement with Monte Carlo simulations using the Geant4 code. The runaway transport in the presence of the internal and externally applied magnetic perturbations has been studied. The diffusion coefficient and the value of the magnetic fluctuation for runaways were derived as a function of B{sub t}. It was found that an increase of runaway losses from the plasma with the decreasing toroidal magnetic field is accompanied with a growth of the magnetic fluctuation in the plasma. The magnetic shielding picture could be confirmed which predicts that the runaway loss occurs predominantly for low energy runaways (few MeV) and considerably less for the high energy ones. In the case of the externally applied magnetic perturbations by means of the dynamic
Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor
Energy Technology Data Exchange (ETDEWEB)
Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.
Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.
Dynamics of nano-dust in tokamak edge plasma
Energy Technology Data Exchange (ETDEWEB)
Krasheninnikov, S.I., E-mail: email@example.com [University of California at San Diego, La Jolla, CA 92093 (United States); Soboleva, T.K. [ICN, Universidad Nacional Autonoma de Mexico, Mexico DF (Mexico); Mendis, D.A. [University of California at San Diego, La Jolla, CA 92093 (United States)
The dynamics of nano-scale dust for tokamak edge conditions is reviewed. It is shown that unlike micron-scale grains, where the ion-grain friction is the dominant force acting on the grain, nano-dust dynamics is the subject of both friction and Lorentz forces. Possible impact of nano-dust on MARFE is investigated and it is found that dust, providing plasma particle sink and thus causing plasma flow, can play a dominant role in the localization of impurity in low temperature region. It is also shown that dust can significantly reduce the growth rate of flute instability.
Steady-state current drive in tokamaks workshop summary
Energy Technology Data Exchange (ETDEWEB)
The purpose of the workshop was to identify the most promising techniques and to outline the expectations for tokamak reactor concepts. The group which included beam and rf specialists were asked to assist in the preparation of specific recommendations for the establishment of a program directed at demonstrating current drive. This report includes the recommendations and conclusions as prepared by A. Bers, D. Jassby, and T. Hsu and the summary papers submitted by the participants describing the various experiments and studies presented at the workshop.
Nonlinear saturation of ballooning modes in tokamaks and stellarators
Bauer, F.; Garabedian, P.; Betancourt, O.
The spectral code BETAS computes plasma equilibrium in a toroidal magnetic field B = [unk]s × [unk]Ψ with remarkable accuracy because the finite difference scheme employed in the radial direction allows for discontinuities of the flux function Ψ across the nested surfaces s = const. Instability of higher modes in stellarators like the Heliotron E can be detected in roughly an hour on the best supercomputers by calculating bifurcated equilibria that are defined over just one field period. The method has been validated by comparing results about nonlinear saturation of ballooning modes in tokamaks with numerical data from the PEST code. PMID:16593984
Simulations of Boundary Turbulence in Tokamak Experiments
Energy Technology Data Exchange (ETDEWEB)
Nevins, W M; Xu, X Q; Carlstrom, T N; Cohen, R H; Groebner, R; Jennings, T; LaBombard, B; Maqueda, R A; Mazurenko, A; McKee, G R; Moyer, R; Mossessian, D; Porkolab, M; Porter, G D; Rensink, M E; Rhodes, T L; Rognlien, T D; Rost, C; Snipes, J; Stotler, D P; Terry, J; Zweben, S
Comparisons between the boundary plasma turbulence observed in the BOUT code and experiments on C-Mod, NSTX, and DIII-D are presented. BOUT is a 3D non-local electromagnetic turbulence simulation code which models boundary-plasma turbulence in a realistic divertor geometry using the modified Braginskii equations for plasma vorticity, density, the electron and ion temperatures and parallel momenta. Many features of the Quasi-Coherent (QC) mode, observed at high densities during enhanced D-alpha (EDA) H-Mode in Alcator C-Mod, are reproduced in BOUT simulations. The spatial structure of boundary plasma turbulence as observed by gas puff imaging (GPI) from discharges on NSTX and C-Mod are in general (NSTX) to good (CMod) agreement with BOUT simulations. Finally, BOUT simulations of DIII-D L-mode experiments near the Hmode transition threshold are in broad agreement with the experimental results.
Design of set-point weighting PIλ + Dμ controller for vertical magnetic flux controller in Damavand tokamak.
Rasouli, H; Fatehi, A
In this paper, a simple method is presented for tuning weighted PI(λ) + D(μ) controller parameters based on the pole placement controller of pseudo-second-order fractional systems. One of the advantages of this controller is capability of reducing the disturbance effects and improving response to input, simultaneously. In the following sections, the performance of this controller is evaluated experimentally to control the vertical magnetic flux in Damavand tokamak. For this work, at first a fractional order model is identified using output-error technique in time domain. For various practical experiments, having desired time responses for magnetic flux in Damavand tokamak, is vital. To approach this, at first the desired closed loop reference models are obtained based on generalized characteristic ratio assignment method in fractional order systems. After that, for the identified model, a set-point weighting PI(λ) + D(μ) controller is designed and simulated. Finally, this controller is implemented on digital signal processor control system of the plant to fast/slow control of magnetic flux. The practical results show appropriate performance of this controller.
Assessing Earth and Environmental Science Enrollment Trends in Texas Public High Schools
Sanders, Joan G.
Scope and Method of Study: This study assesses the status of Earth and environmental sciences education in Texas Public High Schools by analyzing enrollment proportions of 11th and 12th grade students in 607 Independent School Districts (ISD) for the 2010-2011 academic school year using a quantitative, non-experimental alpha research design. This…
Plasma diagnostics in spherical tokamaks with silicon charged-particle detectors
Energy Technology Data Exchange (ETDEWEB)
Netepenko, A., E-mail: email@example.com; Boeglin, W. U. [Department of Physics, Florida International University, Miami, Florida 33199 (United States); Darrow, D. S.; Ellis, R.; Sibilia, M. J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)
Detection of charged fusion products, such as protons and tritons resulting from D(d, p) t reactions, can be used to determine the position and time dependent fusion reaction rate profile in spherical tokamak plasmas with neutral beam heating. We have developed a prototype instrument consisting of 6 ion-implanted-silicon surface barrier detectors combined with collimators in such a way that each detector can accept 3 MeV protons and 1 MeV tritons and thus provides a curved view across the plasma cross section. The combination of the results from all six detectors will provide information on the spatial distribution of the fusion reaction rate. The expected time resolution of about 1 ms makes it possible to study changes in the reaction rate due to slow variations in the neutral beam density profile, as well as rapid changes resulting from MHD instabilities. Details of the new instrument, its data acquisition system, simulation results, and electrical noise testing results are discussed in this paper. First experimental data are expected to be taken during the current experimental campaign at NSTX-U.
First measurement of the edge charge exchange recombination spectroscopy on EAST tokamak
Energy Technology Data Exchange (ETDEWEB)
Li, Y. Y., E-mail: firstname.lastname@example.com; Fu, J.; Jiang, D.; Lyu, B.; Hu, C. D.; Wan, B. N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yin, X. H.; Feng, S. Y. [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); Shi, Y. J. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei (China); Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of); Yi, Y.; Ye, M. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei (China); Zhou, X. J. [Anhui Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Hefei 230031 (China)
An edge toroidal charge exchange recombination spectroscopy (eCXRS) diagnostic, based on a heating neutral beam injection (NBI), has been deployed recently on the Experimental Advanced Superconducting Tokamak (EAST). The eCXRS, which aims to measure the plasma ion temperature and toroidal rotation velocity in the edge region simultaneously, is a complement to the exiting core CXRS (cCXRS). Two rows with 32 fiber channels each cover a radial range from ∼2.15 m to ∼2.32 m with a high spatial resolution of ∼5-7 mm. Charge exchange emission of Carbon VI CVI at 529.059 nm induced by the NBI is routinely observed, but can be tuned to any interested wavelength in the spectral range from 400 to 700 nm. Double-slit fiber bundles increase the number of channels, the fibers viewing the same radial position are binned on the CCD detector to improve the signal-to-noise ratio, enabling shorter exposure time down to 5 ms. One channel is connected to a neon lamp, which provides the real-time wavelength calibration on a shot-to-shot basis. In this paper, an overview of the eCXRS diagnostic on EAST is presented and the first results from the 2015 experimental campaign will be shown. Good agreements in ion temperature and toroidal rotation are obtained between the eCXRS and cCXRS systems.
The generalized accessibility and spectral gap of lower hybrid waves in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Takahashi, Hironori
The generalized accessibility of lower hybrid waves, primarily in the current drive regime of tokamak plasmas, which may include shifting, either upward or downward, of the parallel refractive index (n{sub {parallel}}), is investigated, based upon a cold plasma dispersion relation and various geometrical constraint (G.C.) relations imposed on the behavior of n{sub {parallel}}. It is shown that n{sub {parallel}} upshifting can be bounded and insufficient to bridge a large spectral gap to cause wave damping, depending upon whether the G.C. relation allows the oblique resonance to occur. The traditional n{sub {parallel}} upshifting mechanism caused by the pitch angle of magnetic field lines is shown to lead to contradictions with experimental observations. An upshifting mechanism brought about by the density gradient along field lines is proposed, which is not inconsistent with experimental observations, and provides plausible explanations to some unresolved issues of lower hybrid wave theory, including generation of {open_quote}seed electrons.{close_quote}
Radiative instabilities in the tokamak scrape-off layer during edge-localized mode activity
Energy Technology Data Exchange (ETDEWEB)
Helander, P.; Krasheninnikov, S.I. [Massachusetts Institute of Technology, Plasma Fusion Center, Cambridge, Massachusetts 02139 (United States); Morozov, D.K.; Soboleva, T.K. [Instituto de Ciencias Nucleares, Universidad Nacional Autonoma de Mexico, Mexico D.F. (Mexico)
In order to reduce the heat flux entering the divertor, it is desirable to have strong impurity radiation in the scrape-off layer (SOL) of reactor-size tokamaks like the International Thermonuclear Experimental Reactor [{ital International} {ital Thermonuclear} {ital Experimental} {ital Reactor} ({ital ITER}) {ital Conceptual} {ital Design} {ital Activity} {ital Final} {ital Report}, ITER Documentation Series No. 16 (International Atomic Energy Agency, Vienna, 1991)]. Under such circumstances, however, the SOL plasma is likely to be unstable to the radiative condensation instability. In the present paper, an investigation is undertaken to study the effects of edge-localized mode (ELM) activity on this instability. In the linear regime, it is demonstrated that high-frequency (``grassy``) ELM`s may parametrically excite acoustic waves. The possibility of nonlinear radiative collapse with concomitant stratification of the plasma is discussed, and solutions describing nonlinear traveling waves are derived in which the plasma goes over from equilibrium state to another. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}.
33 CFR 165.804 - Snake Island, Texas City, Texas; mooring and fleeting of vessels-safety zone.
... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Snake Island, Texas City, Texas... Guard District § 165.804 Snake Island, Texas City, Texas; mooring and fleeting of vessels—safety zone. (a) The following is a safety zone: (1) The west and northwest shores of Snake Island; (2) The...
75 FR 36710 - The Texas Engineering Experiment Station/Texas A&M University System; Notice of Acceptance for...
... COMMISSION The Texas Engineering Experiment Station/Texas A&M University System; Notice of Acceptance for... Facility Operating License No. R-83 (``Application''), which currently authorizes the Texas Engineering Experiment Station/Texas A&M University System (TEES, the licensee) to operate the Nuclear Science Center...
Hand-eye coordinative remote maintenance in a tokamak vessel
Energy Technology Data Exchange (ETDEWEB)
Qiu, Qiang, E-mail: email@example.com; Gu, Kai, E-mail: email@example.com; Wang, Pengfei, E-mail: firstname.lastname@example.com; Bai, Weibang, E-mail: email@example.com; Cao, Qixin, E-mail: email@example.com
Highlights: • If there is not rotation between the visual coordinate frame (O{sub e}X{sub e}Y{sub e}) and hand coordinate frame (O{sub h}X{sub h}Y{sub h}), a person can coordinate the movement between hand and eye easily. • We establish an alignment between the movement of the operator's hand and the visual scene of the end-effector as displayed on the monitor. • A potential function is set up in a simplified vacuum vessel model to provide a fast collision checking, and the alignment between repulsive force and Omega 7 feedback force is accomplished. • We carry out an experiment to evaluate its performance in a remote handling task. - Abstract: The reliability is vitally important for the remote maintenance in a tokamak vessel. In order to establish a more accurate and safer remote handling system, a hand-eye coordination method and an artificial potential function based collision avoidance method were proposed in this paper. At the end of this paper, these methods were implemented to a bolts tightening maintenance task, which was carried out in our 1/10 scale tokamak model. Experiment results have verified the value of the hand-eye coordination method and the collision avoidance method.
Lithium beam diagnostic system on the COMPASS tokamak
Energy Technology Data Exchange (ETDEWEB)
Anda, G.; Bencze, A. [Wigner – RCP, HAS, Budapest (Hungary); Berta, M., E-mail: firstname.lastname@example.com [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Hacek, P. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)
Highlights: • Li-beam diagnostic system on the COMPASS tokamak is an improved and compact system to allow testing of Atomic Beam Probe. • The possibility to measure background corrected density profiles on the few microseconds time scale. • First Li-beam diagnostic system with recirculating neutralizer. • The system includes the redesigned ion source with longer lifetime. - Abstract: An improved lithium beam based beam emission spectroscopy system – installed on COMPASS tokamak – is described. The beam energy enhanced up to 120 keV for Atomic Beam Probe measurement. The size of the ion source is doubled, using a newly developed thermionic heater instead of the conventionally used heating (tungsten or molybdenum) filament. The neutralizer is also improved. It produces the same sodium vapor in a cell but minimize the loss condensing the vapor on a cold surface which is led back (in fluid state) into the sodium oven. This way we call it recirculating neutralizer. The observation system consists of a CCD camera and an avalanche photodiode array.
Energy Technology Data Exchange (ETDEWEB)
Campos, A.; Skinner, C.H.
Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. In the tokamak environment, large particles or fi bers can fall on the electrostatic detector potentially causing a permanent short. An electrostatic dust detector developed in the laboratory is being applied to the National Spherical Torus Experiment (NSTX). We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments at atmospheric pressure with varying nozzle designs, backing pressures, puff durations and exit fl ow orientations have given an optimal confi guration that effectively removes particles from a 25 cm² area. Similar removal effi ciencies were observed under a vacuum base pressure of 1 mTorr. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tri-polar grid of fi ne interdigitated traces has been designed that generates an electrostatic traveling wave for conveying dust particles to a "drain." First trials with only two working electrodes have shown particle motion in optical microscope images.
SLPX: superconducting long-pulse tokamak experiment. [NbTi
Energy Technology Data Exchange (ETDEWEB)
Jassby, D.L.; File, J.; Bronner, G.
The principal objectives of the SLPX (Superconducting Long-Pulse Experiment) are: (1) to demonstrate quasi-steady operation of 3 to 5 MA hydrogen and deuterium tokamak plasmas at high temperature and high thermal wall loading, and (2) to develop reliable operation of prototypical tokamak reactor magnetics systems featuring a toroidal assembly of high-field niobium-tin coils, and a system of pulsed niobium-titanium superconducting poloidal-field coils. This paper describes the status of the engineering design features of the SLPX, with emphasis on the magnetics systems. The toroidal-field coils have an aperture of 3.1 x 4.8 m and can operate with a maximum field at the conductor of 12 T. The superconducting poloidal field magnetics system consists of a pulsed NbTi central solenoid and a set of dc NbTi equilibrium-field coils. The entire machine is enclosed in an outer vacuum container equipped with re-entrant ports that provide ambient access to the room-temperature plasma vessel.
Equilibrium, confinement and stability of runaway electrons in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Spong, D A
Some of the ramifications of the runaway population in tokamak experiments are investigated. Consideration is given both to the normal operating regime of tokamaks where only a small fraction of high energy runaways are present and to the strong runaway regime where runaways are thought to carry a significant portion of the toroidal current. In particular, the areas to be examined are the modeling of strong runaway discharges, single particle orbit characteristics of runaways, macroscopic beam-plasma equilibria, and stability against kink modes. A simple one-dimensional, time-dependent model has been constructed in relation to strong runaway discharges. Single particle orbits are analyzed in relation to both the strong runaway regime and the weak regime. The effects of vector E x vector B drifts are first considered in strong runaway discharges and are found to lead to a slow inward shrinkage of the beam. Macroscopic beam-plasma equilibria are treated assuming a pressureless relativistic beam with inertia and using an ideal MHD approximation for the plasma. The stability of a toroidal relativistic beam against kink perturbations is examined using several models. (MOW)
Alternating current plasma operation in the STOR-M tokamak
Mitarai, O.; Xiao, Chijin; Zhang, Liyan; McColl, D.; Zhang, Wei; Conway, G.; Hirose, A.; Skarsgard, H. M.
One cycle alternating current (AC) plasma operation without a dwell time has been achieved in the STOR-M tokamak with good reproducibility using a newly developed ohmic heating circuit. The plasma current of +24 kA is smoothly ramped down in 10 ms with a rampdown rate of around 2.0 kA/ms and then ramped up to between -20 and -24 kA directly without a dwell time. The plasma density of up to (3.7+or-0.6)*1018 m-3 remains at the current reversal as observed in recent soft landing experiments. The key to a successful, reproducible and direct transition in AC tokamak operations on STOR-M is to control both the total vertical field by a feedback control system and the plasma density by careful gas puffing during the current reversal phase. This experiment has demonstrated that the initial loop voltage for the second negative current is minimized when the dwell time approaches zero, and the AC operation without dwelling is possible whenever the plasma current can be softly terminated with a finite residual plasma density
Recent Results from the STOR-M Tokamak
Hirose, A.; Dreval, M.; Elgriw, S.; Mitarai, O.; Pant, A.; Peng, M.; Rohraff, D.; Singh, A. K.; Trembach, D.; Xiao, C.
This paper reports on two recent experiments carried out on the STOR-M tokamak. The first experiment studied the nature of MHD activities based on singular value decomposition algorithm during the improved confinement phase induced by compact torus injection. The typical MHD modes with mode numbers m = 2, 3, and 4 are suppressed during the improved confinement phase. Shortly before the termination of the improved confinement phase, MHD activities reemerge, starting with a gong-mode-like burst followed by oscillations of a rotating m = 2. The second experiment was successful current start-up with a simulated spherical tokamak configuration where the inner Ohmic heating coils surrounding the iron core are deactivated in STOR-M. Current start-up was also achieved with all the vertical equilibrium field coils deactivated. In the latter case, the vertical equilibrium field was provided solely by the image vertical field produced by the magnetization current in the iron core and compensated for by the current through the feedback control vertical field windings. The observed waveforms agree well with numerical simulations.
Impact of magnetic perturbation fields on tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Fietz, Sina; Maraschek, Marc; Suttrop, Wolfgang; Zohm, Hartmut [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Classen, Ivo [FOM-Institute DIFFER, Nieuwegein (Netherlands); Collaboration: the ASDEX Upgrade Team
Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.
Dynamics and transport of dust particles in tokamak edge plasmas
Energy Technology Data Exchange (ETDEWEB)
Krasheninnikov, S I [University of California, San Diego, La Jolla, CA 92093 (United States); Soboleva, T K [UNAM, Mexico D.F., Mexico and Kurchatov Institute, Moscow (Russian Federation)
We discuss the dust particle dynamics in tokamak edge plasmas, with special emphasis on dust particle transport in the sheath and plasma recycling regions. We demonstrate that being dragged by plasma flows in the vicinity of the material surface, dust particles can be accelerated to speeds of {approx}10{sup 3}-10{sup 4} cm s{sup -1}. The opposite direction of plasma recycling flow as well as the frictional forces at the inner and outer divertor legs, propel the dust particles in opposite toroidal directions depending on their location. The interactions of a dust particle with a corrugated surface or plasma turbulence can cause it to exit the recycling region and fly through the scrape-off layer plasma towards the tokamak core. It is conceivable that dust formation in and transport from the divertor region can play an important role in core plasma contamination. However, even then, the dust particle density around the separatrix is {approx}10{sup -2} cm{sup -3}, which makes it difficult to detect.
Configuration and operation of detritiation systems for ITER Tokamak Complex
Energy Technology Data Exchange (ETDEWEB)
Beloglazov, S., E-mail: firstname.lastname@example.com [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Camp, P. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Hayashi, T. [Japan Atomic Energy Agency, 2-2-2 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan); Lepetit, L.; Perevezentsev, A. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Yamanishi, Y. [Japan Atomic Energy Agency, 2-2-2 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan)
The ventilation systems design for the ITER nuclear buildings ensures radioactive contamination is confined so that workers, the public and the environment are protected. Nuclear buildings are divided into confinement sectors which connect to the Heating, Ventilation and Air Conditioning (HVAC) system and detritiation system (DS). The Tokamak Complex DS provides centralized air purification for the building confinement sectors. A distributed arrangement of ventilation piping provides networks necessary for two key functions, these being Vent Detritiation (VD), to maintain sub-atmospheric pressure, and Air Detritiation (AD) to collect tritium released into the confinement sector. For the VD function, air extracted from the particular confinement sector is directed to the DS for processing prior to exhaust to the environment. This paper presents the configuration of the DS of the Tokamak Complex and addresses details of the design of the distributed piping network. Dynamic flow and pressure drop modelling has been applied to support the development of the system configuration and provide data for sizing the system and selecting components. Further design development is discussed in view of the safety requirements for operation of the system during design basis events such as earthquake or fire.
Nuclear shielding of openings in ITER Tokamak building
Energy Technology Data Exchange (ETDEWEB)
Dammann, A., E-mail: firstname.lastname@example.com [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)
Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.
Control of particles flux in a tokamak with an events structure; Controle des flux de particules dans un Tokamak au moyen d`une structure a events
Energy Technology Data Exchange (ETDEWEB)
Tsitrone, E.
Two key problems in the development of a controlled fusion reactor are: -the control of the ashes resulting from the fusion reaction (helium) and of the impurities coming from the wall erosion, which affect the central plasma performances by diluting the fuel and dissipating a part of the produced energy by radiation. - the removal of the heat carried to the walls by charged particles, which is highly concentrated (peak values of several tens of MW per m{sup 2}). Two types of systems are generally used for the plasma-wall interface: throat limiter and axisymmetric divertor. Neither is an ideal candidate to control simultaneously the heat and particle fluxes. This thesis investigates an alternative configuration, the vented limiter, tested for the first time on the Tore Supra tokamak. The vented limiter principle lies on the recycling neutrals collection by slots, in such a way that local thermal overload is avoided. It is shown experimentally that the surface temperature of the prototype installed in Tore Supra remains uniform. As far as the particle collection is concerned, even though the pressure in the vented limiter is lower than the pressure in the throat limiter by a factor 3 for deuterium and 4 helium, it is sufficient to control the plasma density. Moreover, as with a throat limiter, the pressure exhibits a quadratic evolution with the plasma density. To interpret these results, a model describing the plasma recycling on the limiter and the pumping by the slots has been developed. The model has been validated by a comparison with the experimental data. It was then used to propose an optimized version of the prototype with reshaped slots. This should improve the pumping efficiency by a factor 2, in deuterium as well as in helium, but without removing the discrepancy between both pumping efficiencies. (Abstract Truncated)
Development of 3D ferromagnetic model of tokamak core with strong toroidal asymmetry
DEFF Research Database (Denmark)
Markovič, Tomáš; Gryaznevich, Mikhail; Ďuran, Ivan
Fully 3D model of strongly asymmetric tokamak core, based on boundary integral method approach (i.e. characterization of ferromagnet by its surface) is presented. The model is benchmarked on measurements on tokamak GOLEM, as well as compared to 2D axisymmetric core equivalent for this tokamak......, presented in previous work. Linearized model well describes quantitative characteristics of BR field, generated by poloidal field coils located close to core central column, and distorted by ferromagnet. A discrepancy is seen between linearized form of model for BR field generated by coils under...
Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak; Projeto e analise dos circuitos de producao de campo magnetico toroidal e de formacao do plasma do Tokamak ETE (Experimento Tokamak Esferico)
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Filipe F.P.W.; Bosco, Edson del
This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.
Pellet injection and confinement in the tore supra tokamak; Injection de glacons et confinement dans le tokamak tore supra
Energy Technology Data Exchange (ETDEWEB)
Maget, P
Pellet injection in the centre of tokamak plasmas can lead to an improved confinement regime called PEP (Pellet Enhanced Performance). The present work is dedicated to the mechanisms involved in the PEP regimes obtained in the tokamak Tore Supra. A neoclassical approach of transport shows that it is the anomalous transport, due to plasma turbulence, that causes the enhanced confinement. A linear model describing electrostatic instabilities has been developed in order to study the roles of density profile and current profile during the PEP, in the limit of large growth rates. The effect ofradial shear in flows is taken into account by removing the ExB shear flow rate from the linear growth rate, as suggested by non-linear numerical simulations of turbulence. A local transport coefficient is estimated from the knowledge of the linear growth rate and the mode width. We find that the peaked density profile in PEP regime lowers the diffusion coefficient, and that the velocity shear amplifies this effect. The evolution of the current profile is also stabilizing, but this parameter is not known with sufficient accuracy, so that its role in Tore Supra PEP experiments remains uncertain. (author)
The benefits of transportation investment in Texas.
This report highlights the benefits and return on investment of transportation funding in Texas. In total, over : the next decade, Texans will invest $131 billion in statewide infrastructure with a total economic benefit of : an estimated $373 billio...
Texas RPO workshop implementation project summary.
This report documents rural planning organization (RPO) workshops conducted throughout Texas. An RPO is a voluntary organization created and governed by locally elected officials responsible for transportation decisions at the local level. RPOs addre...
Achieving the Texas Higher Education Vision
National Research Council Canada - National Science Library
Benjamin, Roger
The Texas higher education system faces severe challenges in responding to the twin demands placed on it by economic growth and by the increasing problems of access to higher education that many Texans experience...
Regional tertiary cross sections: Texas Gulf Coast
Energy Technology Data Exchange (ETDEWEB)
Debout, D.G.; Luttrell, P.E.; Seo, J.H.
Regional studies of the Frio Formation along the Texas Gulf Coast were conducted to evaluate potential geothermal energy from deep, geopressured sandstone reservoirs. Published regional cross sections, unpublished cross sections provided by several major oil companies, and extensive micropaleontological and electrical-log files at the Bureau of Economic Geology served as basic data. These sections are meant to show gross regional distribution of sand and shale facies both laterally and vertically throughout the entire Tertiary section along the Texas Gulf Coast.
Making Texas Restaurants Healthier for Children
Centers for Disease Control (CDC) Podcasts
Sylvia Crixell, PhD, RD, Professor of Nutrition at Texas State University, discusses her study which details the success of a community-based program in Texas aimed at combatting childhood obesity by improving childrenâs menus in restaurants. Created: 12/30/2014 by Preventing Chronic Disease (PCD), National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP). Date Released: 12/30/2014.
Posthurricane Survey of Experimental Dunes on Padre Island, Texas.
bonariensis), coast brookweed (Samolus ebracteatus), sand rosegentian (Sabatia arenicola), * and longleaf flaveria (Fiaveria oppositifolia ) occurred...and green carpet weed ( Mollugo verticillata). From 152 to 229 meters away from the bare dunes, vegetation cover * increased to 42 percent, and 25... Mollugo verticellata 0 1 8 0 O enothera drwniondii 0 22 0 0 Poipremum procumbens 0 64 6 15 Phyla incisa 0 0 12 0 Sisyrinchiwn biforme 0 0 1 0 Vernonia
Geodesic acoustic eigenmode for tokamak equilibrium with maximum of local GAM frequency
Energy Technology Data Exchange (ETDEWEB)
Lakhin, V.P. [NRC "Kurchatov Institute", Moscow (Russian Federation); Sorokina, E.A., E-mail: email@example.com [NRC "Kurchatov Institute", Moscow (Russian Federation); Peoples' Friendship University of Russia, Moscow (Russian Federation)
The geodesic acoustic eigenmode for tokamak equilibrium with the maximum of local GAM frequency is found analytically in the frame of MHD model. The analysis is based on the asymptotic matching technique.
Rising Above the Storm: DIG TEXAS
Ellins, K. K.; Miller, K. C.; Bednarz, S. W.; Mosher, S.
For a decade Texas educators, scientists and citizens have shown a commitment to earth science education through planning at the national and state levels, involvement in earth science curriculum and teacher professional development projects, and the creation of a model senior level capstone Earth and Space Science course first offered in 2010 - 2011. The Texas state standards for Earth and Space Science demonstrate a shift to rigorous content, career relevant skills and use of 21st century technology. Earth and Space Science standards also align with the Earth Science, Climate and Ocean Literacy framework documents. In spite of a decade of progress K-12 earth science education in Texas is in crisis. Many school districts do not offer Earth and Space Science, or are using the course as a contingency for students who fail core science subjects. The State Board for Educator Certification eliminated Texas' secondary earth science teacher certification in 2009, following the adoption of the new Earth and Space Science standards. This makes teachers with a composite teacher certification (biology, physics and chemistry) eligible to teach Earth and Space Science, as well other earth science courses (e.g., Aquatic Science, Environmental Systems/Science) even if they lack earth science content knowledge. Teaching materials recently adopted by the State Board of Education do not include Earth and Space Science resources. In July 2011 following significant budget cuts at the 20 Education Service Centers across Texas, the Texas Education Agency eliminated key staff positions in its curriculum division, including science. This "perfect storm" has created a unique opportunity for a university-based approach to confront the crisis in earth science education in Texas which the Diversity and Innovation in the Geosciences (DIG) TEXAS alliance aims to fulfill. Led by the Texas A&M University College of Geosciences and The University of Texas Jackson School of Geosciences, with
Overview of the ITER Tokamak complex building and integration of plant systems toward construction
Energy Technology Data Exchange (ETDEWEB)
Cordier, Jean-Jacques, E-mail: email@example.com [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others
The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed
M.H.D. activity associated with the q=1 surface in the Tore-Supra tokamak; Activite M.H.D. associee a la surface q=1 dans le tokamak Tore-Supra
Energy Technology Data Exchange (ETDEWEB)
Cristofani, P.
In order to increase the temperature, density and confinement time of the plasma energy inside tokamak devices, several heating and fuel injection techniques have been used. However, the increase of the energy content of the central part of the plasma leads to instabilities in the confinement magnetic structure which can degrade the confinement properties and the temperature performances. Inside the plasma, the ``q=1`` surface plays an important role in the confinement process. The aim of this thesis is to study the experimental physics related to this surface with the analysis of the ``saw-tooth`` periodical internal relaxations and of the ``snake`` structure. The first chapter gives a general introduction about thermonuclear fusion and a description of the plasma and of its equilibrium. Chapter 2 is devoted to the description of the soft X-ray tomography, the diagnostic technique used in this work. In chapter 3, a theoretical presentation of plasma stability and a comparison with experimental results obtained in the Tore-Supra tokamak are given. The observations of saw-tooth instabilities are presented with the principal theoretical models which are used to explain this phenomenon. The snake density instability localized in the central part of the plasma is described in chapter 4 with an attempt of interpretation. The equation of the size evolution of a magnetic island was modified to test different models which can explain the snake stability. One model is based on the modification of the bootstrap current induced by the presence of the snake, and on the local modification of the current induced by the accumulation of impurities inside the snake. (J.S.). 107 refs.
Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak.
Purohit, S; Joisa, Y S; Raval, J V; Ghosh, J; Tanna, R; Shukla, B K; Bhatt, S B
Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented.
The regime of the improved confinement with deuterium pellet injected into plasmas of tokamak T-10 with W and Li limiters
Ryzhakov, D. V.; Pavlov, Yu D.; Borschegovskiy, A. A.; Gorshkov, A. V.; Kapralov, V. G.; Klyuchnikov, L. A.; Krylov, S. V.; Malzev, S. G.; Sergeev, D. S.
In this paper, we present the first, after replacing a graphite limiter with a tungsten limiter, experimental results of the regimes of improved plasma confinement in the T-10 tokamak when injecting deuterium pellets. Comparison with the results of previous experiments with a graphite limiter shows the preservation of the improved confinement effect. Preliminary results of the experiments on the change in poloidal angle of injection of pellets allow us to say that with the central injection, the maximum effect of improved confinement is observed.
Emission in the 50-80 A region from highly ionized silver in PLT tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Schwob, J.L.; Wouters, A.; Suckewer, S.; Cohen, S.A.; Finkenthal, M.
The spectrum of silver emitted by Princeton Large Torus (PLT) tokamak plasmas has been recorded in the 25 to 150 A region by a multichannel time-resolving grazing-incidence spectrometer. Silver atoms have been introduced in the tokamak plasma using the laser blow-off technique. For the first time, lines emitted within the 3p-3d transitions of Ag XXIX, Ag XXX, and Ag XXXI ions, between 50 and 80 A, have been identified.
Fractional variational problems and particle in cell gyrokinetic simulations with fuzzy logic approach for tokamaks
Directory of Open Access Journals (Sweden)
Rastović Danilo
Full Text Available In earlier Rastovic's papers [1] and [2], the effort was given to analyze the stochastic control of tokamaks. In this paper, the deterministic control of tokamak turbulence is investigated via fractional variational calculus, particle in cell simulations, and fuzzy logic methods. Fractional integrals can be considered as approximations of integrals on fractals. The turbulent media could be of the fractal structure and the corresponding equations should be changed to include the fractal features of the media.
Transport of fast electrons in lower hybrid current drive plasmas in the HT-7 tokamak
Energy Technology Data Exchange (ETDEWEB)
Chen, Z Y [Institute of Plasma Physics, College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Fang, D; Dai, F; Duan, Z Q; Zhu, J X; Sun, W M [Department of Physics, Yunnan Normal University, Kunming 650092 (China); Wan, B N; Shi, Y J, E-mail: email@example.com [Institute of Plasma Physics, Chinese Academy of Sciences, Hefe 230031 (China)
The transport of fast electrons in lower hybrid current drive (LHCD) plasmas in the HT-7 tokamak was investigated in this work. The evolution of fast electron bremsstrahlung emission profiles after switching off the lower hybrid power was analyzed. We found that the dynamics of the fast electrons is governed by the slowing-down process, and the current density profile can be controlled by LHCD in the HT-7 tokamak.
Development of laser-based technology for the routine first wall diagnostic on the tokamak EAST: LIBS and LIAS
Hu, Z.; Gierse, N.; Li, C.; Liu, P.; Zhao, D.; Sun, L.; Oelmann, J.; Nicolai, D.; Wu, D.; Wu, J.; Mao, H.; Ding, F.; Brezinsek, S.; Liang, Y.; Ding, H.; Luo, G.; Linsmeier, C.; EAST team
A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.
Global two-fluid simulations of geodesic acoustic modes in strongly shaped tight aspect ratio tokamak plasmas
Robinson, J. R.; Hnat, B.; Thyagaraja, A.; McClements, K. G.; Knight, P. J.; Kirk, A.; MAST Team
Following recent observations suggesting the presence of the geodesic acoustic mode (GAM) in ohmically heated discharges in the Mega Amp Spherical Tokamak (MAST) [J. R. Robinson et al., Plasma Phys. Controlled Fusion 54, 105007 (2012)], the behaviour of the GAM is studied numerically using the two fluid, global code CENTORI [P. J. Knight et al. Comput. Phys. Commun. 183, 2346 (2012)]. We examine mode localisation and effects of magnetic geometry, given by aspect ratio, elongation, and safety factor, on the observed frequency of the mode. An excellent agreement between simulations and experimental data is found for simulation plasma parameters matched to those of MAST. Increasing aspect ratio yields good agreement between the GAM frequency found in the simulations and an analytical result obtained for elongated large aspect ratio plasmas.
Global two-fluid simulations of geodesic acoustic modes in strongly shaped tight aspect ratio tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Robinson, J. R.; Hnat, B. [Physics Department, University of Warwick, Coventry, CV4 7AL (United Kingdom); Thyagaraja, A. [H.H. Wills Physics Laboratory, University of Bristol, Bristol BS8 1TL (United Kingdom); McClements, K. G.; Knight, P. J.; Kirk, A. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Collaboration: MAST Team
Following recent observations suggesting the presence of the geodesic acoustic mode (GAM) in ohmically heated discharges in the Mega Amp Spherical Tokamak (MAST) [J. R. Robinson et al., Plasma Phys. Controlled Fusion 54, 105007 (2012)], the behaviour of the GAM is studied numerically using the two fluid, global code CENTORI [P. J. Knight et al. Comput. Phys. Commun. 183, 2346 (2012)]. We examine mode localisation and effects of magnetic geometry, given by aspect ratio, elongation, and safety factor, on the observed frequency of the mode. An excellent agreement between simulations and experimental data is found for simulation plasma parameters matched to those of MAST. Increasing aspect ratio yields good agreement between the GAM frequency found in the simulations and an analytical result obtained for elongated large aspect ratio plasmas.
Time resolved neutron flux diagnostics for quasi-steady-state operation study of the HT-7 tokamak
Zhu, Yubao; Chen, Juequan; Li, Guiming
Time resolved neutron flux diagnostic systems based on BF3 proportional counter and ZnS(Ag) scintillator have been developed and implemented on the HT-7 superconducting tokamak. A ten-channel flexible data acquisition system designed with a PCI-8554 general digital counter and industry PC is equipped. Calibrations are made with several neutron sources. The consistencies of experimental data from two techniques have been proven; the BF3 based system is more reliable with better detection efficiency. The measured neutron yield shows good agreement with the simple numerical calculation. The observed photo-neutron production indicates that photon-nuclear reactions are dominant in several special cases such as low density and disruption conditions. Good agreement on ion temperature deduced from neutron diagnosis and neutral particle analyzer under high parameter plasma conditions implies that neutron flux diagnostics can be used as an effective higher temporal resolution ion temperature monitor.
Unraveling the plasma-material interface with real time diagnosis of dynamic boron conditioning in extreme tokamak plasmas
Domínguez-Gutiérrez, F. Javier; Bedoya, Felipe; Krstić, Predrag S.; Allain, Jean P.; Irle, Stephan; Skinner, Charles H.; Kaita, Robert; Koel, Bruce
We present a study of the role of boron and oxygen in the chemistry of deuterium retention in boronized ATJ graphite irradiated by the extreme environment of a tokamak deuterium plasma. The experimental results were obtained by the first XPS measurements inside the plasma chamber of the National Spherical Torus Experiment Upgrade, between the plasma exposures. The subtle interplay of boron, carbon, oxygen and deuterium chemistry is explained by reactive molecular dynamics simulations, verified by quantum-classical molecular dynamics and successfully compared to the measured data. The calculations deciphered the roles of oxygen and boron for the deuterium retention and predict deuterium uptake into a boronized carbon surface close in value to that previously predicted for a lithiated and oxidized carbon surface.
First time observation of local current shrinkage during the MARFE behavior on the J-TEXT tokamak
Shi, Peng; Zhuang, G.; Gentle, K.; Hu, Qiming; Chen, Jie; Li, Qiang; Liu, Yang; Gao, Li; Zhang, Xiaolong; Liu, Hai; Chen, Zhipeng; Zhu, Lizhi; Li, Fuming; Zhou, Yinan; Zeng, Zhong; Liu, Linzi; He, Jiyang
Multifaceted asymmetric radiation as well as strong poloidal asymmetry of the electron density from the edge, dubbed as 'MARFE', has been observed in high electron density Ohmically heated plasmas on J-TEXT tokamak. Equilibrium reconstruction based on the measured data from the 17-channel FIR polarimeter-interferometer indicates that an asymmetric plasma current density distribution forms at the edge region and the plasma current shrinkage locates at the MARFE affected region. Furthermore, associated with the localized plasma current shrinkage, a locked mode MHD activity is excited, which then terminate the discharge with a major disruption. Localized plasma current shrinkage at the MARFE region is considered to be the direct cause for the density limit disruptions, and the proposed interpretation is consistent with the experimental observations.
Overview of the TCV tokamak program: scientific progress and facility upgrades
Coda, S.; Ahn, J.; Albanese, R.; Alberti, S.; Alessi, E.; Allan, S.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Ariola, M.; Bernert, M.; Beurskens, M.; Bin, W.; Blanchard, P.; Blanken, T. C.; Boedo, J. A.; Bolzonella, T.; Bouquey, F.; Braunmüller, F. H.; Bufferand, H.; Buratti, P.; Calabró, G.; Camenen, Y.; Carnevale, D.; Carpanese, F.; Causa, F.; Cesario, R.; Chapman, I. T.; Chellai, O.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Costea, S.; Crisanti, F.; Cruz, N.; Czarnecka, A.; Decker, J.; De Masi, G.; De Tommasi, G.; Douai, D.; Dunne, M.; Duval, B. P.; Eich, T.; Elmore, S.; Esposito, B.; Faitsch, M.; Fasoli, A.; Fedorczak, N.; Felici, F.; Février, O.; Ficker, O.; Fietz, S.; Fontana, M.; Frassinetti, L.; Furno, I.; Galeani, S.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garrido, I.; Geiger, B.; Giovannozzi, E.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Graves, J. P.; Guirlet, R.; Hakola, A.; Ham, C.; Harrison, J.; Hawke, J.; Hennequin, P.; Hnat, B.; Hogeweij, D.; Hogge, J.-Ph.; Honoré, C.; Hopf, C.; Horáček, J.; Huang, Z.; Igochine, V.; Innocente, P.; Ionita Schrittwieser, C.; Isliker, H.; Jacquier, R.; Jardin, A.; Kamleitner, J.; Karpushov, A.; Keeling, D. L.; Kirneva, N.; Kong, M.; Koubiti, M.; Kovacic, J.; Krämer-Flecken, A.; Krawczyk, N.; Kudlacek, O.; Labit, B.; Lazzaro, E.; Le, H. B.; Lipschultz, B.; Llobet, X.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Maget, P.; Maljaars, E.; Malygin, A.; Maraschek, M.; Marini, C.; Martin, P.; Martin, Y.; Mastrostefano, S.; Maurizio, R.; Mavridis, M.; Mazon, D.; McAdams, R.; McDermott, R.; Merle, A.; Meyer, H.; Militello, F.; Miron, I. G.; Molina Cabrera, P. A.; Moret, J.-M.; Moro, A.; Moulton, D.; Naulin, V.; Nespoli, F.; Nielsen, A. H.; Nocente, M.; Nouailletas, R.; Nowak, S.; Odstrčil, T.; Papp, G.; Papřok, R.; Pau, A.; Pautasso, G.; Pericoli Ridolfini, V.; Piovesan, P.; Piron, C.; Pisokas, T.; Porte, L.; Preynas, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Reich, M.; Reimerdes, H.; Reux, C.; Ricci, P.; Rittich, D.; Riva, F.; Robinson, T.; Saarelma, S.; Saint-Laurent, F.; Sauter, O.; Scannell, R.; Schlatter, Ch.; Schneider, B.; Schneider, P.; Schrittwieser, R.; Sciortino, F.; Sertoli, M.; Sheikh, U.; Sieglin, B.; Silva, M.; Sinha, J.; Sozzi, C.; Spolaore, M.; Stange, T.; Stoltzfus-Dueck, T.; Tamain, P.; Teplukhina, A.; Testa, D.; Theiler, C.; Thornton, A.; Tophøj, L.; Tran, M. Q.; Tsironis, C.; Tsui, C.; Uccello, A.; Vartanian, S.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vijvers, W. A. J.; Vlahos, L.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weisen, H.; Wischmeier, M.; Zestanakis, P.; Zuin, M.; the EUROfusion MST1 Team
The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from
Backstepping Control of the Current Profile in the DIII-D Tokamak
Boyer, M. D.; Barton, J.; Schuster, E.; Walker, M. L.; Humphreys, D. A.
Control of the spatial profile of the plasma current in tokamaks has been demonstrated to be a key condition for advanced scenarios with improved confinement and steady-state operation. Non-model-based controllers tested at DIII-D have shown limitations, motivating the design of model-based controllers that account for the dynamics of the q profile. In this work, we utilize a control-oriented model of the current profile evolution in DIII-D to design a backstepping boundary control law for regulating the current profile around a desired feed-forward trajectory. The control scheme makes use of the total plasma current, total power, and line averaged density as actuators. A simulation study is done to test the control law against uncertainties in the model parameters and initial conditions, as well as input disturbances. Finally, the implementation of the controller in the DIII-D plasma control system is discussed and experimental results are presented. Supported by the NSF CAREER award program ECCS-0645086 and the US DOE under DE-FG02-09ER55064 and DE-FC02-04ER54698.
Simulation of the ELMs triggering by lithium pellet on EAST tokamak using BOUT + +
Wang, Y. M.; Xu, X. Q.; Wang, Z.; Sun, Z.; Hu, J. S.; Gao, X.
A new lithium granule injector (LGI) was developed on EAST. Using the LGI, lithium granules can be efficiently injected into EAST tokamak with the granule radius 0.2-1 mm and the granules velocity 30-110 m/s. ELM pacing was realized during EAST shot #70123 at time window from 4.4-4.7s, the average velocity of the pellet was 75 m/s and the average injection rate is at 99Hz. The BOUT + + 6-field electromagnetic turbulence code has been used to simulate the ELM pacing process. A neutral gas shielding (NGS) model has been implemented during the pellet ablation process. The neutral transport code is used to evaluate the ionized electron and Li ion densities with the charge exchange as a dominant factor in the neutral cloud diffusion process. The snapshot plasma profiles during the pellet ablation and toroidal symmetrization process are used in the 6-field turbulence code to evaluate the impact of the pellets on ELMs. Destabilizing effects of the peeling-ballooning modes are found with lithium pellet injection, which is consistent with the experimental results. A scan of the pellet size, shape and the injection velocity will be conducted, which will benefit the pellet injection design in both the present and future devices. Prepared by LLNL under Contract DE-AC52-07NA27344 and this work is supported by the National Natural Science Fonudation of China (Grant No. 11505221) and China Scholarship Council (Grant No. 201504910132).
Gyrokinetic characterization of the isotope effect in turbulent transport at the FT-2 tokamak
Niskala, P.; Gurchenko, A. D.; Gusakov, E. Z.; Altukhov, A. B.; Esipov, L. A.; Kantor, M. Yu; Kiviniemi, T. P.; Kouprienko, D.; Korpilo, T.; Lashkul, S. I.; Leerink, S.; Perevalov, A. A.; Rochford, R.
Isotope effect allows fusion devices to perform better when heavier hydrogen isotopes are used as fuel, but the reason for this improvement is not yet understood. We present the first direct evidence of the isotope effect on particle confinement in the FT-2 tokamak and investigate it via gyrokinetic simulations. Experimental measurements for comparable hydrogen and deuterium discharges show that the particle confinement time increases by 40% for the heavier isotope species. The isotope effect on particle flux is reproduced in global and local gyrokinetic simulations. Global ELMFIRE simulations demonstrate a systemic reduction in particle fluxes across the radial range, showing a ratio of fluxes {{{Γ }}}{{H}}/{{{Γ }}}{{D}}=1.3 at the edge and {{{Γ }}}{{H}}/{{{Γ }}}{{D}}=1.4 at r/a=0.6. Local GENE simulations agree qualitatively with the result. Besides the fluctuation level, smaller scales and a favorable shift in the cross-phase between the turbulent fluctuations are found to contribute to the isotope effect in the simulations.
Design and application of an EPICS compatible slow plant system controller in J-TEXT tokamak
Energy Technology Data Exchange (ETDEWEB)
Zhang, J.; Zhang, M. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zheng, W., E-mail: email@example.com [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhuang, G.; Ding, T. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)
Highlights: • Underlying functionalities are encapsulated into plug-and-play modules. • The slow controller is EPICS compatible. • The slow controller can work as PSH. - Abstract: J-TEXT tokamak has recently implemented J-TEXT COntrol, Data Access and Communication (CODAC) system on the principle of ITER CODAC. The control network in J-TEXT CODAC system is based on Experimental Physics and Industrial Control System (EPICS). However, former slow plant system controllers in J-TEXT did not support EPICS. Therefore, J-TEXT has designed an EPICS compatible slow controller. And moreover, the slow controller also acts the role of Plant System Host (PSH), which helps non-EPICS controllers to keep working in J-TEXT CODAC system. The basic functionalities dealing with user defined tasks have been modularized into driver or plug-in modules, which are plug-and-play and configured with XML files according to specific control task. In this case, developers are able to implement various kinds of control tasks with these reusable modules, regardless of how the lower-lever functions are implemented, and mainly focusing on control algorithm. And it is possible to develop custom-built modules by themselves. This paper presents design of the slow controller. Some applications of the slow controller have been deployed in J-TEXT, and will be introduced in this paper.
Magnetic flux pumping mechanism prevents sawtoothing in 3D nonlinear MHD simulations of tokamak plasmas
Krebs, Isabel; Jardin, Stephen C.; Guenter, Sibylle; Lackner, Karl; Hoelzl, Matthias; Strumberger, Erika; Ferraro, Nate
3D nonlinear MHD simulations of tokamak plasmas have been performed in toroidal geometry by means of the high-order finite element code M3D-C1. The simulations are set up such that the safety factor on axis (q0) is driven towards values below unity. As reported in and the resulting asymptotic states either exhibit sawtooth-like reconnection cycling or they are sawtooth-free. In the latter cases, a self-regulating magnetic flux pumping mechanism, mainly provided by a saturated quasi-interchange instability via a dynamo effect, redistributes the central current density so that the central safety factor profile is flat and q0 1 . Sawtoothing is prevented if β is sufficiently high to allow for the necessary amount of flux pumping to counterbalance the tendency of the current density profile to centrally peak. We present the results of 3D nonlinear simulations based on specific types of experimental discharges and analyze their asymptotic behavior. A set of cases is presented where aspects of the current ramp-up phase of Hybrid ASDEX Upgrade discharges are mimicked. Another set of simulations is based on low-qedge discharges in DIII-D.
Control of magnetic islands in the STOR-M tokamak using resonant helical fields
Elgriw, S.; Liu, D.; Asai, T.; Hirose, A.; Xiao, C.
The resonant interaction between magnetohydrodynamic (MHD) instability modes and the externally applied helical magnetic field is demonstrated in the Saskatchewan Torus-Modified (STOR-M) tokamak. The study is conducted both numerically and experimentally using a 2D MHD equilibrium code in the former and an (l = 2, n = 1) helical coil carrying a short current pulse in the latter. It is shown numerically that the resonant helical current can efficiently suppress the magnetic islands resonating on the (m = 2, n = 1) magnetic surface when the value of the safety factor at the plasma edge is relatively low (STOR-M during low-q ohmic discharges with high MHD activities. The amplitude and frequency of (2, 1) Mirnov fluctuations are significantly reduced after the activation of the resonant field. Lesser suppression in sideband islands is also observed. Moreover, a phase of improved plasma confinement, characterized by a reduction in Hα emission level, a reduction in loop voltage and an increase in the soft x-ray emission, is induced after application of the resonant field.
ECE and ECH application for investigation of plasma self-organization at T-10 tokamak
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Ploskirev E.G.
Full Text Available A behaviour of the high energy electrons appointed by their ECE simultaneously on two first harmonics are used for the electron distribution function analysis. Experiments were fulfilled in Ohmic regime and with on-axis ECH. The explored spectrum of emission from the extreme column periphery comes into existence after a relaxation of the primary flow of electrons with energy more than 1 MeV on the plasma waves. Maximum energy of the generated electrons does not vary during all discharge time also as the spectrum shape of its emission. The form of spectrum does not depend on electron temperature and density but its character width is directly proportional to the value of magnetic field. The appointed connection of the dynamical features between the peripheral high energy electrons and the periodical kinetic instability in the central plasma area (mode m/n=1/1 confirms an existence of the wave transport from the center to the plasma edge. The set of experimental data corresponds to the theory of the stationary electron distribution function formation by the potential plasma waves which apparently are the main mechanism of plasma self-organization in tokamak.
Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak
Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))
Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.
Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*
Energy Technology Data Exchange (ETDEWEB)
Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja
The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.
Energy Technology Data Exchange (ETDEWEB)
In recent experiments with combined fast wave current drive (FWCD) and deuterium neutral beam injection on the DIII-D tokamak [Luxon and Davis, Fusion Technol. 8, 441 (1985)], an enhanced fusion reactivity and fast ion energy content have been observed in the presence of FWCD, with a concomitant low FWCD efficiency [Petty et al., Radio Frequency Power in Plasmas (AIP, New York, 1997), p. 225]. In this paper, we investigate whether high-harmonic ion cyclotron damping could be responsible for the low FWCD efficiency in these experiments, since a number of high-harmonic hydrogen and deuterium cyclotron resonance layers existed in the plasma. The main analysis tool is the ICRF code PION [Eriksson, Hellsten and Willen, Nucl. Fusion 33, 1037 (1993)], modified to allow multiple frequencies simultaneously as was done in the DIII-D experiments. According to the PION modeling, high harmonic damping of fast wave power can give rise to enhanced fusion reactivity and fast ion energy content, which is consistent with the experimental observations.
Acceleration optimization of real-time equilibrium reconstruction for HL-2A tokamak discharge control
Rui, MA; Fan, XIA; Fei, LING; Jiaxian, LI
Real-time equilibrium reconstruction is crucially important for plasma shape control in the process of tokamak plasma discharge. However, as the reconstruction algorithm is computationally intensive, it is very difficult to improve its accuracy and reduce the computation time, and some optimizations need to be done. This article describes the three most important aspects of this optimization: (1) compiler optimization; (2) some optimization for middle-scale matrix multiplication on the graphic processing unit and an algorithm which can solve the block tri-diagonal linear system efficiently in parallel; (3) a new algorithm to locate the X&O point on the central processing unit. A static test proves the correctness and a dynamic test proves the feasibility of using the new code for real-time reconstruction with 129 × 129 grids; it can complete one iteration around 575 μs for each equilibrium reconstruction. The plasma displacements from real-time equilibrium reconstruction are compared with the experimental measurements, and the calculated results are consistent with the measured ones, which can be used as a reference for the real-time control of HL-2A discharge.
Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor
Energy Technology Data Exchange (ETDEWEB)
D.P. Stotler; C.S. Pitcher; C.J. Boswell; B. LaBombard; J.L. Terry; J.D. Elder; S. Lisgo
A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter.
Adjoint Monte Carlo Simulation of Fusion Product Activation Probe Experiment in ASDEX Upgrade tokamak
Äkäslompolo, Simppa; Tardini, Giovanni; Kurki-Suonio, Taina
The activation probe is a robust tool to measure flux of fusion products from a magnetically confined plasma. A carefully chosen solid sample is exposed to the flux, and the impinging ions transmute the material makig it radioactive. Ultra-low level gamma-ray spectroscopy is used post mortem to measure the activity and, thus, the number of fusion products. This contribution presents the numerical analysis of the first measurement in the ASDEX Upgrade tokamak, which was also the first experiment to measure a single discharge. The ASCOT suite of codes was used to perform adjoint/reverse Monte-Carlo calculations of the fusion products. The analysis facilitated, for the first time, a comparison of numerical and experimental values for absolutely calibrated flux. The results agree to within 40%, which can be considered remarkable considering the fact that all features of the plasma cannot be accounted in the simulations. Also an alternative probe orientation was studied. The results suggest that a better optimized...
Cryopump development of the 5 MW neutral beam injector system on HL-2M tokamak
Yang, Xianfu; Cao, Jianyong; Liu, He; Wei, Huiling; Liu, Weicheng; Ma, Hong
In order to carry out the long pulse and high power neutral beam injector (NBI) heating experiment on the HL-2M tokamak and satisfy its high vacuum pumping speed in the NBI, two built-in cryopumps with large area and based on a three-stage adsorption structure have been developed. The design idea and size of the three stage structure, the manufacturing technique, and the test experiment for cryopumps have been described in this paper. The experimental result shows that the cooldown time of liquid nitrogen pipes and liquid helium pipes are 1.5 h and 10 min, respectively. With activated charcoal bonded on the condensation plates, the pumping speed reaches 1.13 × 106 l/s, and the consumption of liquid helium is 2.53 g/s and 3.11 g/s for the 4-component pump and the 5-component pump, respectively. The conductance probability of the pump inlet is 0.25 for hydrogen.
Pre-ionization by AC Ohmic coil operation in the TST-2 spherical tokamak
Ejiri, A.; Takase, Y.; Tsujii, N.; Togashi, H.; Furui, H.; Shinya, T.; Roidl, B.; Sonehara, M.; Yajima, S.; Yoshida, Y.; Yamazaki, H.; Kitayama, A.; Sato, A.; Takei, Y.; Tajiri, Y.; Matsumoto, N.; Mitarai, O.
AC Ohmic coil operation experiments with frequencies up to 10 kHz were performed on the TST-2 spherical tokamak device, and the pre-ionization process was studied in detail. The minimum loop voltage for pre-ionization was 0.4 V, which corresponds to 0.5 V m-1 at the inboard limiter. Dependences of growth rate and saturation level of the process on various parameters were obtained, and they are compared with a time-dependent 0-dimensional model based on Townsend avalanche and loss along field lines. Most of the dependences are reproduced qualitatively by the model, and quantitative differences are within a factor of several. However, the external vertical field dependence of the appearance time, which is defined as the time to observe a plasma, and the isotope effect cannot be reproduced by the model. An ambipolar diffusion state which is predicted theoretically but mitigated experimentally is discussed. It was found that secondary electron emission at the limiter surfaces is a candidate mechanism to mitigate the state.
GPEC, a real-time capable Tokamak equilibrium code
Rampp, Markus; Fischer, Rainer
A new parallel equilibrium reconstruction code for tokamak plasmas is presented. GPEC allows to compute equilibrium flux distributions sufficiently accurate to derive parameters for plasma control within 1 ms of runtime which enables real-time applications at the ASDEX Upgrade experiment (AUG) and other machines with a control cycle of at least this size. The underlying algorithms are based on the well-established offline-analysis code CLISTE, following the classical concept of iteratively solving the Grad-Shafranov equation and feeding in diagnostic signals from the experiment. The new code adopts a hybrid parallelization scheme for computing the equilibrium flux distribution and extends the fast, shared-memory-parallel Poisson solver which we have described previously by a distributed computation of the individual Poisson problems corresponding to different basis functions. The code is based entirely on open-source software components and runs on standard server hardware and software environments. The real-...
Tokamak fusion reactors with less than full tritium breeding
Energy Technology Data Exchange (ETDEWEB)
Evans, K. Jr.; Gilligan, J.G.; Jung, J.
A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed.
Fast temperature fluctuation measurements in SOL of tokamak TCV
DEFF Research Database (Denmark)
Horacek, J.; Nielsen, Anders Henry; Pitts, R.A.
coupling both across the plasma sheath and in the probe circuit itself. Comparisons are also made between the results from higher frequency sweeping and the standard values derived from a slower sweep to show that the fast measurement is reliable. Considerable effort has been expended in recent years......A fast scanning assembly has been widely used on the TCV tokamak to insert a probe head equipped with an array of single Langmuir probe tips up to the separatrix at the plasma midplane. Using fast voltage sweeping, we obtain IV-characteristics every 8 μs, allowing an estimate of the electron......-characteristics, some effort is required to demonstrate the credibility of the Te derived from the characteristics. Following the methodology proposed in [3], we use both numerical (5spice code) and lab simulations of the equivalent probe circuit, together with a simplified plasma circuit to study the capacitative...
WILDCAT: a catalyzed D-D tokamak reactor
Energy Technology Data Exchange (ETDEWEB)
Evans, K. Jr.; Baker, C.C.; Brooks, J.N.
WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.
First results from gamma ray diagnostics in EAST Tokamak
Energy Technology Data Exchange (ETDEWEB)
Zhou, R. J.; Hu, L. Q.; Zhong, G. Q., E-mail: email@example.com; Cao, H. R.; Liu, G. Z.; Li, K.; Zhang, Y.; Lin, S. Y.; Zhang, J. Z. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)
Gamma ray diagnostics has been developed in the EAST tokamak recently. Six BGO scintillator detectors are arranged on the down-half cross-section and pointed at the up-half cross-section of plasma, with space resolution about 15 cm and energy range from 0.3 MeV to 6 MeV. Three main gamma ray peaks in the energy spectra have been observed and are identified as the results of nuclear reactions {sup 207}Pb(n, n′){sup 207m}Pb, H(n, γ) D, and D(p, γ){sup 3}He, respectively. Upgrading of the system is in progress by using LaBr3(Ce) scintillator, fast photo-multiplier tubes, and a fully digital data acquisition system based on high sample frequency digitizers with digital pulse processing algorithms.
Poloidal magnetics of a divertor compact ignition tokamak
Energy Technology Data Exchange (ETDEWEB)
Strickler, D.J.; Peng, Y.K.M.; Jardin, S.C.
A technique is presented for calculating bounds on the poloidal field (PF) coil currents required to constrain critical plasma shape parameters when plasma pressure and current density profiles are changed. Such considerations are important in the conceptual design of the PF coils for the Compact Ignition Tokamak (CIT) and their electrical power systems in view of the uncertainty in plasma profiles and operating scenarios. Four relatively independent coil groups are sufficient to find a coil current distribution and equilibrium satisfying a prescribed plasma major radius, minor radius, and divertor strike point coordinates. The variation in the coil current distribution with plasma profiles tends to be large for external PF systems and provides a measure by which coil configurations may be compared. 6 refs., 7 figs., 4 tabs.
Neutral beam optimisation for the spherical tokamak ST40
Energy Technology Data Exchange (ETDEWEB)
Salmi, A., E-mail: email@example.com [VTT, P.O. Box 1000, FIN-02044 VTT (Finland); Gryaznevich, M.; Buxton, P.; Nightingale, M. [Tokamak Energy Ltd., Culham Science Centre, Abingdon, Oxon OX143DB (United Kingdom); Tala, T. [VTT, P.O. Box 1000, FIN-02044 VTT (Finland)
Orbit following Monte Carlo code (ASCOT) calculations of neutral beam (NB) induced torque, current and heating have been performed for a selected set of plasma scenarios to optimise the neutral beam injection system for the proposed high magnetic field spherical tokamak ST40. It is found that there are strong variations especially in the current drive efficiency and in the toroidal torque depending on the NB alignment and injection energy. The optimal alignments as well as the optimal injection energy depend both on the scenario and whether the current drive or the toroidal torque is to be maximised. Due to the relatively small calculated current drive efficiency (0.05–0.2 MA/MW) we find that for the studied scenarios a relatively central low field side (LFS) deposition below midplane provides the best overall performance still at a tolerable shine through power.
Transport properties of interacting magnetic islands in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Gianakon, T.A.; Callen, J.D.; Hegna, C.C.
This paper explores the equilibrium and transient transport properties of a mixed magnetic topology model for tokamak equilibria. The magnetic topology is composed of a discrete set of mostly non-overlapping magnetic islands centered on the low-order rational surfaces. Transport across the island regions is fast due to parallel transport along the stochastic magnetic field lines about the separatrix of each island. Transport between island regions is assumed to be slow due to a low residual cross-field transport. In equilibrium, such a model leads to: a nonlinear dependence of the heat flux on the pressure gradient; a power balance diffusion coefficient which increases from core to edge; and profile resiliency. Transiently, such a model also exhibits a heat pulse diffusion coefficient larger than the power balance diffusion coefficient.
STARFIRE: a commercial tokamak fusion power plant study
Energy Technology Data Exchange (ETDEWEB)
STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.
Anomalous Transport in Tokamaks: Examples of Its Suppression
Hirose, A.; Ding, W.; Elia, M.; Xiao, C.; Yamagiwa, M.
Studies conducted towards an understanding of improved plasma confinement in small and large tokamaks are presented. In STOR-M, Ohmic H-modes can be induced by various means. The most recent finding is that injection of a high density compact torus triggers a sustained H-mode. For bias induced H-mode, direct evaluation of fluctuation induced particle flux has been made for Ohmic H-modes. The particle flux is intermittent and the average changes its sign from radially outward in L-mode to radially inward in H-mode. The direction and magnitude of the flux are predominantly determined by the phase difference between the density and potential fluctuations. Nonlinear correlation studies have also been made between two radial positions. In L-mode, the inner region behaves as a master and outer region as slave. In H-mode, the relationship is reversed.
On q dependence of thermal transport in tokamaks
Hirose, A.; Livingstone, S.; Singh, A. K.
Analysis based on a gyro-kinetic ballooning stability code predicts that both the ion and electron thermal diffusivities, due to the ion temperature gradient (ITG) and electron temperature gradient (ETG) modes, respectively, increase with the safety factor q almost linearly. In the case of ITG driven ion thermal diffusivity, the q dependence originates from the coupling to the ion acoustic mode, and in the case of the electron thermal diffusivity due to the ETG mode, it emerges from the coupling to the skin size drift mode. In the ETG mode, charge neutrality does not hold for typical tokamak discharges, and mixing length estimates yield a thermal diffusivity large enough to be relevant to experiments.
A Distributed Synchronization and Timing System on the EAST Tokamak
Luo, Jiarong; Wu, Yichun; Shu, Yantai
A key requirement for the EAST distributed control system (EASTDCS) is time synchronization to an accuracy of AVR microcontroller and the Nut/OS embedded Real Time Operating System (RTOS). The DSTS provides the control and the data acquisition systems with reference clocks (0.01 Hz 10 MHz) and delayed trigger times ( 1 mus 4294 s). These are produced by a Core Module Unit (CMU) connected by optical fibres to many Local Synchronized Node Units (LSNU). The fibres provide immunity from electrical noise and are of equal length to match clock and trigger delays between systems. This paper describes the architecture of the DSTS on the EAST tokamak and provides an overview of the characteristics of the main and local units.
Kinetic effects and reversal flows in the tokamak edge plasma
Energy Technology Data Exchange (ETDEWEB)
Krasheninnikov, S.I.; Soboleva, T.K. (Russian Scientific Center, Kurchatov Inst., Moscow (Russia)); Batischev, O.V.; Zmievskaya, G.I.; Levchenko, V.D.; Ovchenkov, P.A.; Sigov, Yu.S.; Silaev, I.I. (M.V. Keldysh Inst. of Applied Mathematics, Russian Academy of Science, Moscow (Russia))
Preliminary results of the edge tokamak plasma simulation in the frame of 3D kinetic Vlasov-Fokker-Planck equations are presented. Currently two versions of the code are developed: One of them is based on a stochastic modeling method, another uses a finite difference approach via split techniques. In both versions, the self-consistent electric field may be calculated from the quasi-neutrality condition, while the sheath potential is obtained from the ambipolarity of plasma flux towards the neutralizing plate. Simplified equations, describing the dense SOL plasma parameters distributions and allowing the analytical solutions are obtained. It is shown, that for the high particle recycling in SOL, the radial scales of the temperature [Delta][sub T] and densities, [Delta][sub n] are close to each other. The reversal flow is formed within a narrow layer near the separatrix magnetic surface. (orig.).
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas.
Green, D L; Berry, L A; Chen, G; Ryan, P M; Canik, J M; Jaeger, E F
Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (∼kV m(-1)) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.
Divertor research on the DIII-D tokamak
Energy Technology Data Exchange (ETDEWEB)
Hill, D.N.; Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States)] [and others
In this paper the authors summarize recent progress on DIII-D in developing techniques for divertor power and particle control relevant to next generation tokamaks such as the proposed ITER and TPX devices. Density control and helium removal by divertor pumping have been demonstrated for the first time in high confinement ELMing H-mode discharges ({tau} {approximately} 2 {times} {tau}{sub ITER-89P}) following installation of a divertor cryopumping system. The peak divertor heat flux in similar H-mode discharges has been reduced through production of a radiating mantle with neon or argon puffing (reductions of 3--5). A number of diagnostics have been added to improve the understanding of the physical processes involved. They are now designing modified double-null divertor structures for DIII-D that will provide improved particle control for high-triangularity VH-mode plasmas while at the same time allowing for gas puffing to reduce the divertor heat flux.
Equilibrium system analysis in a tokamak ignition experiment. Final report
Energy Technology Data Exchange (ETDEWEB)
Carrera, R.; Weldon, W.F.; Woodson, H.H.
The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.
Equilibrium system analysis in a tokamak ignition experiment
Energy Technology Data Exchange (ETDEWEB)
Carrera, R.; Weldon, W.F.; Woodson, H.H.
The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.
Micro-tearing modes in the Mega Ampere Spherical Tokamak
Applegate, D J; Connor, J W; Cowley, S C; Dorland, W; Hastie, R J; Joiner, N; 10.1088/0741-3335/49/8/001
Recent gyrokinetic stability calculations have revealed that the spherical tokamak is susceptible to tearing parity instabilities with length scales of a few ion Larmor radii perpendicular to the magnetic field lines. Here we investigate this 'micro-tearing' mode in greater detail to uncover its key characteristics, and compare it with existing theoretical models of the phenomenon. This has been accomplished using a full numerical solution of the linear gyrokinetic-Maxwell equations. Importantly, the instability is found to be driven by the free energy in the electron temperature gradient as described in the literature. However, our calculations suggest it is not substantially affected by either of the destabilising mechanisms proposed in previous theoretical models. Instead the instability is destabilised by interactions with magnetic drifts, and the electrostatic potential. Further calculations reveal that the mode is not significantly destabilised by the flux surface shaping or the large trapped particle f...
A general comparison between tokamak and stellarator plasmas
Directory of Open Access Journals (Sweden)
Yuhong Xu
Full Text Available This paper generally compares the essential features between tokamaks and stellarators, based on previous review work individually made by authors on several specific topics, such as theories, bulk plasma transport and edge divertor physics, along with some recent results. It aims at summarizing the main results and conclusions with regard to the advantages and disadvantages in these two types of magnetic fusion devices. The comparison includes basic magnetic configurations, magnetohydrodynamic (MHD instabilities, operational limits and disruptions, neoclassical and turbulent transport, confinement scaling and isotopic effects, plasma rotation, and edge and divertor physics. Finally, a concept of quasi-symmetric stellarators is briefly referred along with a comparison of future application for fusion reactors.
Stability and properties of electron-driven fishbones in tokamaks
Merle, Antoine
In tokamaks, the stability of magneto-hydrodynamic modes can be modified by populations of energetic particles. In ITER-type fusion reactors, such populations can be generated by fusion reactions or auxiliary heating. The electron-driven fishbone mode results from the resonant interaction of the internal kink mode with the slow toroidal precessional motion of energetic electrons and is frequently observed in present-day tokamaks with Electron Cyclotron Resonance Heating or Lower Hybrid Current Drive. In Tore Supra, electron-driven fishbones are observed during LHCD-powered discharges in which a high-energy tail of the electronic distribution function is created. Although the destabilization of those modes is related to the existence of a fast particle population, the modes are observed at a frequency that is lower than expected. The linear stability analysis of electron-driven fishbone modes is the main focus of this thesis. The fishbone dispersion relation is derived in a form that accounts for the contribution of the parallel motion of passing particles to the resonance condition. The MIKE code is developed to compute and solve the dispersion relation of electron-driven fishbones. The code is successfully benchmarked against theory using simple analytical distributions. Using the code MIKE with parametric distributions and equilibria, we show that both barely trapped and barely passing electrons resonate with the mode and can drive it unstable. More deeply trapped and passing electrons have a non-resonant effect on the mode that is, respectively, stabilizing and destabilizing. MIKE simulations using complete ECRH-like distribution functions show that energetic barely passing electrons can contribute to drive a mode unstable at a relatively low frequency. This observation could provide some insight to the understanding of Tore Supra experiments.
Transport in the plateau regime in a tokamak pedestal
Seol, J.; Shaing, K. C.
In a tokamak H-mode, a strong E × B flow shear is generated during the L-H transition. Turbulence in a pedestal is suppressed significantly by this E × B flow shear. In this case, neoclassical transport may become important. The neoclassical fluxes are calculated in the plateau regime with the parallel plasma flow using their kinetic definitions. In an axisymmetric tokamak, the neoclassical particles fluxes can be decomposed into the banana-plateau flux and the Pfirsch-Schlüter flux. The banana-plateau particle flux is driven by the parallel viscous force and the Pfirsch-Schlüter flux by the poloidal variation of the friction force. The combined quantity of the radial electric field and the parallel flow is determined by the flux surface averaged parallel momentum balance equation rather than requiring the ambipolarity of the total particle fluxes. In this process, the Pfirsch-Schlüter flux does not appear in the flux surface averaged parallel momentum equation. Only the banana-plateau flux is used to determine the parallel flow in the form of the flux surface averaged parallel viscosity. The heat flux, obtained using the solution of the parallel momentum balance equation, decreases exponentially in the presence of sonic Mp without any enhancement over that in the standard neoclassical theory. Here, Mp is a combination of the poloidal E × B flow and the parallel mass flow. The neoclassical bootstrap current in the plateau regime is presented. It indicates that the neoclassical bootstrap current also is related only to the banana-plateau fluxes. Finally, transport fluxes are calculated when Mp is large enough to make the parallel electron viscosity comparable with the parallel ion viscosity. It is found that the bootstrap current has a finite value regardless of the magnitude of Mp.
Bulk ion heating with ICRF waves in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Mantsinen, M. J., E-mail: email@example.com [Catalan Institution for Research and Advanced Studies, Barcelona (Spain); Barcelona Supercomputing Center, Barcelona (Spain); Bilato, R.; Bobkov, V. V.; Kappatou, A.; McDermott, R. M.; Odstrčil, T.; Tardini, G.; Bernert, M.; Dux, R.; Maraschek, M.; Noterdaeme, J.-M.; Ryter, F.; Stober, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Nocente, M. [Dipartimento di Fisica "G. Occhialini", Università degli Studi di Milano-Bicocca, Milano (Italy); Istituto di Fisica del Plasma "P. Caldirola", CNR, Milano (Italy); Hellsten, T. [Dept. of Fusion Plasma Physics, EES, KTH, Stockholm (Sweden); Mantica, P.; Tardocchi, M. [Istituto di Fisica del Plasma "P. Caldirola", CNR, Milano (Italy); Nielsen, S. K.; Rasmussen, J.; Stejner, M. [Technical University of Denmark, Department of Physics, Lyngby (Denmark); and others
Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER and DEMO operation. This is of particular importance for the bulk ion heating capabilities of ICRF waves. Efficient bulk ion heating with the standard ITER ICRF scheme, i.e. the second harmonic heating of tritium with or without {sup 3}He minority, was demonstrated in experiments carried out in deuterium-tritium plasmas on JET and TFTR and is confirmed by ICRF modelling. This paper focuses on recent experiments with {sup 3}He minority heating for bulk ion heating on the ASDEX Upgrade (AUG) tokamak with ITER-relevant all-tungsten PFCs. An increase of 80% in the central ion temperature T{sub i} from 3 to 5.5 keV was achieved when 3 MW of ICRF power tuned to the central {sup 3}He ion cyclotron resonance was added to 4.5 MW of deuterium NBI. The radial gradient of the T{sub i} profile reached locally values up to about 50 keV/m and the normalized logarithmic ion temperature gradients R/LT{sub i} of about 20, which are unusually large for AUG plasmas. The large changes in the T{sub i} profiles were accompanied by significant changes in measured plasma toroidal rotation, plasma impurity profiles and MHD activity, which indicate concomitant changes in plasma properties with the application of ICRF waves. When the {sup 3}He concentration was increased above the optimum range for bulk ion heating, a weaker peaking of the ion temperature profile was observed, in line with theoretical expectations.
Electron cyclotron resonance heating in the microwave tokamak experiment
Energy Technology Data Exchange (ETDEWEB)
Allen, S.L.; Casper, T.A.; Fenstermacher, M.E. [and others
This paper presents the results from a series of Electron Cyclotron Resonance Heating (ECRH) experiments on the Microwave Tokamak Experiment (MTX). On-axis heating at B{sub T} = 5T (f{sub ce} = 140 GHz) has been performed at electron densities up to cutoff. We have used both a long-pulse gryotron ({approximately}200 kW, {approximately}0.1s) and a pulsed Free Electron Laser (FEL) as microwave sources. Gyrotron experiments with power densities corresponding to 4 MW m{sup {minus}3}. A far infrared (FIR) polarimeter measured peaking of plasma current profiles in some discharges during the ECRH pulse. During high-power single-pulse FEL experiments, single-pass microwave !transmission measurements show nonlinear effects; i.e., higher transmission than predicted by linear theory. A corrugated-wall duct was used in the tokamak port to increase the gradient of the parallel refractive index n{sub parallel} of the incident wave, and increased absorption was observed. Evidence of electron tail heating during FEL pulses was observed on soft x-ray and ECE diagnostics. These results are in agreement with predictions of nonlinear theory; extrapolation of this theory to reactor-like conditions indicates efficient absorption and heating. A Laser Assisted Particle Probe Spectroscopy (LAPPS) diagnostic provided estimates of the vacuum electric field of the FEL which were consistent with the measured power. Multiple pulse operation of the ETA-II accelerator for the FEL has also been demonstrated, indicating the feasibility of high-average power FEL operation.
Continuous, saturation, and discontinuous tokamak plasma vertical position control systems
Energy Technology Data Exchange (ETDEWEB)
Mitrishkin, Yuri V., E-mail: firstname.lastname@example.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: firstname.lastname@example.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: firstname.lastname@example.com [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: firstname.lastname@example.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)
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.
First-principles-driven model-based current profile control for the DIII-D tokamak via LQI optimal control
Boyer, Mark D.; Barton, Justin; Schuster, Eugenio; Luce, Tim C.; Ferron, John R.; Walker, Michael L.; Humphreys, David A.; Penaflor, Ben G.; Johnson, Robert D.
In tokamak fusion plasmas, control of the spatial distribution profile of the toroidal plasma current plays an important role in realizing certain advanced operating scenarios. These scenarios, characterized by improved confinement, magnetohydrodynamic stability, and a high fraction of non-inductively driven plasma current, could enable steady-state reactor operation with high fusion gain. Current profile control experiments at the DIII-D tokamak focus on using a combination of feedforward and feedback control to achieve a targeted current profile during the ramp-up and early flat-top phases of the shot and then to actively maintain this profile during the rest of the discharge. The dynamic evolution of the current profile is nonlinearly coupled with several plasma parameters, motivating the design of model-based control algorithms that can exploit knowledge of the system to achieve desired performance. In this work, we use a first-principles-driven, control-oriented model of the current profile evolution in low confinement mode (L-mode) discharges in DIII-D to design a feedback control law for regulating the profile around a desired trajectory. The model combines the magnetic diffusion equations with empirical correlations for the electron temperature, resistivity, and non-inductive current drive. To improve tracking performance of the system, a nonlinear input transformation is combined with a linear-quadratic-integral (LQI) optimal controller designed to minimize a weighted combination of the tracking error and controller effort. The resulting control law utilizes the total plasma current, total external heating power, and line averaged plasma density as actuators. A simulation study was used to test the controller's performance and ensure correct implementation in the DIII-D plasma control system prior to experimental testing. Experimental results are presented that show the first-principles-driven model-based control scheme's successful rejection of input
Core-ion temperature measurement of the ADITYA tokamak using passive charge exchange neutral particle energy analyzer
Energy Technology Data Exchange (ETDEWEB)
Pandya, Santosh P.; Ajay, Kumar; Mishra, Priyanka; Dhingra, Rajani D.; Govindarajan, J. [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India)
Core-ion temperature measurements have been carried out by the energy analysis of passive charge exchange (CX) neutrals escaping out of the ADITYA tokamak plasma (minor radius, a= 25 cm and major radius, R= 75 cm) using a 45 Degree-Sign parallel plate electrostatic energy analyzer. The neutral particle analyzer (NPA) uses a gas cell configuration for re-ionizing the CX-neutrals and channel electron multipliers (CEMs) as detectors. Energy calibration of the NPA has been carried out using ion-source and {Delta}E/E of high-energy channel has been found to be {approx}10%. Low signal to noise ratio (SNR) due to VUV reflections on the CEMs was identified during the operation of the NPA with ADITYA plasma discharges. This problem was rectified by upgrading the system by incorporating the additional components and arrangements to suppress VUV radiations and improve its VUV rejection capabilities. The noise rejection capability of the NPA was experimentally confirmed using a standard UV-source and also during the plasma discharges to get an adequate SNR (>30) at the energy channels. Core-ion temperature T{sub i}(0) during flattop of the plasma current has been measured to be up to 150 eV during ohmically heated plasma discharges which is nearly 40% of the average core-electron temperature (typically T{sub e}(0) {approx} 400 eV). The present paper describes the principle of tokamak ion temperature measurement, NPA's design, development, and calibration along with the modifications carried out for minimizing the interference of plasma radiations in the CX-spectrum. Performance of the NPA during plasma discharges and experimental results on the measurement of ion-temperature have also been reported here.
Interaction between fast ions and ion cyclotron heating in a tokamak plasma; Interaction des ions rapides avec les ondes a la frequence cyclotronique ionique dans un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bergeaud, V
In an ignited fusion reactor, the plasma temperature is sustained by the fusion reactions. However, before this regime is reached, it is necessary to bring an additional power to the plasma. One of the methods that enables the coupling of power is the use of an electromagnetic wave in the ion cyclotron range of frequencies (ICRF). This thesis deals with the interaction between ICRF heating and the fast ions. The thesis contains a theoretical study of the influence of ICRF heating on the ion distribution function. A particular emphasis is put on the importance of the toroidal spectrum of the modes of propagation of the wave in the tokamak. It is necessary to take into account all these modes in order to correctly assess the strength of the wave particle interaction, especially for high energy particles (of the order of hundreds of keV). The classical treatment of the wave particle interaction is based on the hypothesis that the cyclotron phase of the particle and the wave phase are de-correlated between successive resonant interactions. One is therefore led to consider ICRF heating as a diffusive process. This hypothesis is reconsidered in this thesis and it is shown that strong correlations exist in a large part of the velocity space. For this study, a numerical code that computes the full trajectory of particles interacting with a complete electromagnetic field has been developed. The thesis also deals with the problem of fast ion losses due to the breaking of the toroidal symmetry of the confinement magnetic field (called the ripple modulation). Between two toroidal coils, local magnetic wells exist, and particles can be trapped there. When trapped they undergo a vertical drift that makes them quit the plasma rapidly. The ripple modulation also causes an enhancement of the radial diffusion, thereby increasing the losses. A Monte Carlo model describing these mechanisms is presented. This model is validated thanks to a comparison with an experimental database from
Interaction of fast ions with ion cyclotron electromagnetic waves in tokamak plasma; Interaction des ions rapides avec les ondes a la frequence cyclotronique ionique dans un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bergeaud, V
In an ignited fusion reactor, the plasma temperature is sustained by the fusion reactions. However, before this regime is reached, it is necessary to bring an additional power to the plasma. One of the methods that enables the coupling of power is the use of an electromagnetic wave in the ion cyclotron range of frequencies (ICRF). This thesis deals with the interaction between ICRF heating and the fast ions. The thesis contains a theoretical study of the influence of ICRF heating on the ion distribution function. A particular emphasis is put on the importance of the toroidal spectrum of the modes of propagation of the wave in the tokamak. It is necessary to take into account all these modes in order to correctly assess the strength of the wave particle interaction, especially for high energy particles (of the order of hundreds of keV). The classical treatment of the wave particle interaction is based on the hypothesis that the cyclotron phase of the particle and the wave phase are de-correlated between successive resonant interactions. One is therefore led to consider ICRF heating as a diffusive process. This hypothesis is reconsidered in this thesis and it is shown that strong correlations exist in a large part of the velocity space. For this study, a numerical code that computes the full trajectory of particles interacting with a complete electromagnetic field has been developed. The thesis also deals with the problem of fast ion losses due to the breaking of the toroidal symmetry of the confinement magnetic field (called the ripple modulation). Between two toroidal coils, local magnetic wells exist, and particles can be trapped there. When trapped they undergo a vertical drift that makes them quit the plasma rapidly. The ripple modulation also causes an enhancement of the radial diffusion, thereby increasing the losses. A Monte Carlo model describing these mechanisms is presented. This model is validated thanks to a comparison with an experimental database from
Texas site selection and licensing status
Energy Technology Data Exchange (ETDEWEB)
Avant, R.V. Jr.
Texas has identified a potential site in Hudspeth County in far West Texas near the town of Fort Hancock. Over the past year the Texas Low-Level Radioactive Waste Disposal Authority has been conducting detailed geology, hydrology, meteorology, soils, and flora and fauna evaluations. An authorization by the Board of Directors of the Authority to proceed with a license application, assuming that the detailed evaluation indicates that the site is suitable, is expected by September. A prototype license has been prepared in anticipation of the order to proceed with licensing, and the formal license application is expected to be submitted to the Texas Department of Health-Bureau of Radiation Control in December, meeting the license application milestone. Although site selection processes in all siting areas across the country have experienced organized opposition, El Paso County has funded a particularly well-organized, well-financed program to legally and technically stop consideration of the Fort Hancock site prior to the licensing process. Many procedural, regulatory, and technical issues have been raised which have required responses from the Authority in order to proceed with licensing. This has provided a unique perspective of what to expect from well-organized opposition at the licensing stage. This paper presents an update on the Texas siting activity with detailed information on the site evaluation and license application. Experience of dealing with issues raised by opposition relating to NRC guidelines and rules is also discussed.
Commercial truck platooning demonstration in Texas – level 2 automation.
Through this project, the Texas Department of Transportation (TxDOT) funded the creation of a comprehensive truck platooning demonstration in Texas, serving as a proactive effort in assessing innovative operational strategies to position TxDOT as a l...
Preliminary Analysis of Ex-Vessel Steam Explosion using TEXAS-V code for APR1400
Energy Technology Data Exchange (ETDEWEB)
Song, Sung Chu; Lee, Jung Jae; Cho, Yong Jin; Hwang, Taesuk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
The purpose of this study is to explore input development and the audit calculation using TEXAS-V code for ex-vessel steam explosion for a flooded reactor cavity of APR1400. TEXAS computational models are one of the simplified tools for simulations of fuel-coolant interaction during mixing, triggering and explosion phase. The models of TEXAS code were validated by performing the fundamental experimental investigation in the KROTOS facility at JRC, Ispra. The experiments such as KROTOS and FARO experiment are aimed at providing benchmark data to examine the effect of fuel-coolant initial conditions and mixing on explosion energetics with alumina and prototypical core material. TEXAS-V code used in this study was to analyze and predict the ex-vessel steam explosion for a reactor scale. The input deck to simulate the flooded reactor cavity of APR1400 is developed and base case calculation is performed. This study will provide a base for further study. The code will be of use for the evaluation and sensitivity study of ex-vessel steam explosion for ERVC strategy in the future studies. Analysis result of this study is similar to the result of other study. Through this study, it is found that TEXAS-V could be the used as a tool for predicting the steam explosion load on a reactor scale, as fast running computer code. In addition, TEXAS-V code could be to evaluate the impact of various uncertainties, which are not clearly understood yet, to provide a conservative envelope for the steam explosion.
Analysis of toroidal momentum dissipation by non-axisymmetric fields in high beta, low aspect ratio tokamak experiments
Zhu, Wubiao
Sustained passive stabilization of ideal MHD modes in tokamaks and the spherical torus (ST) can be obtained by maintaining high plasma rotation. However, the rotation has been theoretically predicted and experimentally found to decay, eliminating passive stabilization and impeding sustainment of high beta. Understanding the physical mechanisms leading to plasma momentum dissipation is extremely important to determine how the favorable plasma rotation can be sustained and maximized and how the plasma rotation profile can be controlled in the future tokamaks. The present work first quantitatively examines the agreement between electromagnetic torque theory and localized resonant plasma rotation damping by resistive MHD instabilities. The drag caused by the interaction of the tearing mode with the wall eddy currents can quantitatively explain localized resonant plasma toroidal rotation damping induced by the tearing mode. The remainder of the study focuses on quantitative comparison of theory to the observed global plasma rotation damping by applied non-axisymmetric fields and ideal MHD instabilities. Plasmas with beta below, approaching, and above the calculated no-wall beta limit are created to study the non-resonant plasma toroidal rotation damping physics. At low beta, external applied field perturbations are used to study the braking effects of n = 1 and n = 3 field configurations. At beta close to the no-wall limit, resonant field amplification (RFA)/stabilized RWM effects are added to the model in computing the braking magnetic field. At beta well above the no-wall limit, the unstable RWM damps the plasma rotation strongly, and the theoretically computed mode eigenfunction is used to determine the field. An NBI source term, resonant EM torque, fluid viscous force and neoclassical toroidal viscosity (NTV) torque in both plateau and collisionless 1/nu are included in the model. Inclusion of a broad toroidal and poloidal field spectrum is required for quantitative
Physics of the interaction between runaway electrons and the background plasma of the current quench in tokamak disruptions
Reux, Cedric
Runaway electrons are created during disruptions of tokamak plasmas. They can be accelerated in the form of a multi-MA beam at energies up to several 10's of MeV. Prevention or suppression of runaway electrons during disruptions will be essential to ensure a reliable operation of future tokamaks such as ITER. Recent experiments showed that the suppression of an already accelerated beam with massive gas injection was unsuccessful at JET, conversely to smaller tokamaks. This was attributed to a dense, cold background plasma (up to several 1020 m-3 accompanying the runaway beam. The present contribution reports on the latest experimental results obtained at JET showing that some mitigation efficiency can be restored by changing the features of the background plasma. The density, temperature, position of the plasma and the energy of runaways were characterized using a combined analysis of interferometry, soft X-rays, bolometry, magnetics and hard X-rays. It showed that lower density background plasmas were obtained using smaller amounts of gas to trigger the disruption, leading to an improved penetration of the mitigation gas. Based on the observations, a physical model of the creation of the background plasma and its subsequent evolution is proposed. The plasma characteristics during later stages of the disruption are indeed dependent on the way it was initially created. The sustainment of the plasma during the runaway beam phase is then addressed by making a power balance between ohmic heating, power transfer from runaway electrons, radiation and atomic processes. Finally, a model of the interaction of the plasma with the mitigation gas is proposed to explain why massive gas injection of runaway beams works only in specific situations. This aims at pointing out which parameters bear the most importance if this mitigation scheme is to be used on larger devices like ITER. Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium
Teachers and Teaching Conditions in Rural Texas: Policy Brief
Jimerson, Lorna
Over four milliion children go to public schools in Texas; of these, almost half a million (474,000) students attend school in rural areas. Thirty-six percent of rural Texas students are members of a minority group, 46% are poor, and more than 31,000 students in rural Texas do not speak English well. These are Texas-style large numbers that begin…
Electron heating via mode converted ion Bernstein waves in the Alcator C-Mod tokamak
Energy Technology Data Exchange (ETDEWEB)
Bonoli, P.T.; OShea, P.; Brambilla, M.; Golovato, S.N.; Hubbard, A.E.; Porkolab, M.; Takase, Y.; Boivin, R.L.; Bombarda, F.; Christensen, C.; Fiore, C.L.; Garnier, D.; Goetz, J.; Granetz, R.; Greenwald, M.; Horne, S.F.; Hutchinson, I.H.; Irby, J.; Jablonski, D.; LaBombard, B.; Lipschultz, B.; Marmar, E.; May, M.; Mazurenko, A.; McCracken, G.; Nachtrieb, R.; Niemczewski, A.; Ohkawa, H.; Pappas, D.A.; Reardon, J.; Rice, J.; Rost, C.; Schachter, J.; Snipes, J.A.; Stek, P.; Takase, K.; Terry, J.; Wang, Y.; Watterson, R.L.; Welch, B.; Wolfe, S.M. [Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)
Highly localized direct electron heating [full width at half-maximum (FWHM){congruent}0.2a] via mode converted ion Bernstein waves has been observed in the Alcator C-Mod Tokamak [I. H. Hutchinson {ital et al.}, Phys. Plasmas {bold 1}, 1511 (1994)]. Electron heating at or near the plasma center (r/a{ge}0.3) has been observed in H({sup 3}He) discharges at B{sub 0}=(6.0{endash}6.5)T and n{sub e}(0){congruent}1.8{times}10{sup 20}m{sup {minus}3}. [Here, the minority ion species is indicated parenthetically.] Off-axis heating (r/a{ge}0.5) has also been observed in D({sup 3}He) plasmas at B{sub 0}=7.9T. The concentration of {sup 3}He in these experiments was in the range of n{sub 3{sub He}}/n{sub e}{congruent}(0.2{endash}0.3) and the locations of the mode conversion layer and electron heating peak could be controlled by changing the {sup 3}He concentration or toroidal magnetic field (B{sub 0}). The electron heating profiles were deduced using a rf modulation technique. Detailed comparisons with one-dimensional and toroidal full-wave models in the ion cyclotron range of frequencies have been carried out. One-dimensional full-wave code predictions were found to be in qualitative agreement with the experimental results. Toroidal full-wave calculations indicated the importance of volumetric and wave focusing effects in the interpretation of the experimental results. {copyright} {ital 1997 American Institute of Physics.}
Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator
Energy Technology Data Exchange (ETDEWEB)
Ahn, J-W; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.
Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for
Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator
Ahn, J.-W.; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.
Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for