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Discharges with high bootstrap current fraction on Tore Supra

Description: Bootstrap current is regarded as a serious candidate for non-inductively driving a significant fraction of the total current. High bootstrap fraction discharges have already been achieved and analysed in several tokamaks, including JT-60, DIII-D and TFTR. Tore Supra (R=2. 36 m, a=0.80 m) is particularly suited for the study of non-inductive discharges and long pulse operation. It is equipped with several of non-inductive current drive/heating systems including Lower Hybrid Current Drive (LHCD), Fast Wave Electron Heating (FWEH), and in the future Electron Cyclotron Heating. Fully non-inductive discharges with enhanced confinement (LHEP mode) have already been obtained in Tore Supra with LHCD. High {Beta}p ({le}1.6) regimes current nave also been achieved in the presence of FWEH. In particular, a discharge with 70% of the total current generated by the bootstrap current was observed. In this context, non-inductive current density profile determination is essential for understanding current drive experiments and ultimately for implementing current profile control. This paper briefly describes two methods developed on Tore Supra to determine the non-inductive current density profiles. The agreement between the two methods has been tested by applying them to ohmic discharges. These methods are then applied to the high bootstrap fraction discharges heated by FWEH. The non-inductive current density profile of these discharges are carried out. and the results are finally compared to several models of bootstrap current including Hirsman`s with low collisionality, matrix formulation and both Kessel and Houlberg matrix formulation.
Date: December 31, 1995
Creator: Joffrin, E.; Saoutic, B.; Basiuk, V.; Forest, C. & Houlberg, W.A.
Partner: UNT Libraries Government Documents Department

Electron cyclotron heating and current drive approach for low-temperature startup plasmas using O-X-EBW mode conversion

Description: A mechanism for heating and driving currents in very overdense plasmas is considered based on a double-mode conversion: Ordinary mode to Extraordinary mode to electron Bernstein wave. The possibility of using this mechanism for plasma buildup and current ramp in the National Spherical Torus Experiment is investigated.
Date: June 1, 1997
Creator: Batchelor, D.B. & Bigelow, T.S.
Partner: UNT Libraries Government Documents Department

Fast wave antenna array feed circuits tolerant of time-varying loading for DIII-D

Description: Three different transmission line configurations for operating a four-element antenna array with a single rf power source are compared. The goal of this study is to obtain a system that presents a matched load to the generator despite variation of the loading of the array elements due to changing plasma conditions.
Date: April 1, 1997
Creator: Pinsker, R.I.; Moeller, C.P. & Phelps, D.A.
Partner: UNT Libraries Government Documents Department

Fast ion orbits in spherical tokamaks

Description: In a spherical tokamak, the 1/R variation of the toroidal field is extreme, and for a given value of the safety factor a relatively low average toroidal field can be used, together with large plasma current and large plasma minor radius and elongation. The poloidal and toroidal fields are then of similar size. In consequence, the orbits of fast ions depart considerably from the guiding center orbits because of gyromotion in the small magnetic fields in the low field side.
Date: July 20, 1995
Creator: Solano, E.R.
Partner: UNT Libraries Government Documents Department

Results of DIII-D operation with new enabling technologies

Description: Recent experiments on DIII-D have been carried out to understand and explore optimized tokamak operating modes by exploiting control of the plasma current and pressure profiles using new RF current drive and divertor technology. DIII-D emphasizes plasma shape and divertor experiments using a digital plasma control system and extensive diagnostics to develop improved understanding and control of transport barriers in high performance plasmas. The emphasis of the program is to extend the duration of high performance operating modes beyond the plasma current relaxation time by using ICRF and ECH current drive. Engineering features of the new RF systems being developed for these experiments as well as new divertor results are described. DIII-D employs multi-element ICRF antennas for fast-wave electron heating and on-axis current drive and is beginning 110 GHz ECH experiments with MW-level gyrotrons for off-axis current drive. DIII-D employs active cryogenic divertor neutral particle pumping for plasma density and plasma pressure profile control. A divertor modification is now being implemented on DIII-D to pump higher triangularity plasmas and to better baffle neutral backflow from the recycling divertor region.
Date: March 1, 1997
Creator: Simonen, T.C.
Partner: UNT Libraries Government Documents Department

ICRF heating and current drive experiments on TFTR

Description: Recent experiments in the Ion Cyclotron Range of Frequencies (ICRF) at TFTR have focused on the RF physics relevant to advanced tokamak D-T reactors. Experiments performed either tested confinement in reactor relevant plasmas or tested specific ICRF heating scenarios under consideration for reactors. H-minority heating was used to supply identical heating sources for matched D-T and D only L-mode plasmas to determine the species scaling for energy confinement. Second harmonic tritium heating was performed with only thermal tritium ions in an L-mode target plasma, verifying a possible start-up scenario for the International Thermonuclear Experimental Reactor (ITER). Direct electron heating in Enhanced Reverse Shear (ERS) plasmas has been found to delay the back transition out of the ERS state. D-T mode conversion of the fast magnetosonic wave to an Ion Berstein Wave (IBW) for off-axis heating and current drive has been successfully demonstrated for the first time. Parasitic Li{sup 7} cyclotron damping limited the fraction of the power going to the electrons to less than 30%. Similar parasitic damping by Be{sup 9} could be problematic in ITER. Doppler shifted fundamental resonance heating of beam ions and alpha particles has also been observed.
Date: December 31, 1996
Creator: Rogers, J.H.; Hosea, J.C. & Phillips, C.K.
Partner: UNT Libraries Government Documents Department

The ion cyclotron system for the KSTAR tokamak

Description: The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (Ro = 1.8 m, a = 0.5 m, {kappa} {lt}= 2, BT{lt}= 3.5 T, Ip {lt}= 2 MA, {tau} pulse {lt}= 300 s) is being constructed to do long pulse, high beta, advanced operating mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of rf power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast wave current drive (FWCD), and mode conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current drive efficiency. This paper describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher; IC physics requirements and analyses are covered in a companion paper.
Date: January 1, 1998
Creator: Swain, D.W.; Ryan, P.W.; Taylor, D.J.; Hong, B.G.; Bae, Y.D. & Hwang, C.K.
Partner: UNT Libraries Government Documents Department

The technology of ICRF systems

Description: The technology of ICRF systems has made substantial gains in the past few years. The total power levels have increased from the 5-MW level to greater than 20 MW; the pulse lengths have increased from the 100-msec level to 30 seconds. Recently, fast wave current drive (FWCD) has been demonstrated using phased arrays of ICRF antennas. In order to achieve such large gains, substantial changes and improvements were needed in the level of design analysis, fabrication techniques, and system controls.
Date: December 1995
Creator: Baity, F. W.; Barber, G. C. & Bigelow, T. S.
Partner: UNT Libraries Government Documents Department

Fast wave current drive in DIII-D

Description: The non-inductive current drive from fast Alfven waves launched by a directional four-element antenna was measured in the DIII-D tokamak. The fast wave frequency (60 MHz) was eight times the deuterium cyclotron frequency at the plasma center. An array of rf pickup loops at several locations around the torus was used to verify the directivity of the four-element antenna. Complete non-inductive current drive was achieved using a combination of fast wave current drive (FWCD) and electron cyclotron current drive (ECCD) in discharges for which the total plasma current was inductively ramped down from 400 to 170 kA. For discharges with steady plasma current, up to 110 kA of FWCD was inferred from an analysis of the loop voltage, with a maximum non-inductive current (FWCD, ECCD, and bootstrap) of 195 out of 310 kA. The FWCD efficiency increased linearly with central electron temperature. For low current discharges, the FWCD efficiency was degraded due to incomplete fast wave damping. The experimental FWCD was found to agree with predictions from the CURRAY ray-tracing code only when a parasitic loss of 4% per pass was included in the modeling along with multiple pass damping.
Date: February 1, 1995
Creator: Petty, C.C.; Callis, R.W.; Chiu, S.C.; deGrassie, J.S.; Forest, C.B.; Freeman, R.L. et al.
Partner: UNT Libraries Government Documents Department


Description: Green's-function techniques are used to calculate electron cyclotron current drive (ECCD) efficiency in general tokamak geometry in the low-collisionality regime. Fully relativistic electron dynamics is employed in the theoretical formulation. The high-velocity collision model is used to model Coulomb collisions and a simplified quasi-linear rf diffusion operator describes wave-particle interactions. The approximate analytic solutions which are benchmarked with a widely used ECCD model, facilitate time-dependent simulations of tokamak operational scenarios using the non-inductive current drive of electron cyclotron waves.
Date: March 1, 2003
Partner: UNT Libraries Government Documents Department

First results on fast wave current drive in advanced tokamak discharges in DIII-D

Description: Initial experiments have been performed on the DIII-D tokamak on coupling, direct electron heating, and current drive by fast waves in advanced tokamak discharges. These experiments showed efficient central heating and current drive in agreement with theory in magnitude and profile. Extrapolating these results to temperature characteristic of a power plant (25 keV) gives current drive efficiency of about 0.3 MA/m{sup 2}.
Date: July 1, 1995
Creator: Prater, R.; Cary, W.P. & Baity, F.W.
Partner: UNT Libraries Government Documents Department

Solenoid-free Plasma Start-up in NSTX using Transient CHI

Description: Experiments in NSTX have now unambiguously demonstrated the coupling of toroidal plasmas produced by the technique of CHI to inductive sustainment and ramp-up of the toroidal plasma current. This is an important step because an alternate method for plasma startup is essential for developing a fusion reactor based on the spherical torus concept. Elimination of the central solenoid would also allow greater flexibility in the choice of the aspect ratio in tokamak designs now being considered. The transient CHI method for spherical torus startup was originally developed on the HIT-II experiment at the University of Washington.
Date: November 3, 2008
Creator: Raman, R.; Nelson, B. A.; Mueller, D.; Jarboe, T. R.; Bell, M. G.; LeBlanc, B. et al.
Partner: UNT Libraries Government Documents Department

Stable Spheromaks with Profile Control

Description: A spheromak equilibrium with zero edge current is shown to be stable to both ideal MHD and tearing modes that normally produce Taylor relaxation in gun-injected spheromaks. This stable equilibrium differs from the stable Taylor state in that the current density j falls to zero at the wall. Estimates indicate that this current profile could be sustained by non-inductive current drive at acceptable power levels. Stability is determined using the NIMROD code for linear stability analysis. Non-linear NIMROD calculations with non-inductive current drive could point the way to improved fusion reactors.
Date: January 29, 2008
Creator: Fowler, T K & Jayakumar, R
Partner: UNT Libraries Government Documents Department

Compact tokamak reactors. Part 1 (analytic results)

Description: We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model.
Date: September 13, 1996
Creator: Wootton, A. J.; Wiley, J. C.; Edmonds, P. H. & Ross, D. W.
Partner: UNT Libraries Government Documents Department

Electron cyclotron current drive in DIII-D

Description: Clear measurements of the localized current density driven by electron cyclotron waves have been made on the DIII-D tokamak. Direct evidence of the current drive is seen on the internal magnetic field measurements by motional Stark effect spectroscopy. Comparison with theoretical calculations in the collisionless limit shows the experimental current drive exceeds the predictions by a substantial amount for currents driven near the half radius. In all cases the experimental current density profile is broader than the predicted one.
Date: May 1, 1999
Creator: Luce, T.C.; Lin-Liu, Y.R.; Lohr, J.M.; Petty, C.C.; Politzer, P.A.; Prater, R. et al.
Partner: UNT Libraries Government Documents Department

Upgrade of the DIII-D RF systems

Description: The DIII-D Advanced Tokamak Program requires the ability to modify the current density profile for extended time periods in order to achieve the improved plasma conditions now achieved with transient means. To support this requirement DIII-D has just completed a major addition to its ion cyclotron range of frequency (ICRF) systems. This upgrade project added two new fast wave current drive (FWCD) systems, with each system consisting of a 2 MW, 30 to 120 MHz transmitter, an all ceramic insulated transmission line, and water-cooled four-strap antenna. With this addition of 4 MW of FWCD power to the original 2 MW, 30 to 60 MHz capability, experiments can be performed with centrally localized current drive enhancement. For off-axis current modification, plans are in place to add 110 GHz electron cyclotron heating (ECH) power to DIII-D. Initially, 3 MW of power will be available with plans to increase the power to 6 MW and to 10 MW.
Date: October 1, 1995
Creator: Callis, R. W.; Cary, W. P. & O`Neill, R. C.
Partner: UNT Libraries Government Documents Department

Modeling of electron cyclotron current drive experiments on DIII-D

Description: Electron Cyclotron Current Drive (ECCD) is considered a leading candidate for current profile control in Advanced Tokamak (AT) operation. Localized ECCD has been clearly demonstrated in recent proof-of-principle experiments on DIII-D. The measured ECCD efficiency near the magnetic axis agrees well with standard theoretical predictions. However, for off-axis current drive the normalized experimental efficiency does not decrease with minor radius as expected from the standard theory; the observed reduction of ECCD efficiency due to trapped electron effects in the off-axis cases is smaller than theoretical predictions. The standard approach of modeling ECCD in tokamaks has been based on the bounce-average calculations, which assume the bounce frequency is much larger than the effective collision frequency for trapped electrons at all energies. The assumption is clearly invalid at low energies. Finite collisionality will effectively reduce the trapped electron fraction, hence, increase current drive efficiency. Here, a velocity-space connection formula is proposed to estimate the collisionality effect on electron cyclotron current drive efficiency. The collisionality correction gives modest improvement in agreement between theoretical and recent DIII-D experimental results.
Date: May 1, 1999
Creator: Lin-Liu, Y.R.; Chan, V.S.; Luce, T.C.; Prater, R.; Sauter, O. & Harvey, R.W.
Partner: UNT Libraries Government Documents Department

Initial tests and operation of a 110 GHz, 1 MW gyrotron with evacuated waveguide system on the DIII-D tokamak

Description: A gyrotron producing nominally 1 MW at 110 GHz has been installed at the DIII-D tokamak and operated in a program of initial tests with a windowless evacuated transmission line. The alignment and first test operation were performed in an air environment at atmospheric pressure. Under these conditions, the tube produced rf output in excess of 800 kW for pulse lengths greater than 10 msec and power near 500 kW for pulse lengths of about 100 msec into a free space dummy load. The gyrotron was operated into evacuated corrugated waveguide in the full power parameter regime for pulse lengths of up to 500 msec injecting greater than 0.5 MW into DIII-D for a preliminary series of experiments. Generated powers greater than 900 kW were achieved. A parasitic oscillation at various frequencies between 20 and 100 MHz, which was generated during the pulsing of the gyrotron electron beam, was suppressed somewhat by a capacitive filter attached to the gyrotron itself. Addition of a magnetic shield intended to alter the magnetic field geometry below the cathode eliminated internal tube sparks. Rework of the external power and interlock circuitry to improve the immunity to electromagnetic interference was also done in parallel so that the fast interlock circuitry could be used. The latest results of the test program, the design of the free space load and other test hardware, and the transmission line will be presented.
Date: August 1, 1996
Creator: Lohr, J.; Ponce, D. & Tooker, J.F.
Partner: UNT Libraries Government Documents Department

High-harmonic ion cyclotron heating and current drive in ultra-small aspect ratio tokamaks

Description: Ultra-small aspect ratio tokamaks present a totally new plasma environment for heating and current drive experiments and involve a number of physics issues that have not previously been explored. These devices operate at low magnetic field and relatively high density so that the effective dielectric constant of the plasma to high harmonic fast waves (HHFW), is quite high, and perpendicular wavelength of fast waves is very short. {lambda} {approximately} 2.0 cm compared with {lambda} - 10-20 cm. This makes possible strong electron absorption at high harmonics of the ion cyclotron frequency, {Omega}{sub i}, and at fairly high phase velocity in relation to electron thermal velocity. If the antenna system can control the parallel wave spectrum, this offers the promise of high efficiency off-axis current drive and the possibility for current drive radial profile control. Antenna phasing is ineffective for profile control in conventional tokamaks because of central absorption. There are also challenges for antenna design in this regime because of the high dielectric constant and the large angle of the magnetic field with respect to the equatorial plane ({approximately}45{degrees}), which varies greatly during current ramp. Preliminary experiments in this HHFW regime are being carried out in CDX-U.
Date: November 1, 1996
Creator: Batchelor, D.B.; Jaeger, E.F.; Carter, M.D. & Berry, L.A.
Partner: UNT Libraries Government Documents Department

ICRF antenna modeling and simulation. Final report, March 1, 1993--May 31, 1996

Description: SAIC has undergone a three year research and development program in support of the DOE Office of Fusion Energy`s (OFE) program in Ion Cyclotron Range of Frequencies (ICRF) heating of present, next generation, and future plasma fusion devices. The effort entailed advancing theoretical models and numerical simulation technology of ICRF physics and engineering issues associated predominately with, but not limited to, tokamak Ion Cyclotron Heating (ICH) and fast wave current drive (FWCD). Ion cyclotron heating and current drive is a central element in all current and planned large fusion experiments. In recent years, the variety of uses for ICRF systems has expanded, and includes the following: (1) Heating sufficient to drive plasma to ignition. (a) Second-harmonic T heating. (b) He{sup 3} minority heating. (2) Second-harmonic He{sup 4} heating in H plasma (for non-activated phase). (3) Detailed equilibrium profile control minority heating. (a) Ion minority (He{sup 3}) CD (for profile control on inside of plasma). (b) Ion minority (He{sup 3}) CD (for profile control on outside of plasma). (4) Ion-ion hybrid regime majority ion heating. (5) Electron current drive. (6) Mode conversion to drive current. (7) Deuterium minority heating. (8) Sawtooth instability stabilization. (9) Alpha particle parameter enhancement. (10) The generation of minority tails by ICRF to simulate D-T plasma particle physics in a deuterium plasma. Optimization of ICRF antenna performance for either heating or current drive depends critically on the complex balance and interplay between the plasma physics and the electromechanical system requirements. For example, ITER IC rf designs call for an IC. system frequency range from 20 MHz to 100 MHz. Additionally, antenna designs and operational modes that minimize impurity production and induced sheath formation may degrade current drive efficiency. Such effects have been observed in experiments involving it versus zero antenna phasing.
Date: August 30, 1996
Partner: UNT Libraries Government Documents Department

Applications of fast wave in spherical tokamaks

Description: In spherical tokamaks (ST), the magnetic field strength varies over a wide range across the plasma, and at high betas it deviates significantly from the 1/R dependence of conventional tokamaks. This, together with the high density expected in ST, poses challenging problems for RF heating and current drive. In this paper, the authors investigate the various possible applications of fast waves (FW) in ST. The adjoint technique of calculating current drive is implemented in the raytracing code CURRAY. The applicability of high harmonic and subharmonic FW to steady state ST is considered. They find that high harmonic FW tends to be totally absorbed before reaching the core and may be considered a candidate for off axis current drive while the subharmonic FW tends to be absorbed mainly in the core region and may be considered for central current drive. A difficult problem is the maintenance of current at the startup stage. In the bootstrap ramp-up scenario, the current ramp-up is mainly provided by the bootstrap current. Under this condition, the role of rf becomes mainly the sustainment of plasma through electron heating. Using a slab full-wave code SEMAL, the authors find that the ion-ion-hybrid mode conversion scheme is a promising candidate. The effect of possible existence of edge Alfven resonance and high harmonic cyclotron resonance is investigated and regimes of minimization of edge heating identified.
Date: April 1, 1997
Creator: Chiu, S.C.; Chan, V.S.; Lin-Liu, Y.R.; Miller, R.L.; Prater, R. & Politzer, P.
Partner: UNT Libraries Government Documents Department

High-harmonic fast wave heating experiments in CDX-U

Description: One of the primary objectives of the proposed National Spherical Tokamak Experiment (NSTX) is the investigation of very high {beta} regimes. Consequently, finding efficient methods of non-inductive heating and current drive required to heat and sustain such plasmas is of considerable importance. High-frequency fast waves are a promising candidate in this regard. However, in NSTX, the field-line pitch at the outer midplane will range from 0 up to 60 degrees from plasma start-up to current flattop. Thus, antenna strap orientation with respect to the edge magnetic field may have a serious impact on power coupling and absorption. To address this issue, the vacuum vessel of the Current Drive Experiment -- Upgrade (CDX-U) spherical tokamak has been upgraded to accommodate a rotatable two-strap antenna capable of handling several hundred kilowatts in short pulses. Details of the antenna design and results from loading measurements made as a function of power, strap angle, and strap phasing will be presented. Results from microwave scattering experiments will also be discussed.
Date: December 1, 1997
Creator: Menard, J.; Majeski, R.; Ono, M.; Wilson, J.R.; Munsat, T. & Seki, T.
Partner: UNT Libraries Government Documents Department

4 MW upgrade to DIII-D FWCD system: System commissioning and initial operation

Description: The initial installation of the 4 MW fast wave current drive (FWCD) upgrade started in 1992 with the purchase of two ABB/Thomcast AG rf power amplifiers. These amplifiers cover the frequency range 30 MHz to 120 MHz. A maximum output power of over 2 MW between 30 MHz and 80 MHz and 1 MW at 120 MHz were the specification requirements. The system as installed is comprised of the two mentioned rf amplifiers, coaxial transmission and matching components, rf phase and amplitude monitoring, and a SUN SparcStation 10 control system. Due to various reasons almost every major component in the system required redesign and engineering in order to meet the system requirements. The failures, probable cause and the final redesigns will be discussed as well as some thoughts on how better to specify system requirements for future systems.
Date: October 1995
Creator: Cary, W. P.; Callis, R. W.; de Grassie, J. S.; Harris, T. E.; O`Neill, R. C.; Pinsker, R. I. et al.
Partner: UNT Libraries Government Documents Department

Expectations for the National Spherical Torus Experiment`s high harmonic fast wave system

Description: High harmonic fast waves (HHFW) have been chosen as the primary method to drive steady state currents in the National Spherical Torus Experiment (NSTX). The somewhat limited experience with this frequency range in conventional tokamak plasma indicates that the coupling to electrons should be successful; however, there is no experimental data base for HHFWs in the unique and rapidly varying plasma regimes expected for NSTX. In this paper, the authors describe how the HHFW antenna was designed for NSTX using the computer codes to help make decisions that might affect the system`s performance and operation. The antenna geometry has been optimized to maintain the power handling and phase control requirements within engineering constraints. The physics issues that lead to the choice of poloidal current strap orientation are discussed. Expectations for current profile control using the antenna`s phase control system are also discussed.
Date: November 1, 1998
Creator: Carter, M.D.; Ryan, P.M. & Swain, D.W.
Partner: UNT Libraries Government Documents Department