21 Matching Results

Search Results

Advanced search parameters have been applied.

Operations with tritium neutral beams on TFTR

Description: In late 1993 the Tokamak Fusion Test Reactor began operating with a deuterium-tritium (DT) fuel mixture instead of the pure deuterium which it had used previously. The major portion of this tritium has initially entered the torus as energetic neutral beam particles. Over 600 deuterium-tritium discharges have now been studied with the aid of more than 2000 tritium ion source shots. The maximum total neutral particle power injected with a mix of deuterium and tritium beams has been 39.6 megawatts, and the maximum injected as tritium neutrals has been 24.3 megawatts. Tritium neutral beam operation has become routine during this time.
Date: December 31, 1996
Creator: Grisham, L.R.; O`Connor, T. & Stevenson, T.N.
Partner: UNT Libraries Government Documents Department

Gas utilization in TFTR (Tokamak Fusion Test Reactor) neutral beam injectors

Description: Measurements of gas utilization in a test TFTR neutral beam injector have been performed to study the feasibility of running tritium neutral beams with existing ion sources. Gas consumption is limited by the restriction of 50,000 curies of T/sub 2/ allowed on site. It was found that the gas efficiency of the present long-pulse ion sources is higher than it was with previous short-pulse sources. Gas efficiencies were studied over the range of 35 to 55%. At the high end of this range the neutral fraction of the beam fell below that predicted by room temperature molecular gas flow. This is consistent with observations made on the JET injectors, where it has been attributed to beam heating of the neutralizer gas and a concomitant increase in conductance. It was found that a working gas isotope exchange from H/sub 2/ to D/sub 2/ could be accomplished on the first beam shot after changing the gas supply, without any intermediate preconditioning. The mechanism believed responsible for this phenomenon is heating of the plasma generator walls by the arc and a resulting thermal desorption of all previously adsorbed and implanted gas. Finally, it was observed that an ion source conditioned to 120 kV operation could produce a beam pulse after a waiting period of fourteen hours by preceding the beam extraction with several hi-pot/filament warm-up pulses, without any gas consumption. 18 refs., 7 figs., 2 tabs.
Date: August 1, 1987
Creator: Kamperschroer, J.H.; Gammel, G.M.; Kugel, H.W.; Grisham, L.R.; Stevenson, T.N.; von Halle, A. et al.
Partner: UNT Libraries Government Documents Department

Experience with deuterium-tritium plasmas heated by high power neutral beams

Description: The Tokamak Fusion Test Reactor has operated since November of 1993 with a deuterium-tritium fuel mixture for selected discharges. The majority of the tritium has been introduced as energetic neutral atoms of up to 120 keV injected by the neutral beam systems, with some of the twelve ion sources run on pure tritium and some on deuterium to optimize the fuel mixture in the core plasma. A maximum beam power of 39.6 megawatts has been injected, and deuterium-tritium fusion power production has reached 10.7 megawatts, achieving central fusion power densities comparable to or greater than those expected for the International Thermonuclear Reactor, and allowing the first studies of fusion-produced alpha particle behavior in reactor grade plasmas. Energy confinement in deuterium-tritium plasmas is better than in similar deuterium plasmas for most plasma regimes. Innovative techniques to manipulate the plasma current and pressure profiles are permitting studies of enhanced confinement regimes.
Date: 1996
Creator: Grisham, L. R.; Kamperschroer, J. H.; O`Connor, T.; Oldaker, M.; Stevenson, T. & Von Halle, A.
Partner: UNT Libraries Government Documents Department

Measurement of ion profiles in TFTR neutral beamlines

Description: A technique is described whereby the ion dumps inside the TFTR Neutral Beam Test Stand were used to measure thermal profiles of the full-, half-, and third-energy ions. 136 thermocouples were installed on the full-energy ion dump, allowing full beam contours. Additional linear arrays across the widths of the half- and third-energy ion dumps provided a measure of the shape, in the direction parallel to the grid rails, of the half- and third-energy ions, and, hence, of the molecular ions extracted from the source. As a result of these measurements it was found that the magnet was more weakly focusing, by a factor of two, than expected, explaining past overheating of the full-energy ion dump. Hollow profiles on the half- and third-energy ion dumps were observed, suggesting that extraction of D{sub 2}+ and D{sub 3}+ are primarily from the edge of the ion source. If extraction of half-energy ions is from the edge of the accelerator, a divergence parallel to the grid rails of 0.6{degrees}{plus minus}0.1{degrees} results. It is postulated that a nonuniform gas profile near the accelerator is the cause of the hollow partial-energy ion profiles; the pressure being depressed over the accelerator by particles passing through this highly transparent structure. Primary electrons reaching the accelerator produce nonuniform densities of D{sub 2}+ through the ionization of this across the full-energy dump was examined as a means of reducing the power density. By unbalancing the current in the two coils of the magnet, on a shot by shot basis, by up to 2:1 ratio, it was possible to move the centerline of the full-energy ion beam sideways by {approximately}12.5 cm. The adoption of such a technique, with a ramp of the coil imbalance from 2:1 to 1:2 over a beam pulse, could reduce the power density by a factor of ...
Date: February 1, 1992
Creator: Kamperschroer, J.H.; Grisham, L.R.; Kugel, H.W.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A. et al.
Partner: UNT Libraries Government Documents Department

Measurement of ion profiles in TFTR neutral beamlines

Description: A technique is described whereby the ion dumps inside the TFTR Neutral Beam Test Stand were used to measure thermal profiles of the full-, half-, and third-energy ions. 136 thermocouples were installed on the full-energy ion dump, allowing full beam contours. Additional linear arrays across the widths of the half- and third-energy ion dumps provided a measure of the shape, in the direction parallel to the grid rails, of the half- and third-energy ions, and, hence, of the molecular ions extracted from the source. As a result of these measurements it was found that the magnet was more weakly focusing, by a factor of two, than expected, explaining past overheating of the full-energy ion dump. Hollow profiles on the half- and third-energy ion dumps were observed, suggesting that extraction of D{sub 2}+ and D{sub 3}+ are primarily from the edge of the ion source. If extraction of half-energy ions is from the edge of the accelerator, a divergence parallel to the grid rails of 0.6{degrees}{plus_minus}0.1{degrees} results. It is postulated that a nonuniform gas profile near the accelerator is the cause of the hollow partial-energy ion profiles; the pressure being depressed over the accelerator by particles passing through this highly transparent structure. Primary electrons reaching the accelerator produce nonuniform densities of D{sub 2}+ through the ionization of this across the full-energy dump was examined as a means of reducing the power density. By unbalancing the current in the two coils of the magnet, on a shot by shot basis, by up to 2:1 ratio, it was possible to move the centerline of the full-energy ion beam sideways by {approximately}12.5 cm. The adoption of such a technique, with a ramp of the coil imbalance from 2:1 to 1:2 over a beam pulse, could reduce the power density by a factor of {ge}1.5.
Date: February 1, 1992
Creator: Kamperschroer, J. H.; Grisham, L. R.; Kugel, H. W.; O`Connor, T. E.; Stevenson, T. N.; von Halle, A. et al.
Partner: UNT Libraries Government Documents Department

Status of the Control System on the National Spherical Torus Experiment (NSTX)

Description: In 2003, the NSTX plasma control system was used for plasma shape control using real-time equilibrium reconstruction (using the rtEFIT code - J. Ferron, et al., Nucl. Fusion 38 1055 (1998)). rtEFIT is now in routine use for plasma boundary control [D. A. Gates, et al., submitted to Nuclear Fusion (2005)]. More recently, the system has been upgraded to support feedback control of the resistive wall mode (RWM). This paper describes the hardware and software improvements that were made in support of these physics requirements. The real-time data acquisition system now acquires 352 channels of data at 5kHz for each NSTX plasma discharge. The latency for the data acquisition, which uses the FPDP (Front Panel Data Port) protocol, is measured to be {approx}8 microseconds. A Stand-Alone digitizer (SAD), designed at PPPL, along with an FPDP Input multiplexing module (FIMM) allows for simple modular upgrades. An interface module was built to interface between the FPDP output of the NSTX control system and the legacy Power Conversion link (PCLINK) used for communicating with the PPPL power supplies (first used for TFTR). Additionally a module has been built for communicating with the switching power amplifiers (SPA) recently installed on NSTX. In addition to the hardware developments, the control software [D. Mastrovito, Fusion Eng. And Design 71 65 (2004)] on the NSTX control system has been upgraded. The control computer is an eight processor (8x333MHz G4) built by Sky Computers (Helmsford, MA). The device driver software for the hardware described above will be discussed, as well as the new control algorithms that have been developed to control the switching power supplies for RWM control. An important initial task in RWM feedback is to develop a reliable mode detection algorithm.
Date: August 5, 2005
Creator: Gates, D.A.; Ferron, J.R.; Bell, M.; Gibney, T.; Johnson, R.; Marsala, R.J. et al.
Partner: UNT Libraries Government Documents Department

A Neutral Beam Injector Upgrade for NSTX

Description: The National Spherical Torus Experiment (NSTX) capability with a Neutral Beam Injector (NBI) capable of 80 kiloelectronvolt (keV), 5 Megawatt (MW), 5 second operation. This 5.95 million dollar upgrade reused a previous generation injector and equipment for technical, cost, and schedule reasons to obtain these specifications while retaining a legacy capability of 120 keV neutral particle beam delivery for shorter pulse lengths for possible future NSTX experiments. Concerns with NBI injection included power deposition in the plasma, aiming angles from the fixed NBI fan array, density profiles and beam shine through, orbit losses of beam particles, and protection of the vacuum vessel wall against beam impingement. The upgrade made use of the beamline and cryo panels from the Neutral Beam Test Stand facility, existing power supplies and controls, beamline components and equipment not contaminated by tritium during DT [deuterium-tritium] experiments, and a liquid Helium refrigerator plant to power and cryogenically pump a beamline and three ion sources. All of the Tokamak Fusion Test Reactor (TFTR) ion sources had been contaminated with tritium, so a refurbishment effort was undertaken on selected TFTR sources to rid the three sources destined for the NSTX NBI of as much tritium as possible. An interconnecting duct was fabricated using some spare and some new components to attach the beamline to the NSTX vacuum vessel. Internal vacuum vessel armor using carbon tiles was added to protect the stainless steel vacuum vessel from beam impingement in the absence of plasma and interlock failure. To date, the NBI has operated to 80 keV and 5 MW and has injected requested power levels into NSTX plasmas with good initial results, including high beta and strong heating characteristics at full rated plasma current.
Date: January 18, 2002
Creator: Stevenson, T.; McCormack, B; Loesser, G.D.; Kalish, M.; Ramakrishnan, S.; Grisham, L. et al.
Partner: UNT Libraries Government Documents Department

Plasma Shape Control on the National Spherical Torus Experiment (NSTX) using Real-time Equilibrium Reconstruction

Description: Plasma shape control using real-time equilibrium reconstruction has been implemented on the National Spherical Torus Experiment (NSTX). The rtEFIT code originally developed for use on DIII-D was adapted for use on NSTX. The real-time equilibria provide calculations of the flux at points on the plasma boundary, which is used as input to a shape control algorithm known as isoflux control. The flux at the desired boundary location is compared to a reference flux value, and this flux error is used as the basic feedback quantity for the poloidal-field coils on NSTX. The hardware that comprises the control system is described, as well as the software infrastructure. Examples of precise boundary control are also presented.
Date: April 15, 2005
Creator: Gates, D.A.; Ferron, J.R.; Bell, M.; Gibney, T.; Johnson, R.; Marsala, R.J. et al.
Partner: UNT Libraries Government Documents Department

Operation of a TFTR ion source with a ground potential gas feed into the neutralizer

Description: TFTR long pulse ion sources have been operated with gas fed only into the neutralizer. Gas for the plasma generator entered through the accelerator rather than directly into the arc chamber. This modification has been proposed for tritium beam operation to locate control electronics at ground potential and to simplify tritium plumbing. Source operation with this configuration and with the nominal gas system which feeds gas into both the ion source and the center of the neutralizer are compared. Comparison is based upon accelerator grid currents, beam composition, and neutral power delivered to the calorimeter. Charge exchange in the accelerator can be a significant loss mechanism in both systems at high throughput. A suitable operating point with the proposed system was found that requires 30% less gas than used presently. The extracted D{sup +}, D{sub 2}{sup +}, and D{sub 3}{sup +} fractions of the beam were found to be a function of the gas throughput; at similar throughputs, the two gas feed systems produced similar extracted ion fractions. Operation at the proposed gas efficient point results in a small reduction (relative to the old high throughput mode) in the extracted D{sup +} fraction of the beam from 77% to 71%, with concomitant changes in the D{sub 2}{sup +} fraction from 18% to 26%, and 6% to 3% for D{sub 3}{sup +}. 26 refs., 7 figs.
Date: July 1, 1991
Creator: Kamperschroer, J.H.; Dudek, L.E.; Grisham, L.R.; Newman, R.A.; O'Conner, T.E.; Stevenson, T.N. et al.
Partner: UNT Libraries Government Documents Department

Temporal behavior of neutral particle fluxes in TFTR (Tokamak Fusion Test Reactor) neutral beam injectors

Description: Data from an E {parallel} B charge exchange neutral analyzer (CENA), which views down the axis of a neutral beamline through an aperture in the target chamber calorimeter of the TFTR neutral beam test facility, exhibit two curious effects. First, there is a turn-on transient lasting tens of milliseconds having a magnitude up to three times that of the steady-state level. Second, there is a 720 Hz, up to 20% peak-to-peak fluctuation persisting the entire pulse duration. The turn-on transient occurs as the neutralizer/ion source system reaches a new pressure equilibrium following the effective ion source gas throughput reduction by particle removal as ion beam. Widths of the transient are a function of the gas throughput into the ion source, decreasing as the gas supply rate is reduced. Heating of the neutalizer gas by the beam is assumed responsible, with gas temperature increasing as gas supply rate is decreased. At low gas supply rates, the transient is primarliy due to dynamic changes in the neutralizer line density and/or beam species composition. Light emission from the drift duct corroborate the CENA data. At high gas supply rates, dynamic changes in component divergence and/or spatial profiles of the source plasma are necessary to explain the observations. The 720 Hz fluctuation is attributed to a 3% peak-to-peak ripple of 720 Hz on the arc power supply amplified by the quadratic relationship between beam divergence and beam current. Tight collimation by CENA apertures cause it to accept a very small part of the ion source's velocity space, producing a signal linearly proportional to beam divergence. Estimated fluctuations in the peak power density delivered to the plasma under these conditions are a modest 3--8% peak to peak. The efffects of both phenomena on the injected neutral beam can be ameliorated by careful operion of the ...
Date: September 1, 1989
Creator: Kamperschroer, J.H.; Gammel, G.M.; Roquemore, A.L.; Grisham, L.R.; Kugel, H.W.; Medley, S.S. et al.
Partner: UNT Libraries Government Documents Department

Operation of TFTR neutral beams with heavy ions

Description: High Z neutral atoms have been injected into TFTR plasmas in an attempt to enhance plasma confinement through modification of the edge electric field. TFTR ion sources have extracted 9 A of 62 keV Ne{sup +} for up to 0.2 s during injection into deuterium plasmas, and for 0.5 s during conditioning pulses. Approximately 400 kW of Ne{sup 0} have been injected from each of two ion sources. Operation was at full bending magnet current, with the Ne{sup +} barely contained on the ion dump. Beamline design modifications to permit operation up to 120 keV with krypton or xenon are described. Such ions are too massive to be deflected up to the ion dump. The plan, therefore, is to armor those components receiving these ions. Even with this armor, modest increases in the bending magnet current capability are necessary to safely reach 120 kV with Kr or Xe. Information relevant to heavy ion operation was also acquired when several ion sources were inadvertently operated with water contamination. Spectroscopic analysis of certain pathological pulses indicate that up to 6% of the extracted ions were water. After dissociation in the neutralizer, water yields oxygen ions which, as with Ne, Kr, and Xe, are under-deflected by the magnet. Damage to a calorimeter scraper, due to the focal properties of the magnet, has resulted. A magnified power density of 6 KW/cm{sup 2} for 2 s, from {approximately} 90 kW of O{sup +}, is the suspected cause. 11 refs., 4 figs.
Date: July 1, 1991
Creator: Kamperschroer, J.H.; Stevenson, T.N.; Wright, K.E.; Dudek, L.E.; Grisham, L.R.; Newman, R.A. et al.
Partner: UNT Libraries Government Documents Department

Particle reflection and TFTR neutral beam diagnostics

Description: Determination of two critical neutral beam parameters, power and divergence, are affected by the reflection of a fraction of the incident energy from the surface of the measuring calorimeter. On the TFTR Neutral Beam Test Stand, greater than 30% of the incident power directed at the target chamber calorimeter was unaccounted for. Most of this loss is believed due to reflection from the surface of the flat calorimeter, which was struck at a near grazing incidence (12{degrees}). Beamline calorimeters, of a V''-shape design, while retaining the beam power, also suffer from reflection effects. Reflection, in this latter case, artificially peaks the power toward the apex of the V'', complicating the fitting technique, and increasing the power density on axis by 10 to 20%; an effect of import to future beamline designers. Agreement is found between measured and expected divergence values, even with 24% of the incident energy reflected.
Date: April 1, 1992
Creator: Kamperschroer, J.H.; Grisham, L.R.; Kugel, H.W.; O'Connor, T.E.; Newman, R.A.; Stevenson, T.N. et al.
Partner: UNT Libraries Government Documents Department

Low Z impurity ion extraction from TFTR ion sources

Description: TFTR deuterium neutral beams have been operated unintentionally with significant quantities of extracted water ions. Water has been observed with an Optical Multichannel Analyzer (OMA) during beam extraction when small water leaks were present within the arc chamber. These leaks were thermally induced with the contamination level increasing linearly with pulse length. 6% of the beam current was attributed to water ions for the worst leak, corresponding to an instantaneous value of 12% at the end of a 1.5 s pulse. A pre-calorimeter collimator was damaged as a result of this operation. A similar contamination is observed during initial operation of ion sources exposed to air. This latter contamination is attributed to the synthesis, from adsorbed air, of either D[sub 2]O or the indistinguishable ND[sub 3]. Initial operation of new ion sources typically produces a contamination level of [approximately]2%. These impurities are reduced to undetectable levels after 50 to 100 beam pulses. Once a water molecule is present in the plasma generator, it is predominantly ionized rather than dissociated, resulting in the extraction of only trace amounts of hydrogenated ions. The addition of water to the extracted beam also reduces the optimum perveance, moving the typical underdense operating point closer to optimum, causing the frequency of grid faults to increase. Close to 90% of the water extracted from ion sources with water leaks was deuterated, implying that the potential exists for the production of tritiated water during TFTR's forthcoming DT operation. Isotope exchange in the plasma generator takes place rapidly and is believed to be surface catalyzed. The primary concern is with O implanted into beam absorbers recombining with tritium, and the subsequent hold up of T[sub 2]O on cryopanels. Continuous surveillance with the OMA diagnostic during DT operation will ensure that ion sources with detectable water are not operated ...
Date: April 1, 1993
Creator: Kamperschroer, J.H.; Grisham, L.R.; Newman, R.A.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A. et al.
Partner: UNT Libraries Government Documents Department

Next-Step Spherical Torus Experiment and Spherical Torus Strategy in the Fusion Energy Development Path

Description: A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction.
Date: October 27, 2003
Creator: Ono, M.; Peng, M.; Kessel, C.; Neumeyer, C.; Schmidt, J.; Chrzanowski, J. et al.
Partner: UNT Libraries Government Documents Department

ELMs and the H-mode Pedestal in NSTX

Description: We report on the behavior of ELMs in NBI-heated H-mode plasmas in NSTX. It is observed that the size of Type I ELMs, characterized by the change in plasma energy, decreases with increasing density, as observed at conventional aspect ratio. It is also observed that the Type I ELM size decreases as the plasma equilibrium is shifted from a symmetric double-null toward a lower single-null configuration. Type III ELMs have also been observed in NSTX, as well as a high-performance regime with small ELMs which we designate Type V. These Type V ELMs are consistent with high bootstrap current operation and density approaching Greenwald scaling. The Type V ELMs are characterized by an intermittent n=1 MHD mode rotating counter to the plasma current. Without active pumping, the density rises continuously through the Type V phase. However, efficient in-vessel pumping should allow density control, based on particle containment time estimates.
Date: July 16, 2004
Creator: Maingi, R.; Sabbagh, S.A.; Bush, C.E.; Fredrickson, E.D.; Menard, J.E.; Stutman, D. et al.
Partner: UNT Libraries Government Documents Department

Cryosorption of helium on argon frost TFTR (Tokamak Fusion Test Reactor) neutral beamlines

Description: Helium pumping on argon frost has been investigated on TFTR neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are {approximately} 25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, > 20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a two-fold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 torr-l of helium, the beamline base pressure rose to 2{times}10{sup -5} torr from an initial value of 10{sup -8} torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 torr-l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity. 29 refs., 7 figs.
Date: November 1, 1989
Creator: Kamperschroer, J.H.; Cropper, M.B.; Dylla, H.F.; Garzotto, V.; Dudek, L.E.; Grisham, L.R. et al.
Partner: UNT Libraries Government Documents Department

H-Mode Turbulence, Power Threshold, ELM, and Pedestal Studies in NSTX

Description: High-confinement mode (H-mode) operation plays a crucial role in NSTX [National Spherical Torus Experiment] research, allowing higher beta limits due to reduced plasma pressure peaking, and long-pulse operation due to high bootstrap current fraction. Here, new results are presented in the areas of edge localized modes (ELMs), H-mode pedestal physics, L-H turbulence, and power threshold studies. ELMs of several other types (as observed in conventional aspect ratio tokamaks) are often observed: (1) large, Type I ELMs, (2) ''medium'' Type II/III ELMs, and (3) giant ELMs which can reduce stored energy by up to 30% in certain conditions. In addition, many high-performance discharges in NSTX have tiny ELMs (newly termed Type V), which have some differences as compared with ELM types in the published literature. The H-mode pedestal typically contains between 25-33% of the total stored energy, and the NSTX pedestal energy agrees reasonably well with a recent international multi-machine scaling. We find that the L-H transition occurs on a {approx}100 {micro}sec timescale as viewed by a gas puff imaging diagnostic, and that intermittent quiescent periods precede the final transition. A power threshold identity experiment between NSTX and MAST shows comparable loss power at the L-H transition in balanced double-null discharges. Both machines require more power for the L-H transition as the balance is shifted toward lower single null. High field side gas fueling enables more reliable H-mode access, but does not always lead to a lower power threshold e.g., with a reduction of the duration of early heating. Finally the edge plasma parameters just before the L-H transition were compared with theories of the transition. It was found that while some theories can separate well-developed L- and H-mode data, they have little predictive value.
Date: October 28, 2004
Creator: Maingi, R.; Bush, C.E.; Fredrickson, E.D.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P. et al.
Partner: UNT Libraries Government Documents Department

The National Spherical Torus Experiment (NSTX) Research Program and Progress Towards High Beta, Long Pulse Operating Scenarios

Description: A major research goal of the National Spherical Torus Experiment is establishing long-pulse, high-beta, high-confinement operation and its physics basis. This research has been enabled by facility capabilities developed over the last two years, including neutral-beam (up to 7 MW) and high-harmonic fast-wave heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with <beta {sub T}> up to 35%. Normalized beta values often exceed the no wall limit, and studies suggest that passive wall mode stabilization is enabling this for broad pressure profiles characteristic of H-mode plasmas. The viability of long, high bootstrap-current fraction operations has been established for ELMing H-mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fueling are likely contributing to a reduction in H-mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary-heated plasmas examined thus far. High-harmonic fast-wave (HHFW) effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is by comparing of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. A peak heat flux of 10 MW/m superscript ''2'' has been measured in the H-mode, with large asymmetries in the power deposition being observed between the inner and outer strike points. Noninductive plasma start-up studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current-drive techniques have begun.
Date: October 15, 2002
Creator: Synakowski, E. J.; Bell, M. G.; Bell, R. E.; Bigelow, T.; Bitter, M.; Blanchard, W. et al.
Partner: UNT Libraries Government Documents Department

Progress Towards High Performance, Steady-state Spherical Torus

Description: Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} ...
Date: October 2, 2003
Creator: Ono, M.; Bell, M.G.; Bell, R.E.; Bigelow, T.; Bitter, M.; Blanchard, W. et al.
Partner: UNT Libraries Government Documents Department

Status and Plans for the National Spherical Torus Experimental Research Facility

Description: An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.
Date: July 27, 2005
Creator: Columbia University
Partner: UNT Libraries Government Documents Department