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Recent Fast Wave Coupling and Heating Studies on NSTX, with Possible Implications for ITER

Description: The goal of the high harmonic fast wave (HHFW) research on NSTX is to maximize the coupling of RF power to the core of the plasma by minimizing the coupling of RF power to edge loss processes. HHFW core plasma heating efficiency in helium and deuterium L-mode discharges is found to improve markedly on NSTX when the density 2 cm in front of the antenna is reduced below that for the onset of perpendicular wave propagation (nonset ∝ B*k|| 2/ω). In NSTX, the observed RF power losses in the plasma edge are driven in the vicinity of the antenna as opposed to resulting from multi-pass edge damping. PDI surface losses through ion-electron collisions are estimated to be significant. Recent spectroscopic measurements suggest that additional PDI losses could be caused by the loss of energetic edge ions on direct loss orbits and perhaps result in the observed clamping of the edge rotation. Initial deuterium H-mode heating studies reveal that core heating is degraded at lower kφ (- 8 m-1 relative to 13 m-1) as for the Lmode case at elevated edge density. Fast visible camera images clearly indicate that a major edge loss process is occurring from the plasma scrape off layer (SOL) in the vicinity of the antenna and along the magnetic field lines to the lower outer divertor plate. Large type I ELMs, which are observed at both kφ values, appear after antenna arcs caused by precursor blobs, low level ELMs, or dust. For large ELMs without arcs, the source reflection coefficients rise on a 0.1 ms time scale, which indicates that the time derivative of the reflection coefficient can be used to discriminate between arcs and ELMs.
Date: July 21, 2009
Creator: J.C. Hosea, R.E. Bell, E. Feibush, R.W. Harvey, E.F. Jaeger, B.P LeBlanc, R. Maingi, C.K. Phillips, L. Roquemore, P.M. Ryan, G. Taylor, K. Tritz, E.J. Valeo, J. Wilgen, J.R. Wilson, and the NSTX Team
Partner: UNT Libraries Government Documents Department

Improvement in Plasma Performance with Lithium Coatings in NSTX

Description: Lithium as a plasma-facing material has attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Dramatic effects on plasma performance with lithium-coated plasma-facing components (PFC's) have been demonstrated on many fusion devices, including TFTR, T-11M, and FT-U. Using a liquid-lithium-filled tray as a limiter, the CDX-U device achieved very significant enhancement in the confinement time of ohmically heated plasmas. The recent NSTX experiments reported here have demonstrated, for the first time, significant and recurring benefits of lithium PFC coatings on divertor plasma performance in both L- and H- mode regimes heated by neutral beams.
Date: February 17, 2009
Creator: Kaita, R
Partner: UNT Libraries Government Documents Department

Fivefold confinement time increase in the Madison Symmetric Torus using inductive poloidal current drive

Description: Current profile control is employed in the Madison Symmetric Torus reversed field pinch to reduce the magnetic fluctuations responsible for anomalous transport. An inductive poloidal electric field pulse is applied in the sense to flatten the parallel current profile, reducing the dynamo fluctuation amplitude required to sustain the equilibrium. This technique demonstrates a substantial reduction in fluctuation amplitude (as much as 50%), and improvement in energy confinement (from 1 ms to 5 ms); a record low fluctuation (0.8%) and record high temperature (615 eV) for this device were observed simultaneously during current drive experiments. Plasma beta increases by 50% and the Ohmic input power is three times lower. Particle confinement improves and plasma impurity contamination is reduced. The results of the transient current drive experiments provide motivation for continuing development of steady-state current profile control strategies for the reversed field pinch.
Date: December 1, 1996
Creator: Stoneking, M.R.; Lanier, N.E.; Prager, S.C.; Sarff, J.S. & Sinitsyn, D.
Partner: UNT Libraries Government Documents Department

Strong radial electric field shear and reduced fluctuations in a reversed-field pinch

Description: A strongly sheared radial electric field is observed in enhanced confinement discharges in the MST reversed-field pinch. The strong shear develops in a narrow region in the plasma edge. Electrostatic fluctuations are reduced over the entire plasma edge with an extra reduction in the shear region. Magnetic fluctuations, resonant in the plasma core but global in extent, are also reduced. The reduction of fluctuations in the shear region is presumably due to the strong shear, but the causes of the reductions outside this region have not been established.
Date: May 1, 1997
Creator: Chapman, B. E.; Chiang, C. S.; Prager, S. C. & Sarff, J. S.
Partner: UNT Libraries Government Documents Department

TRANSPORT STUDIES IN DIII-D WITH MODULATED ECH

Description: Experiments have been performed where the T{sub e} profile stiffness was tested at several spatial locations by varying the ECH resonance location. Propagation of the pulses was Fourier analyzed and compared to simulations based on several transport models. The plasma appears to be near the critical T{sub e} gradient for ETG modes and marginally stable to ITG modes. However, the local T{sub e} response to a locally applied heat pulse does not indicate a nonlinear, critical gradient model where T{sub e} is clipped when trying to rise above a critical gradient. The response can be simply understood as the plasma integrating the ECH power, producing an increase in T{sub e} which equilibrates to a new local level with an exponential time constant representing the local confinement time.
Date: July 1, 2002
Creator: DeBOO, J.C.; AUSTIN, M.E.; BRAVENEC, R.V.; KINSEY, J.E; LOHR, J.; LUCE, T.C. et al.
Partner: UNT Libraries Government Documents Department

Theoretical aspects of energy confinement in spheromaks

Description: It is shown that, despite the poor global energy confinement observed in spheromak experiments to date, the long-term prospects may be favorable as spheromaks are scaled to larger size and higher temperatures. The present performance is traced to excessive magnetic energy loss at the edge compared to tokamaks and heat transport due to magnetic fluctuations, both of which should scale away as the temperature increases.
Date: November 16, 1994
Creator: Fowler, T.K.
Partner: UNT Libraries Government Documents Department

Advanced tokamak operating modes in TPX and ITER

Description: A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER.
Date: December 31, 1994
Creator: Nevins, W.M.
Partner: UNT Libraries Government Documents Department

DIII-D Advanced Tokamak Research Overview

Description: This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously {beta}{sub N}H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues.
Date: December 1, 1999
Creator: Chan, V.S.; Greenfield, C.M.; Lao, L.L.; Luce, T.C.; Petty, C.C. & Staebler, G.M.
Partner: UNT Libraries Government Documents Department

JET Radiative Mantle Experiments in ELMy H-Mode

Description: Radiative mantle experiments were performed on JET ELMy H-mode plasmas. The Septum configuration was used where the X-point is embedded into the top of the Septum. Argon radiated 50% of the input power from the bulk plasma while Z{sub eff} rose from an intrinsic level of 1.5 to about 1.7 due to the injected Argon. The total energy content and global energy confinement time decreased 15% when the impurities were introduced. In contrast, the effective thermal diffusivity in the core confinement region (r/a = .4--.8) decreased by 30%. Usually, JET ELMy H-mode plasmas have confinement that is correlated to the edge pedestal pressure. The radiation lowered the edge pedestal and consequently lowered the global confinement. Thus the confinement was changed by a competition between the edge pedestal reduction lowering the confinement and the weaker RI effect upon the core transport coefficients raising the confinement. The ELM frequency increased from 10 Hz Type I ELMs, to 200 Hz type III ELMs. The energy lost by each ELM reduced to 0.5% of the plasma energy content.
Date: September 28, 1999
Creator: Budny, R.; Coffey, I.; Dumortier, P.; Grisolia, C.; Strachan, J.D. & al, et
Partner: UNT Libraries Government Documents Department

Recent Progress on the National Spherical Torus Experiment (NSTX)

Description: Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal {beta}{sub T} (= 2{micro}{sub 0}<p>/B{sub T}{sup 2} where B{sub T} is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized {beta}{sub N} (= {beta}{sub T}aB{sub I}/I{sub p}) {approx} 6% {center_dot} m {center_dot} T/MA.. The highest {beta} discharge exceeded the calculated no-wall {beta} limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to {beta}{sub T} {approx} 20% and {beta}{sub N} = 5.4. Long pulse ({approx}1s) high {beta}{sub p} ({approx}1.5) discharges have also been obtained at higher {beta}{sub {phi}} (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of {approx}1.5 times ITER98pby2 for several {tau}{sub E} are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current.
Date: July 2, 2002
Creator: Gates, D. A.; Bell, M. G.; Bell, R. E.; Bialek, J.; Bigelow, T.; Bitter, M. et al.
Partner: UNT Libraries Government Documents Department

Recent Improvements in Fast Wave Heating in NSTX

Description: Recent improvements in high-harmonic fast wave (HHFW) core heating in NSTX are attributed to using lithium conditioning, and other wall conditioning techniques, to move the onset density for perpendicular fast wave propagation further from the antenna. This has resulted in the first observation of HHFW core electron heating in deuterium plasma at a launched toroidal wavenumber, kφ = -3 m-1, NSTX record core electron temperatures of 5 keV in helium and deuterium discharges and, for the first time, significant HHFW core electron heating of deuterium neutral-beam-fuelled H-mode plasmas. Also, kφ = -8 m-1 heating of the plasma startup and plasma current ramp-up has resulted in significant core electron heating, even at central electron densities as low as ~ 4x1018 m-3.
Date: June 26, 2009
Creator: G. Taylor, R.E. Bell, R.W. Harvey, J.C. Hosea, E.F. Jaeger, B.P. LeBlanc, C.K. Phillips, P.M. Ryan, E.J. Valeo, J.B. Wilgen, J.R. Wilson, and the NSTX Team
Partner: UNT Libraries Government Documents Department

Recent progress on the National Spherical Torus Experiment (NSTX)

Description: Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal {beta}{sub T} (= 2{mu}{sub 0}<p>/B{sub T}{sup 2} where B{sub T} is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized {beta}{sub N} (={beta}{sub T}aB{sub I}/I{sub p}) {approx} 6% {center_dot} m {center_dot} T/MA. The highest {beta} discharge exceeded the calculated no-wall {beta} limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to {beta}{sub T} {approx} 20% and {beta}{sub N} = 5.4. Long pulse ({approx}1s) high {beta}{sub p} ({approx}1.5) discharges have also been obtained at higher B{sub {phi}} (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of {approx} 1.5 times ITER98pby2 for several {tau}{sub E} are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current.
Date: January 1, 2002
Creator: Maqueda, R. J. (Ricardo J.); Wurden, G. A. (Glen A.); Gates, D. A.; Bell, M. G.; Bialek, J.; Bigelow, T. et al.
Partner: UNT Libraries Government Documents Department

Improvement in Plasma Performance with Lithium Coatings in NSTX

Description: Lithium as a plasma-facing material has attractive features, including a reduction in the recycling of hydrogenic species and the potential for withstanding high heat and neutron fluxes in fusion reactors. Dramatic effects on plasma performance with lithium-coated plasma-facing components (PFCOs) have been demonstrated on many fusion devices, including TFTR, [1] T-11M, [2] and FT-U. [3] Using a liquid-lithium-filled tray as a limiter, the CDX-U device achieved very significant enhancement in the confinement time of ohmically heated plasmas. [4] The recent NSTX experiments reported here have demonstrated, for the first time, significant and recurring benefits of lithium PFC coatings on divertor plasma performance in both L- and H- mode regimes heated by neutral beams.
Date: September 12, 2008
Creator: Kaita, R; Ahn, J -W; Allain, J P; Bell, M G; Bell, R; Boedo, J et al.
Partner: UNT Libraries Government Documents Department

Progress towards sustainment of advanced tokamak modes in DIII-D

Description: Improving confinement and beta limits simultaneously in long-pulse ELMy H-mode discharges is investigated. The product {beta}{sub N}H{sub 98y} serves as a useful figure-of-merit for performance, where {beta}{sub N} {triple_bond} {beta}/(I/aB) and H{sub 98y} is the ratio of the thermal confinement time relative to the most recent ELMy H-mode confinement scaling established by the ITER confinement database working group. In discharges with q{sub 0} {approximately} 1 (no sawteeth) and discharges with q{sub min} > 1.5 and negative central magnetic shear, {beta}{sub N} {approximately} 2.9 and H{sub 98y} {approximately} 1.4 are sustained for up to 2 s. Although peaked profiles are observed, steep internal transport barriers are not present. Further increases in {beta}{sub N} in these discharges is limited by neoclassical tearing modes (NTM) in the positive shear region. In another recently developed regime, {beta}{sub N} {approximately} 3.8 and H{sub 98y} {approximately} 1.8 has been sustained during large infrequent ELMs in non-sawtoothing discharges with 1{sub 0} {approximately} 1. This level of performance is similar to that obtained in ELM-free regimes such as VH-mode. The limitation on {beta}{sub N} and pulse length in these discharges is also the onset of NTMs.
Date: December 1998
Creator: Rice, B. W.; Burrell, K. H. & Ferron, J. R.
Partner: UNT Libraries Government Documents Department

METHANE PENTRATION IN DIII-D ELMing H-MODE PLASMAS

Description: Carbon penetration into the core plasma during midplane and divertor methane puffing has been measured for DIII-D ELMing H-mode plasmas. The methane puffs are adjusted to a measurable signal, but global plasma parameters are only weakly affected (line average density, <n{sub e}> increases by < 10%, energy confinement time, {tau}{sub E} drops by < 10%). The total carbon content is derived from C{sup +6} density profiles in the core measured as a function of time using charge exchange recombination spectroscopy. The methane penetration factor is defined as the difference in the core content with the puff on and puff off, divided by the carbon confinement time and the methane puffing rate. In ELMing H-mode discharges with ion {del}B drift direction into the X-point, increasing the line averaged density from 5 to 8 x 10{sup 19} m{sup -3} dropped the penetration factor from 6.6% to 4.6% for main chamber puffing. The penetration factor for divertor puffing was below the detection limit (<1%). Changing the ion {del}B drift direction to away from the X-point decreased the penetration factor by more than a factor of five for main chamber puffing.
Date: June 1, 2002
Creator: WEST, W.P.; LASNIER, C.J.; WHYTE, D.G.; ISLER, R.C.; EVANS, T.E.; JACKSON, G.L. et al.
Partner: UNT Libraries Government Documents Department

On the dynamics of turbulent transport near marginal stability

Description: A general methodology for describing the dynamics of transport near marginal stability is formulated. Marginal stability is a special case of the more general phenomenon of self-organized criticality. Simple, one field models of the dynamics of tokamak plasma self-organized criticality have been constructed, and include relevant features such as sheared mean flow and transport bifurcations. In such models, slow mode (i.e. large scale, low frequency transport events) correlation times determine the behavior of transport dynamics near marginal stability. To illustrate this, impulse response scaling exponents (z) and turbulent diffusivities (D) have been calculated for the minimal (Burgers) and sheared flow models. For the minimal model, z = 1 (indicating ballastic propagation) and D {approximately}(S{sub 0}{sup 2}){sup 1/3}, where S{sub 0}{sup 2} is the noise strength. With an identically structured noise spectrum and flow with shearing rate exceeding the ambient decorrelation rate for the largest scale transport events, diffusion is recovered with z = 2 and D {approximately} (S{sub 0}{sup 2}){sup 3/5}. This indicates a qualitative change in the dynamics, as well as a reduction in losses. These results are consistent with recent findings from {rho} scaling scans. Several tokamak transport experiments are suggested.
Date: March 1, 1995
Creator: Diamond, P.H. & Hahm, T.S.
Partner: UNT Libraries Government Documents Department

Stability of negative central magnetic shear discharges in the DIII-D tokamak

Description: Discharges with negative central magnetic shear (NCS) hold the promise of enhanced fusion performance in advanced tokamaks. However, stability to long wavelength magnetohydrodynamic modes is needed to take advantage of the improved confinement found in NCS discharges. The stability limits seen in DIII-D experiments depend on the pressure and current density profiles and are in good agreement with stability calculations. Discharges with a strongly peaked pressure profile reach a disruptive limit at low beta, {beta}{sub N} = {beta} (I/aB){sup -1} {le} 2.5 (% m T/MA), caused by an n = 1 ideal internal kink mode or a global resistive instability close to the ideal stability limit. Discharges with a broad pressure profile reach a soft beta limit at significantly higher beta, {beta}{sub N} = 4 to 5, usually caused by instabilities with n > 1 and usually driven near the edge of the plasma. With broad pressure profiles, the experimental stability limit is independent of the magnitude of negative shear but improves with the internal inductance, corresponding to lower current density near the edge of the plasma. Understanding of the stability limits in NCS discharges has led to record DIII-D fusion performance in discharges with a broad pressure profile and low edge current density.
Date: December 1, 1996
Creator: Strait, E.J.; Chu, M.S. & Ferron, J.R.
Partner: UNT Libraries Government Documents Department

Threshold power and energy confinement for ITER

Description: In order to predict the threshold power for L-H transition and the energy confinement performance in ITER, assembling of database and analyses of them have been progressed. The ITER Threshold Database includes data from 10 divertor tokamaks. Investigation of the database gives a scaling of the threshold power of the form P{sub thr} {proportional_to} B{sub t} n{sub e}{sup 0.75} R{sup 2} {times} (n{sub e} R{sup 2}){sup +-0.25}, which predicts P{sub thr} = 100 {times} 2{sup 0{+-}1} MW for ITER at n{sub e} = 5 {times} 10{sup 19} m{sup {minus}3}. The ITER L-mode Confinement Database has also been expanded by data from 14 tokamaks. A scaling of the thermal energy confinement time in L-mode and ohmic phases is obtained; {tau}{sub th} {approximately} I{sub p} R{sup 1.8} n{sub e}{sup 0.4{sub P{sup {minus}0.73}}}. At the ITER parameter, it becomes about 2.2 sec. For the ignition in ITER, more than 2.5 times of improvement will be required from the L-mode. The ITER H-mode Confinement Database is expanded from data of 6 tokamaks to data of 11 tokamaks. A {tau}{sub th} scaling for ELMy H-mode obtained by a standard regression analysis predicts the ITER confinement time of {tau}{sub th} = 6 {times} (1 {+-} 0.3) sec. The degradation of {tau}{sub th} with increasing n{sub e} R{sup 2} (or decreasing {rho}{sub *}) is not found for ELMy H-mode. An offset linear law scaling with a dimensionally correct form also predicts nearly the same {tau}{sub th} value.
Date: December 31, 1996
Creator: Takizuka, T.
Partner: UNT Libraries Government Documents Department

Comprehensive energy transport scalings derived from DIII-D similarity experiments

Description: The dependences of heat transport on the dimensionless plasma physics parameters has been measured for both L-mode and H-mode plasmas on the DIII-D tokamak. Heat transport in L-mode plasmas has a gyroradius scaling that is gyro-Bohm-like for electrons and worse than Bohm-like for ions, with no measurable beta or collisionality dependence; this corresponds to having an energy confinement time that scales like {tau}{sub E} {proportional_to} n{sup 0.5}P{sup {minus}0.5}. H-mode plasmas have gyro-Bohm-like scaling of heat transport for both electrons and ions, weak beta scaling, and moderate collisionality scaling. In addition, H-mode plasmas have a strong safety factor scaling ({chi} {approximately} q{sup 2}) at all radii. Combining these four dimensionless parameter scalings together gives an energy confinement time scaling for H-mode plasmas like {tau}{sub E} {proportional_to} B{sup {minus}1}{rho}{sup {minus}3.15}{beta}{sup 0.03}v{sup {minus}0.42}q{sub 95}{sup {minus}1.43} {proportional_to} I{sup 0.84}B{sup 0.39}n{sup 0.18}P{sup {minus}0.41}L{sup 2.0}, which is similar to empirical scalings derived from global confinement databases.
Date: December 1, 1998
Creator: Petty, C.C.; Luce, T.C. & Baity, F.W.
Partner: UNT Libraries Government Documents Department

Task III: UCSD/DIII-D/Textor FY-97-98 Accomplishments

Description: OAK (B204) Task III: UCSD/DIII-D/Textor FY-97-98 Accomplishments. A comprehensive report on the physics of pump limiters and particularly, the characterization of ALT-II, was published in Nuclear Fusion, bringing the project to a closure. The performance of the toroidal pump limiter was characterized under full auxiliary heating of 7 MW of NBI and ICRH and full pumping, as stated in the project milestones. Relevant highlights are: (1) Pumping with ALT-II allows for density control. (2) The achieved exhaust efficiency is 4% during NBI operation and near 2% during OH or ICRH operation. (3) We have shown that an exhaust efficiency of 2% is sufficient to satisfy the ash removal requirements of fusion reactors. (4) The plasma particle efflux and the pumped flux both increase with density and heating power. (5) The particle confinement time is less than the energy confinement time by a factor of 4. In summary, pumped belt limiters could provide the density control and ash exhaust requirements of fusion reactors.
Date: September 5, 2000
Creator: Boedo, J.A.
Partner: UNT Libraries Government Documents Department

Enhanced D-T supershot performance at high current using extensive lithium conditioning in TFTR

Description: A substantial improvement in supershot fusion plasma performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive Li conditioning of the TFTR limiter. This combination has resulted in not only significantly higher global energy confinement times than had previously been obtained in high current supershots, but also the highest ratio of central fusion output power to input power observed to date.
Date: May 1, 1995
Creator: Mansfield, D.K.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Bell, R.E. et al.
Partner: UNT Libraries Government Documents Department

Analysis of a dedicated rotation experiment in TFTR

Description: The results and analysis of a well-diagnosed, dedicated rotation experiment in TFTR are presented. Various neoclassical and anomalous theories for momentum transport are described and compared with the experimental data. The gyroviscocity theory is able to predict the measured central toroidal rotation speed, momentum confinement time and radial torque flow profile reasonably well when a poloidal asymmetry factor {tilde {Theta}} = 1.5 is used. The cold-ion-perpendicular-viscocity theory requires the assumption of an implausibly large number of cold ions in order to predict the magnitude of the experimental torque flow. The ion-temperature-gradient-mode theory, the untrapped-particle-electrostatic-mode theory and the stochastic-magnetic-perturbation theory all predict torque flows that differ greatly in magnitude, radial profile and parametric dependence from the experimental values.
Date: March 1, 1992
Creator: Stacey, W.M.
Partner: UNT Libraries Government Documents Department

Studies of energetic confined alphas using the pellet charge exchange diadgnostic on TFTR

Description: Results from recent DT experiments on TFIR to measure the energy distribution and radial density profile of fast confined alphas with the use of Li pellets and neutral particle analysis are presented. When a pellet is injected into the plasma, a toroidally extended ablation cloud is formed that travels with the pellet. A small fraction of the fusion alphas incident on the cloud are converted to helium neutrals as a result of electron capture processes. The escaping energetic helium neutrals are analyzed and detected by the neutral particle analyzer. Radially resolved energy spectra of trapped confined alphas in 0.5-2 MeV range and radial alpha density profiles are presented in this paper. The experimental data are compared with modeling results obtained with the TRANSP Monte-Carlo Code and with a specially developed Fokker-Planck Post Processor (FPP) that uses the alpha source distribution produced by TRANSP. Comparison of the experimental data with TRANSP and FPP show that the alphas in the plasma core of sawtooth free discharges in TFIR are well confined and slow down classically. The energy and radial profiles distributions outside the plasma core show the influence of stochastic ripple losses on alphas. Measurements for sawtoothing plasmas show a significant outward radial transport of trapped alphas.
Date: July 1995
Creator: Petrov, M. P.; Budny, R. V. & Duong, H. H.
Partner: UNT Libraries Government Documents Department

INVESTIGATION OF A PLASMA MODE IN EBTS.

Description: A plasma related mode has been identified when EBTS operated with long trap length. The mode frequency scaling showed monotonic increased with confinement time. Initial scaling qualitatively suggested the mode to an electron beam driven ion cyclotron instability. However, a more quantitative evaluation is indicative of a drift mode. Nevertheless, the possibility of a structure mode, though unlikely, can not be completely excluded. The process of proper instability identification and stabilization is described.
Date: November 6, 2000
Creator: HERSHCOVITCH,A.
Partner: UNT Libraries Government Documents Department