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M Groth,
G D Porter,
T D Rognlien,
S Wiesen,
M Wischmeier,
M N A Beurskens,
X Bonnin,
B D Bray,
S Brezinsek,
N H Brooks, [......],
A W Leonard,
A Meigs,
H W Müller,
M E Rensink,
D L Rudakov,
J G Watkins,
E Wolfrum,
the DIII-D,
ASDEX Upgrade Teams,
JET EFDA Contributors
[show abstract]
[hide abstract]
ABSTRACT: Deuterium fueling profiles across the separatrix have been calculated with the edge fluid codes UEDGE, SOLPS and EDGE2D/EIRENE for lower single null, ohmic and low-confinement plasmas in DIII-D, ASDEX Upgrade and JET. The fueling profiles generally peak near the divertor x-point, and broader profiles are predicted for the open divertor geometry and horizontal targets in DIII-D than for the more closed geometries and vertical targets in AUG and JET. Significant fueling from the low-field side midplane may also occur when assuming strong radial ion transport in the far scrape-off layer. The dependence of the fueling profiles on upstream density is investigated for all three devices, and between the different codes for a single device. The validity of the predictions is assessed for the DIII-D configuration by comparing the measured ion current to the main chamber walls at the low-field side and divertor targets, and deuterium emission profiles across the divertor legs, and the high-field and low-field side midplane regions to those calculated by UEDGE and SOLPS.
Plasma Physics and Controlled Fusion 11/2011; 53(12):124017. · 2.42 Impact Factor
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T.W. Petrie,
T.E. Evans,
N.H. Brooks,
M.E. Fenstermacher,
J.R. Ferron,
C.T. Holcomb,
B. Hudson,
A.W. Hyatt,
T.C. Luce, C.J. Lasnier,
S. Mordijck,
R.A. Moyer,
T.H. Osborne,
P.A. Politzer,
M.E. Rensink,
M.J. Schaffer,
P.B. Snyder,
J.G. Watkins
[show abstract]
[hide abstract]
ABSTRACT: The range in density and collisionality for which resonant magnetic perturbations (RMPs) are effective in suppressing edge-localized modes (ELMs) in the presence of a radiating divertor was found to be modest for representative H-mode plasmas in DIII-D. When deuterium and argon gas injection rates were increased during RMP, both the electron collisionality in the pedestal and the maximum electron pressure gradient (∇Pe,MAX) in the pedestal also increased. As ∇Pe,MAX approached values consistent with the peeling–ballooning stability limit, as determined by edge stability analysis, ELMing activity re-emerged. For cases with the same injected neutral beam power, argon accumulation in the main plasma was greater in the RMP ELM-suppressed cases than in comparable non-RMP ELMing H-mode cases. Reductions in the core concentration of injected argon were observed for both RMP and non-RMP H-mode cases when their respective deuterium injection rates were increased. Although complete ELM suppression in RMP radiating divertor plasmas in DIII-D was only accessible over a limited range in pedestal density and collisionality, significant ELM mitigation with heat flux reduction was possible over a wider range. Comparing RMP radiating divertor discharges after the re-appearance of ELMing activity during gas puffing with a standard ELMing plasma for cases with the same pedestal density reveals that the RMP discharges have (1) lower average electron temperature at the midplane separatrix, implying lower average electron temperature at the divertor target, (2) lower time-averaged peak heat flux and (3) lower transient peak heat flux from ELMs even at the same pedestal collisionality.
Nuclear Fusion 05/2011; 51(7):073003. · 4.09 Impact Factor
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J. A. Boedo,
E. A. Belli,
E. Hollmann,
W. M. Solomon,
D. L. Rudakov,
J. G. Watkins,
R. Prater,
J. Candy,
R. J. Groebner,
K. H. Burrell,
J. S. deGrassie, C. J. Lasnier,
A. W. Leonard,
R. A. Moyer,
G. D. Porter,
N. H. Brooks,
S. Muller,
G. Tynan,
E. A. Unterberg
[show abstract]
[hide abstract]
ABSTRACT: First measurements of the D+ parallel velocity, V∥D+, in L-mode discharges in the DIII-D [
J. L. Luxon, Nucl. Fusion 42, 614 (2002)
] tokamak boundary region at two poloidal locations, θ ∼ 0° and θ ∼ 255°, made using Mach probes, feature a peak with velocities of up to 80 km/s at the midplane last closed flux surface (LCFS), as high as ten times the charge exchange recombination C6+ toroidal velocity, VϕC6+, in the same location. The V∥D+ profiles are very asymmetric poloidally, by a factor of 8–10, and feature a local peak at the midplane. This peak, 1–2 cm wide, is located at or just inside the LCFS, and it suggests a large source of momentum in that location. This momentum source is quantified at ∼ 0.31 N m by using a simple momentum transport model. This is the most accurate measurement of the effects of so called “intrinsic” edge momentum source to date. The V∥D+ measurements are quantitatively consistent with a purely neoclassical computational modeling of V∥D+ by the code NEO [
E. A. Belli and J. Candy, Plasma Phys. Controlled Fusion 50, 095010 (2008)
], using VϕC6+ as input, for ρ ∼ 0.7–0.95 at the two poloidal locations, where V∥D+ measurements exist. The midplane NEO-calculated V∥D+ grows larger than V∥C6+ in the steeper edge gradient region and trends to agreement with the probe-measured V∥D+ data near ρ ∼ 1, where the local V∥D+ velocity peak exists. The measurements and computations were made in OH and L-mode discharges on an upper single null, with ion ∇BT drift away from the divertor. The rotating layer finding is similar in auxiliary heated discharges with and without external momentum input, except that at higher density the edge velocity weakens.
Physics of Plasmas 03/2011; 18(3):032510-032510-9. · 2.15 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: An experiment aimed at benchmarking the ITER scrape-off layer (SOL) power width scaling in limited L-mode discharges has been conducted on DIII-D. Scans of the main scaling parameters were performed in an inner-wall-limited (IWL) magnetic configuration. Using the near-SOL density and temperature e-folding lengths, lambda(n), lambda(T), determined from reciprocating Langmuir probe measurements, SOL power flux density e-folding lengths, lambda(q), are derived. A few lower single null (LSN) discharges were also run for comparison. The results are generally in agreement with the ITER design assumptions, finding that lambda(n) and lambda(T) are correlated (lambda(T) similar to 1.2 lambda(n)) and both lambda(n) and lambda(T) are on average 2.1-2.5 times larger in IWL configurations than in LSN. In moderate elongation (kappa similar to 1.4) IWL discharges, lambda(q) is largest and agrees with the assumed ITER scaling within the estimated uncertainty (a factor of similar to 2). In IWL discharges lambda(q) measurements are consistent with the expectations of SOL power balance. (C) 2010 Elsevier B.V. All rights reserved.
Journal of Nuclear Materials. 01/2011; 415(1):S387-S390.
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R.J. Groebner,
P.B. Snyder,
T.H. Osborne,
A.W. Leonard,
T.L. Rhodes,
L. Zeng,
E.A. Unterberg,
Z. Yan,
G.R. McKee, C.J. Lasnier,
J.A. Boedo,
J.G. Watkins
[show abstract]
[hide abstract]
ABSTRACT: The spatial and temporal evolution of the total pedestal pressure profile has been measured during the pedestal evolution between successive edge localized modes (ELMs) of type-I ELMing H-mode discharges in DIII-D. Measurements are used to test a model that predicts that kinetic ballooning modes (KBMs) provide a strong constraint on the pedestal pressure gradient obtained during an inter-ELM cycle and cause the pedestal width to scale as the square root of the pedestal poloidal beta. Discharges in two different parameter regimes are examined for evidence that the evolution of the pressure gradient reaches a limit prior to the onset of an ELM. Both discharges show evidence of rapid evolution of the pressure profile very early in the recovery phase from an ELM. In one discharge, the pressure gradient reached approximate steady state within ~3 ms after the ELM event. In the other discharge, the pressure gradient just inboard of the last closed flux surface reached steady state early in the ELM recovery phase even as the pedestal expanded into the core and the maximum pressure gradient continued to rise during the remainder of the ELM cycle. Simple quantitative theoretical metrics show that pressure gradients in both discharges reached levels that were large enough to excite KBMs. In addition, the peeling–ballooning theory for the onset of type-I ELMs and the EPED1 model for pedestal height and width make predictions consistent with the data of both discharges.
Nuclear Fusion 05/2010; 50(6):064002. · 4.09 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: Edge localized modes (ELMs) are qualitatively and quantitatively modeled in tokamaks using current bursts which have been observed in the scrape-off-layer (SOL) during an ELM crash. During the initial phase of an ELM, a heat pulse causes thermoelectric currents. They first flow in short connection length flux tubes which are initially established by error fields or other nonaxisymmetric magnetic perturbations. The currents change the magnetic field topology in such a way that larger areas of short connection length flux tubes emerge. Then currents predominantly flow in short SOL-like flux tubes and scale with the area of the flux tube assuming a constant current density. Quantitative predictions of flux tube patterns for a given current are in excellent agreement with measurements of the heat load and current flow at the DIII-D target plates during an ELM cycle.
Physical Review Letters 04/2010; 104(17):175001. · 7.37 Impact Factor
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D L Rudakov,
C P C Wong,
A Litnovsky,
W R Wampler,
J A Boedo,
N H Brooks,
M E Fenstermacher,
M Groth,
E M Hollmann,
W Jacob, [......],
A G McLean,
M Marot,
R A Moyer,
T W Petrie,
V Philipps,
R D Smirnov,
P C Stangeby,
J G Watkins,
W P West,
J H Yu
[show abstract]
[hide abstract]
ABSTRACT: Divertor and midplane material evaluation systems (DiMES and MiMES) in the DIII-D tokamak are used to address a variety of plasma–material interaction (PMI) issues relevant to ITER. Among the topics studied are carbon erosion and re-deposition, hydrogenic retention in the gaps between plasma-facing components (PFCs), deterioration of diagnostic mirrors from carbon deposition and techniques to mitigate that deposition, and dynamics and transport of dust. An overview of the recent experimental results is presented.
Physica Scripta 12/2009; 2009(T138):014007. · 1.20 Impact Factor
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D.L. Rudakov,
A. Litnovsky,
W.P. West,
J.H. Yu,
J.A. Boedo,
B.D. Bray,
S. Brezinsek,
N.H. Brooks,
M.E. Fenstermacher,
M. Groth, [......],
A.G. McLean,
R.A. Moyer,
A.Yu. Pigarov,
V. Philipps,
A. Pospieszczyk,
R.D. Smirnov,
J.P. Sharpe,
W.M. Solomon,
J.G. Watkins,
C.P.C. Wong
[show abstract]
[hide abstract]
ABSTRACT: Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicrometre sized dust is routinely observed using Mie scattering from a Nd : Yag laser. The source is strongly correlated with the presence of type I edge localized modes (ELMs). Larger size (0.005–1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust; on the other hand, large flakes or debris falling into the plasma may induce a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micrometre-size particles into plasma discharges. In DIII-D, a sample holder filled with 30–40 mg of dust is inserted in the lower divertor and exposed, via sweeping of the strike points, to the diverted plasma flux of high-power ELMing H-mode discharges. After a brief dwell (~0.1 s) of the outer strike point on the sample holder, part of the dust penetrates into the core plasma, raising the core carbon density by a factor of 2–3 and resulting in a twofold increase in the radiated power. In TEXTOR, instrumented dust holders with 1–45 mg of dust are exposed in the scrape-off-layer 0–2 cm radially outside of the last closed flux surface in discharges heated with 1.4 MW of NBI. Launched in this configuration, the dust perturbed the edge plasma, as evidenced by a moderate increase in the edge carbon content, but did not penetrate into the core plasma.
Nuclear Fusion 07/2009; 49(8):085022. · 4.09 Impact Factor
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T.W. Petrie,
G.D. Porter,
N.H. Brooks,
M.E. Fenstermacher,
J.R. Ferron,
M. Groth,
A.W. Hyatt,
R.J. La Haye, C.J. Lasnier,
A.W. Leonard,
T.C. Luce,
P.A. Politzer,
M.E. Rensink,
M.J. Schaffer,
M.R. Wade,
J.G. Watkins,
W.P. West
[show abstract]
[hide abstract]
ABSTRACT: The effectiveness of the puff-and-pump technique to enrich a seeded impurity in the divertor relative to the core and, thereby, to maximize radiation in the divertor depends sensitively on both the magnetic geometry and the ion B × ∇B drift direction. In the puff-and-pump scenario used here, argon impurities injected into the private flux region are inhibited from accumulation in the core plasma by enhanced plasma flows to the divertor created by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping near the divertor targets. Modelling of single-null, H-mode plasmas with the UEDGE fluid transport code indicates that particle drifts in the scrape-off layer and divertor strongly affect the locations where the argon seed impurity accumulates. It is also found in double-null cases that argon always shows a larger accumulation in the divertor out of which the ion B × ∇B drift is directed, regardless of the divertor into which the argon is injected. Experiments have shown that the degree to which the deuterium gas-puffing rate inhibits the escape of the seed impurity from the divertor(s) depends critically on the direction of the ion B × ∇B drift and on whether the plasma is single-null or double-null. The transition in behaviour from double-null to single-null character during puff-and-pump occurs for |dRsep| = 0.4 cm when the ion B × ∇B drift was pointing away from the dominant divertor. The lowest argon density buildup in the main plasma of any of the configurations studied during puff-and-pump was achieved in single-null plasmas with the ion B × ∇B drift direction away from the divertor.
Nuclear Fusion 05/2009; 49(6):065013. · 4.09 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: A video camera system is described as that measures the spatial distribution of visible line emission emitted from the main scrape-off layer (SOL) of plasmas in the DIII-D tokamak. A wide-angle lens installed on an equatorial port and an in-vessel mirror, which intercepts part of the lens' view, provide simultaneous tangential views of the SOL on the low-field and high-field sides of the plasma's equatorial plane. Tomographic reconstruction techniques are used to calculate the two-dimensional (2D) poloidal profiles from the raw data, and one-dimensional (1D) poloidal profiles simulating chordal views of other optical diagnostics from the 2D profiles. The 2D profiles can be compared with SOL plasma simulations; the 1D profiles with measurements from spectroscopic diagnostics. Sample results are presented, which elucidate carbon transport in plasmas with toroidally uniform injection of methane and argon transport in disruption mitigation experiments with massive gas jet injection.
The Review of scientific instruments 04/2009; 80(3):033505. · 1.52 Impact Factor
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D. L. Rudakov,
J. A. Boedo,
J. H. Yu,
N. H. Brooks,
M. E. Fenstermacher,
M. Groth,
E. M. Hollmann, C. J. Lasnier,
A. G. McLean,
R. A. Moyer,
P. C. Stangeby,
G. R. Tynan,
W. R. Wampler,
J. G. Watkins,
W. P. West,
C. P. C. Wong,
R. J. Bastasz,
D. Buchenauer,
J. Whaley
[show abstract]
[hide abstract]
ABSTRACT: Erosion of the main chamber plasma-facing components is of concern for ITER. Plasma interaction with the outboard chamber wall is studied in DIII-D using Langmuir probes and optical diagnostics. Fast camera data shows that edge localized modes (ELMs) feature helical filamentary structures propagating towards the outboard wall. Upon reaching the wall, filaments result in regions of local intense plasma-material interaction (PMI) where peak incident particle and heat fluxes are up to two orders of magnitude higher than those between ELMs. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. A moderate change of the gap between the separatrix and the outer wall strongly affects PMI intensity at the wall. Material samples exposed near the outboard wall showed net carbon deposition in high-density discharges (near the Greenwald limit) and tendency towards net erosion in lower density discharges (similar to 0.45 of the Greenwald limit). (C) 2009 Elsevier B.V. All rights reserved.
Journal of Nuclear Materials. 01/2009; 390-91:785-788.
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D L Rudakov,
J H Yu,
J A Boedo,
E M Hollmann,
S I Krasheninnikov,
R A Moyer,
S H Muller,
A Yu Pigarov,
M Rosenberg,
R D Smirnov, [......],
C P C Wong,
A L Roquemore,
C H Skinner,
W M Solomon,
S Ratynskaia,
M E Fenstermacher,
M Groth, C J Lasnier,
A G McLean,
P C Stangeby
[show abstract]
[hide abstract]
ABSTRACT: Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 microm in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.
The Review of scientific instruments 11/2008; 79(10):10F303. · 1.52 Impact Factor
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T.W. Petrie,
N.H. Brooks,
M.E. Fenstermacher,
M. Groth,
A.W. Hyatt,
R.C. Isler, C.J. Lasnier,
A.W. Leonard,
G.D. Porter,
M.J. Schaffer,
J.G. Watkins,
M.R. Wade,
W.P. West
[show abstract]
[hide abstract]
ABSTRACT: 'Puff-and-pump' radiating divertor scenarios, applied to both upper single-null (SN) and double-null (DN) H-mode plasmas, result in a 30–60% increase in radiated power with little or no decrease in τE. Argon was injected into the private flux region of the upper divertor, and plasma flow into the upper divertor was enhanced by a combination of deuterium gas puffing upstream of the divertor targets and particle pumping at the targets. For the same constant deuterium injection rate, argon penetrated the main plasma of SNs more rapidly and reached a higher steady-state concentration when the B × ∇B-ion drift direction was towards the divertor (V∇B↑) rather than away from the divertor (V∇B↓). We also found that the initial rate at which argon accumulated inside DN plasmas was more than twice that of comparable SN plasmas having the same B × ∇B-ion drift direction. In DNs, the radiated power was not shared equally between divertors during argon injection. Only when the B × ∇B ion drift direction was away from the divertor were both significant increases in divertor radiated power and an accumulation of argon in the divertor observed, based on spectroscopic measurements of Ar II. Our data suggest that an unbalanced DN shape where the B × ∇B-ion drift is directed away from the dominant divertor may provide the best chance of successfully coupling a radiating divertor approach with a higher performance H-mode plasma.
Nuclear Fusion 03/2008; 48(4):045010. · 4.09 Impact Factor
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T.E. Evans,
M.E. Fenstermacher,
R.A. Moyer,
T.H. Osborne,
J.G. Watkins,
P. Gohil,
I. Joseph,
M.J. Schaffer,
L.R. Baylor,
M. Bécoulet, [......],
M.W. Jakubowski, C.J. Lasnier,
M. Lehnen,
A.W. Leonard,
J. Lonnroth,
E. Nardon,
V. Parail,
O. Schmitz,
B. Unterberg,
W.P. West
[show abstract]
[hide abstract]
ABSTRACT: Large Type-I edge localized modes (ELMs) are completely eliminated with small n = 3 resonant magnetic perturbations (RMP) in low average triangularity, , plasmas and in ITER similar shaped (ISS) plasmas, , with ITER relevant collisionalities . Significant differences in the RMP requirements and in the properties of the ELM suppressed plasmas are found when comparing the two triangularities. In ISS plasmas, the current required to suppress ELMs is approximately 25% higher than in low average triangularity plasmas. It is also found that the width of the resonant q95 window required for ELM suppression is smaller in ISS plasmas than in low average triangularity plasmas. An analysis of the positions and widths of resonant magnetic islands across the pedestal region, in the absence of resonant field screening or a self-consistent plasma response, indicates that differences in the shape of the q profile may explain the need for higher RMP coil currents during ELM suppression in ISS plasmas. Changes in the pedestal profiles are compared for each plasma shape as well as with changes in the injected neutral beam power and the RMP amplitude. Implications of these results are discussed in terms of requirements for optimal ELM control coil designs and for establishing the physics basis needed in order to scale this approach to future burning plasma devices such as ITER.
Nuclear Fusion 01/2008; 48(2):024002. · 4.09 Impact Factor
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I Joseph,
T E Evans,
A M Runov,
M E Fenstermacher,
M Groth,
S V Kasilov, C J Lasnier,
R A Moyer,
G D Porter,
M J Schaffer,
R Schneider,
J G Watkins
[show abstract]
[hide abstract]
ABSTRACT: The effect of resonant magnetic perturbations on heat transport in DIII-D H-mode plasmas has been calculated by combining the TRIP3D field line tracing code with the E3D two-fluid transport code. Simulations show that the divertor heat flux distribution becomes non-axisymmetric because heat flux is efficiently guided to the divertor along the three-dimensional invariant manifolds of the magnetic field. Calculations demonstrate that heat flux is spread over a wider area of the divertor target, thereby reducing the peak heat flux delivered during steady-state operation. Filtered optical cameras have observed non-axisymmetric particle fluxes at the strike point and Langmuir probes have observed non-axisymmetric floating potentials. On the other hand, the predicted magnitude of stochastic thermal transport is too large to match the pedestal plasma profiles measured by Thomson scattering and charge exchange recombination spectroscopy. The Braginskii thermal conductivity overestimates the experimental heat transport in the pedestal because the mean free paths of both species are longer than estimates of the parallel thermal correlation lengths, and collisionless transport models are probably required for accurate description. However, even the collisionless estimates for electron thermal transport are too large by one to two orders of magnitude. Thus, it is likely that another mechanism such as rotational screening of resonant perturbations limits the stochastic region and reduces transport inside of the pedestal.
Nucl. Fusion. 01/2008; 48.
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T W Petrie,
N H Brooks,
M E Fenstermacher,
J R Ferron,
M Groth,
A W Hytatt,
R J La Haye, C J Lasnier,
A W Leonard,
T C Luce,
P A Politzer,
G D Porter,
M J Schaffer
01/2008;
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A. Litnovsky,
D.L. Rudakov,
G. De Temmerman,
P. Wienhold,
V. Philipps,
U. Samm,
A.G. McLean,
W.P. West,
C.P.C. Wong,
N.H. Brooks, [......],
M.E. Fenstermacher,
M. Groth, C.J. Lasnier,
R.L. Boivin,
A.W. Leonard,
A. Romanyuk,
T. Hirai,
G. Pintsuk,
U. Breuer,
A. Scholl
Fusion Engineering and Design 01/2008; 83(1):79-89. · 1.49 Impact Factor
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R.S. Granetz,
E.M. Hollmann,
D.G. Whyte,
V.A. Izzo,
G.Y. Antar,
A. Bader,
M. Bakhtiari,
T. Biewer,
J.A. Boedo,
T.E. Evans, [......],
R.A. Moyer,
P.B. Parks,
M.L. Reinke,
D.L. Rudakov,
E.J. Strait,
J.L. Terry,
J. Wesley,
W.P. West,
G. Wurden,
J. Yu
[show abstract]
[hide abstract]
ABSTRACT: High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 × or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER.
Nuclear Fusion 08/2007; 47(9):1086. · 4.09 Impact Factor
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[show abstract]
[hide abstract]
ABSTRACT: In recent campaigns, the photron ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the national spherical torus experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE localized modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX small type V ELM regime[2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera[4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.
Fusion Engineering, 2007. SOFE 2007. 2007 IEEE 22nd Symposium on; 07/2007
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D L Rudakov,
W P West,
M Groth,
J H Yu,
C C Wong,
J A Boedo,
N H Brooks,
T E Evans,
M E Fenstermacher,
E M Hollmann,
A W Hyatt, C J Lasnier,
A G McLean,
R A Moyer,
A Pigarov,
R Smirnov,
W M Solomon,
J G Watkins
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ABSTRACT: Dust is commonly found in fusion devices. Though generally of no concern in the present day machines, dust may pose serious safety and operational concerns for ITER. Micron-size dust usually dominates the samples collected from tokamaks. During a plasma discharge micron-size dust particles can become highly mobile and travel over distances of a few meters. Once inside the plasma, dust particles heat up to over 3000 K and emit thermal radiation that can be detected by visible imaging techniques. Observations of naturally occurring and artificially introduced dusts have been performed in DIII-D divertor and scrape-off layer (SOL) using standard frame rate CMOS cameras, a gated-intensified CID camera, and a fast-framing CMOS camera. In the first 2-3 plasma discharges after a vent with personnel entry inside the vacuum vessel ('dirty vent') dust levels were quite high with thousands of particles observed in each discharge. Individual particles moving at velocities of up to a few hundred m/s and breakup of larger particles into pieces were observed. After about 15 discharges dust was virtually gone during the stationary portion of a discharge, and appeared at much reduced levels during the plasma initiation and termination phases. After a few days of plasma operations (about 70 discharges) dust levels were further reduced to just a few observed events per discharge except in discharges with current disruptions that produced significant amounts of dust. An injection of a few milligram of micron-size (6 micron median diameter) carbon dust into a high-power lower single-null ELMing H-mode discharge with strike points swept across the lower divertor floor was performed. A significant increase of the core carbon radiation was observed for about 250 ms after the injection, as the total radiated power increased twofold. Dust particles from the injection were observed by the fast framing camera in the outboard SOL near the midplane. The amount of dust observed by the fast camera immediately after the injection was
04/2007
