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Direct measurement of divertor exhaust neo enrichment in DIII-D

Description: We report first direct measurements of divertor exhaust gas impurity enrichment, {eta}{sub exh}=(exhaust impurity concentration){divided_by}(core impurity concentration), for both unpumped and D{sub 2} puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D{sub 2} puff into the SOL (scrape-off layer) with pumping increased {eta}{sub exh} threefold over either unpumped conditions or D{sub 2} puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment.
Date: June 1, 1996
Creator: Schaffer, M.J.; Wade, M.R.; Maingi, R.; Monier-Garbet, P.; West, W.P.; Whyte, D.G. et al.
Partner: UNT Libraries Government Documents Department


Description: Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density n{sub e,PED} and plasma stored energy W{sub T} were sensitive to changes in magnetic balance near the double-null (DN) configuration, e.g., both decreased 20%-30% when the DN shifted to a slightly unbalanced DN, where the B x {del}B drift direction pointed away from the main X-point. Recycling at each of the four divertor targets was sensitive to changes in magnetic balance and the B x {del}B drift direction. The poloidal distribution of the recycling in DN is in qualitative agreement with the predictions of UEDGE modeling with particle drifts included. The particle flux at the inner divertor target is shown to be much more sensitive to magnetic balance than the particle flux at the outer divertor target near the DN shape. These results suggest possible advantages and drawbacks for balanced DN operation.
Date: June 1, 2002
Partner: UNT Libraries Government Documents Department

Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

Description: Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.
Date: October 1996
Creator: Anderson, P. M.; Bozek, A. S.; Hollerbach, M. A.; Humphreys, D. A.; Luxon, J. L.; Reis, E. E. et al.
Partner: UNT Libraries Government Documents Department

Plasma flow in the DIII-D divertor

Description: Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor.
Date: July 1998
Creator: Boedo, J. A.; Porter, G. D. & Schaffer, M. J.
Partner: UNT Libraries Government Documents Department

Plasma pressure and flows during divertor detachment

Description: MHD theory applied to tokamak plasma scrape-off layer (SOL) equilibria requires Pfirsch-Schlueter current, which, because the magnetic lines are open, normally closes through electrically conducting divertor or limiter components. During detached divertor operation the Pfirsch-Schlueter current path to the divertor target is sometimes blocked, in which case theory predicts that the plasma develops a poloidal pressure gradient around the upstream SOL and a corresponding parallel flow, in order to satisfy all the conditions of MHD equilibrium. This paper reports the only known examples of detached diverted plasma in the DIII-D tokamak with blocked Pfirsch-Schlueter current, and they show no clear SOL poloidal pressure differences. However, the predicted pressure differences are small, near the limit of detectability with the available diagnostics. In the more usual DIII-D partially detached divertor operation mode, the Pfirsch-Schlueter current appears to never be blocked, and no unusual poloidal pressure differences are observed, as expected. Finally, a local overpressure is observed just inside the magnetic separatrix near the X-point in both attached and detached Ohmically heated plasmas.
Date: August 1, 1998
Creator: Schaffer, M.J.; Brooks, N.H.; Boedo, J.A.; Isler, R.C. & Moyer, R.A.
Partner: UNT Libraries Government Documents Department

Poloidal pressure gradients, divertor detachment and marfes

Description: Because the radiation power density from a marfe scales approximately as the square of its plasma pressure, and since increased radiation would aid divertor detachment for high power tokamaks, this paper identifies regions that might permit locally increased plasma pressure in steady state. The magnetic and dynamic (flow) constraints of magneto-hydrodynamics (MHD) are examined for self-consistent locally increased pressure equilibria, in both the magnetically open tokamak scrape-off layer (SOL) and the closed surfaces just inside the last closed flux surface. In most tokamak geometries it is difficult to recycle particles at a sufficient rate to sustain high pressure marfes, but they might be possible near a divertor X-point.
Date: November 1, 1997
Creator: Schaffer, M.J.
Partner: UNT Libraries Government Documents Department

Helical-D pinch

Description: A stabilized pinch configuration is described, consisting of a D-shaped plasma cross section wrapped tightly around a guiding axis. The {open_quotes}helical-D{close_quotes} geometry produces a very large axial (toroidal) transform of magnetic line direction that reverses the pitch of the magnetic lines without the need of azimuthal (poloidal) plasma current. Thus, there is no need of a {open_quotes}dynamo{close_quotes} process and its associated fluctuations. The resulting configuration has the high magnetic shear and pitch reversal of the reversed field pinch (RFP). (Pitch = P = qR, where R = major radius). A helical-D pinch might demonstrate good confinement at q << 1.
Date: August 1, 1997
Creator: Schaffer, M.J.
Partner: UNT Libraries Government Documents Department

Modeling of DIII-D noble gas puff and pump experiments

Description: Previous DIII-D experiments that induced a D{sup +} flow in the scrap-off layer (SOL) showed that this flow increased the divertor concentration of extrinsically injected impurities (neon, argon). These impurity fueling and exhaust (or puff and pump) experiments raise a number of modeling issues: the effect of edge-localized modes (ELMs) in regulating impurity core accumulation; the particle balance of the extrinsic impurities; the relation between divertor and plenum enrichment; and the effect of features unique to the present DIII-D Advanced Divertor configuration, specifically, the localized back-conductance of D{sub 2} and impurities from the baffle plenum in the outboard divertor region. To aid in understanding the relations between these processes, models have been improved: for core impurity transport to include ELM effects, and for divertor models to treat helium, neon, and argon transport with DIII-D--specific configuration effects. The models have been used to analyze a series of experiments in which neon and argon were first continuously injected (in the divertor private flux region) for 1.5 s, and then exhausted by the DIII-D cryopumping system. Deuterium was puffed at rates of 80 Torr L/s and 150 Torr L/s from the midplane and the divertor private region in these experiments. Results of the simulations are given.
Date: August 1, 1997
Creator: Hogan, J.T.; Wade, M.; Maingi, R.; Owen, L.; Schaffer, M. & West, P.
Partner: UNT Libraries Government Documents Department

Demonstration of the ITER Power Exhaust Solution Using the Puff and Pump Technique on DIII-D

Description: In future, high power density fusion devices, the need to prevent excessive local deposition of the plasma energy efflux on the first-wall surfaces is a critical design consideration in order to maintain the integrity of such surfaces. This requirement must be met without significant impact on plasma purity or overall plasma confinement. For the International Thermonuclear Experimental Reactor (ITER), these constraints have led to the following design criteria [1] P{sub rad}/(P{sub input} + P{sub {alpha}}) = 83%, P{sub rad,core}/(P{sub input} + P{sub {alpha}}) = 33%, P{sub target}/P{sub loss} = 17%, Z{sub eff} &lt; 1.8, and {tau}{sub E}/{tau}{sub E,ITER93H} &gt; 0.85. Here, P{sub loss} is the power flowing out of the core (i.e., P{sub input} + P{sub {alpha}} - P{sub rad,core})and P{sub target} is the power conducted to the target plate. These criteria represent a compromise between obtaining sufficient radiation to reduce the target heat load to a tolerable level, minimizing core fuel dilution, and maintaining sufficient power flow through the edge plasma to maintain H-mode confinement. Past experiments have had difficulty achieving these conditions simultaneously when using seeded impurities, and therefore there has been some concern regarding the viability of the ITER design. However, recent experiments in DIII-D using the puff and pump technique with argon as the seeded impurity have demonstrated the compatibility of these design constraints. In particular, steady-state plasma conditions have been achieved with P{sub rad}/P{sub input} = 72%, P{sub rad,core}/P{sub input} = 16%, P{sub target}/P{sub loss} = 17%, Z{sub eff} = 1.85, and {tau}{sub E}/{tau}{sub E,ITER93H} = 1.05.
Date: July 1, 1999
Creator: Wade, M.R.; West, W.P.; Hill, D.N.; Allen, S.L.; Boedo, J.A.; Brooks, N.H. et al.
Partner: UNT Libraries Government Documents Department

Measurements of flows in the DIII-D divertor by Mach probes

Description: First measurements of Mach number of background plasma in the DIII-D divertor are presented in conjunction with temperature T{sub e} and density n{sub e} using a fast scanning probe array. To validate the probe measurements, the authors compared the T{sub e}, n{sub e} and J{sub sat} data to Thomson scattering data and find good overall agreement in attached discharges and some discrepancy for T{sub e} and n{sub e} in detached discharges. The discrepancy is mostly due to the effect of large fluctuations present during detached plasmas on the probe characteristic; the particle flux is accurately measured in every case. A composite 2-D map of measured flows is presented for an ELMing H-mode discharge and they focus on some of the details. They have also documented the temperature, density and Mach number in the private flux region of the divertor and the vicinity of the X-point, which are important transition regions that have been little studied or modeled. Background parallel plasma flows and electric fields in the divertor region show a complex structure.
Date: June 1, 1998
Creator: Boedo, J.A.; Lehmer, R.; Moyer, R.A.; Watkins, J.G.; Porter, G.D.; Evans, T.E. et al.
Partner: UNT Libraries Government Documents Department

Divertor E X B Plasma Convection in DIII-D

Description: Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for standard (ion VB{sub T} drift toward divertor X-point) and reversed B{sub T} directions; for low (L) and high (H) confinement modes; and for partially detached divertor mode. The data are consistent with recent computational modeling identifying E x B{sub T} circulation, due to potentials sustained by plasma gradients, as the main cause of divertor plasma sensitivity to B{sub T} direction.
Date: July 1, 1999
Creator: Boedo, J.A.; Schaffer, M.J.; Maingi, M.; Lasnier, C.J. & Watkins, J.G.
Partner: UNT Libraries Government Documents Department


Description: Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments and resistive wall mode experiments done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported. Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments. These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.
Date: February 1, 2003
Partner: UNT Libraries Government Documents Department


Description: Intermittent plasma objects (IPOs) featuring higher pressure than the surrounding plasma, and responsible for {approx}50% of the E x B{sub T} radial transport, are observed in the scrape-off layer (SOL) and edge of the DIII-D tokamak. The skewness of probe and BES intermittent data suggest IPO formation at or near the last closed flux surface (LCFS) and the existence of hole-IPO pairs. The particle content of the IPOs at the LCFS is linearly dependent on the discharge density, however, when normalized to the local averaged density, it is fairly insensitive to density variations. It is also shown that the IPOs thermalize with the background plasma within 1 cm of the LCFS. The IPOs appear in the SOL of both L and H mode discharges carrying {approx}50% of the total SOL radial E x B{sub T} transport at all radii. However, the total flux and the IPO contribution, are highly reduced in H-mode conditions due to the increased confinement.
Date: June 1, 2002
Partner: UNT Libraries Government Documents Department


Description: Detailed measurements in two dimensions by probes and Thomson scattering reveal unexpected local electric potential and electron pressure (p{sub e}) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak [J.L. Luxon and L.G. Davis, Fusion Technol. 8, 441 (1985)]. The potential drives E x B circulation about the X-point, thereby exchanging plasma between closed and open magnetic surfaces at rates that can be comparable to the total cross-separatrix transport. The potential is consistent with the classical parallel Ohm's law. A simple model is proposed to explain the pressure and potential hills in low power, nearly detached plasmas. Recent two-dimensional edge transport modeling with plasma drifts also shows X-point pressure and potential hills but by a different mechanism. These experimental and theoretical results demonstrate that low power tokamak plasmas can be far from poloidal uniformity in a boundary layer just inside the separatrix. Additional data, though preliminary and incomplete, suggest that E x B circulation across the separatrix might be a common feature of low confinement behavior.
Date: November 1, 2000
Partner: UNT Libraries Government Documents Department

Divertor particle exhaust and wall inventory on DIII-D

Description: Many tokamaks achieve optimum plasma performance by achieving low recycling; various wall conditioning techniques including helium glow discharge cleaning (HeGDC) are routinely applied to help achieve low recycling. Many of these techniques allow strong, transient wall pumping, but they may not be effective for long-pulse tokamaks, such as the International Thermonuclear Experimental Reactor (ITER), the Tokamak Physics Experiment (TPX), Tore Supra Continu, and JT-60SU. Continuous particle exhaust using an in-situ pumping scheme may be effective for wall inventory control in such devices. Recent particle balance experiments on the Tore Supra and DIII-D tokamaks demonstrated that the wall particle inventory could be reduced during a given discharge by use of continuous particle exhaust. In this paper the authors report the first results of wall inventory control and good performance with the in-situ DIII-D cryopump, replacing the HeGDC normally applied between discharges.
Date: September 1, 1995
Creator: Maingi, R.; Jackson, G.L.; Mahdavi, M.A.; Schaffer, M.J.; Wade, M.R.; Mioduszewski, P.K. et al.
Partner: UNT Libraries Government Documents Department

Edge Stability and Transport Control with Resonant Magnetic Perturbations in Collisionless Tokamak Plasmas

Description: A critical issue for fusion plasma research is the erosion of the first wall of the experimental device due to impulsive heating from repetitive edge magneto-hydrodynamic (MHD) instabilities known as 'edge-localized modes' (ELMs). Here, we show that the addition of small resonant magnetic field perturbations completely eliminates ELMs while maintaining a steady-state high-confinement (H-mode) plasma. These perturbations induce a chaotic behavior in the magnetic field lines, which reduces the edge pressure gradient below the ELM instability threshold. The pressure gradient reduction results from a reduction in particle content of the plasma, rather than an increase in the electron thermal transport. This is inconsistent with the predictions of stochastic electron heat transport theory. These results provide a first experimental test of stochastic transport theory in a highly rotating, hot, collisionless plasma and demonstrate a promising solution to the critical issue of controlling edge instabilities in fusion plasma devices.
Date: June 13, 2006
Creator: Evans, T E; Moyer, R A; Burrell, K H; Fenstermacher, M E; Joseph, I; Leonard, A W et al.
Partner: UNT Libraries Government Documents Department

Comparison of Edge Plasma Perturbation During ELM Control Using One vs Two Toroidal Rows of RMP Coils in ITER Similar Shaped Plasmas on DIII-D

Description: Large Type-I edge localized modes (ELMs) were suppressed by n = 3 resonant magnetic perturbations (RMPs) from a set of internal coils (I-coil) in plasmas with an ITER similar shape at the ITER pedestal collisionality, {nu}*{sub e} {approx} 0.1 and low edge safety factor (q{sub 95} {approx} 3.6), with either a single toroidal row of the internal RMP coils or two poloidally separated rows of coils. ELM suppression with a single row of internal coils was achieved at approximately the same q{sub 95} surface-averaged perturbation field as with two rows of coils, but required higher current per coil. Maintaining complete suppression of ELMs using n = 3 RMPs from a single toroidal row of internal coils was less robust to variations in input neutral beam injection torque than previous ELM suppression cases using both rows of internal coils. With either configuration of RMP coils, maximum ELM size is correlated with the width of the edge region having good overlap of the magnetic islands from vacuum field calculations.
Date: May 21, 2008
Creator: Fenstermacher, M E; Evans, T E; Osborne, T H; Schaffer, M J; deGrassie, J S; Gohil, P et al.
Partner: UNT Libraries Government Documents Department

A Comparison of Radiating Divertor Behavior in Single- and Double-Null Plasmas in DIII-D

Description: '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 {tau}{sub E}. 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 Bx{del}B-ion drift direction was toward the divertor (V{sub {del}B{up_arrow}}) rather than away from the divertor (V{sub {del}B{down_arrow}}). 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 Bx{del}B-ion drift direction. In DNs, the radiated power was not shared equally between divertors during argon injection. Only in the divertor opposite Bx{del}B ion drift direction were both significant increases in divertor radiated power and an accumulation of argon, based on spectroscopic measurements of ArII, observed. Our data suggests that a DN shape that is biased in the direction away from the Bx{del}B-ion drift direction may provide the best prospect of successfully coupling a radiating divertor approach with a higher performance H-mode plasma.
Date: March 25, 2008
Creator: Petrie, T W; Brooks, N H; Fenstermacher, M E; Groth, M; Hyatt, A W; Isler, R C et al.
Partner: UNT Libraries Government Documents Department

Effect of Island Overlap on ELM Suppression by Resonant Magnetic Perturbations in DIII-D

Description: Recent DIII-D [J.L. Luxon, et al., Nucl. Fusion 43, 1813 (2003)] experiments show a correlation between the extent of overlap of magnetic islands induced in the edge plasma by perturbation coils and complete suppression of Type-I edge localized modes (ELMs) in plasmas with ITER-like electron pedestal collisionality {nu}*{sub e} {approx} 0.1, flux surface shape and low edge safety factor (q{sub 95} {approx} 3.6). With fixed n = 3 resonant magnetic perturbation (RMP) strength, ELM suppression is obtained only in a finite window in the edge safety factor (q{sub 95}) consistent with maximizing the resonant component of the applied helical field. ELM suppression is obtained over an increasing range of q{sub 95} by either increasing the n = 3 RMP strength, or by adding n = 1 perturbations to 'fill in' gaps between islands across the edge plasma. The suppression of Type-I ELMs correlates with a minimum width of the edge region having magnetic islands with Chirikov parameter &gt;1.0, based on vacuum calculations of RMP mode components excluding the plasma response or rotational shielding. The fraction of vacuum magnetic field lines that are lost from the plasma, with connection length to the divertor targets comparable to an electron-ion collisional mean free path, increases throughout the island overlap region in the ELM suppressed case compared with the ELMing case.
Date: November 8, 2007
Creator: Fenstermacher, M E; Evans, T E; Osborne, T H; Schaffer, M J; Aldan, M P; deGrassie, J S et al.
Partner: UNT Libraries Government Documents Department

Engineering features of ISX

Description: ISX, an Impurity Study Experiment, is presently being designed at Oak Ridge National Laboratory as a joint scientific effort between ORNL and General Atomic Company. ISX is a moderate size tokamak dedicated to the study of impurity production, diffusion, and control. The significant engineering features of this device are discussed. (auth)
Date: January 1, 1975
Creator: Lousteau, D.C.; Jernigan, T.C.; Schaffer, M.J. & Hussung, R.O.
Partner: UNT Libraries Government Documents Department