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Investigation of density limit processes in DIII-D

A series of experiments has been conducted in DIII-D to investigate density-limiting processes. The authors have studied divertor detachment and MARFEs on closed field lines and find semi-quantitative agreement with theoretical calculations of onset conditions. They have shown that the critical density for MARFE onset at low edge temperature scales as I{sub p}/a{sup 2}, i.e. similar to Greenwald scaling. They have also shown that the scaling of the critical separatrix density with heating power at partial detachment onset agrees with Borass` model. Both of these processes yield high edge density limits for reactors such as ITER. By using divertor pumping and pellet fueling they have avoided these and other processes and accessed densities > 1.5{times} Greenwald limit scaling with H-mode confinement, demonstrating that the Greenwald limit is not a fundamental limit on the core density.

Date: February 1, 1999

Creator: Maingi, R.; Mahdavi, M.A. & Petrie, T.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Divertor plasma physics experiments on the DIII-D tokamak

In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model.

Date: October 1996

Creator: Mahdavi, M.A.; Allen, S.L. & Evans, T.E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Investigation of physical processes limiting plasma density in H-mode on DIII-D

A series of experiments was conducted on the DIII-D tokamak to investigate the physical processes which limit density in high confinement mode (H-mode) discharges. The typical H-mode to low confinement mode (L-mode) transition limit at high density near the empirical Greenwald density limit was avoided by divertor pumping, which reduced divertor neutral pressure and prevented formation of a high density, intense radiation zone (MARFE) near the X-point. It was determined that the density decay time after pellet injection was independent of density relative to the Greenwald limit and increased non-linearly with the plasma current. Magnetohydrodynamic (MHD) activity in pellet-fueled plasmas was observed at all power levels, and often caused unacceptable confinement degradation, except when the neutral beam injected (NBI) power was {le} 3 MW. Formation of MARFEs on closed field lines was avoided with low safety factor (q) operation but was observed at high q, qualitatively consistent with theory. By using pellet fueling and optimizing discharge parameters to avoid each of these limits, an operational space was accessed in which density {approximately} 1.5 {times} Greenwald limit was achieved for 600 ms, and good H-mode confinement was maintained for 300 ms of the density flattop. More significantly, the density was successfully increased to the limit where a central radiative collapse was observed, the most fundamental density limit in tokamaks.

Date: December 1996

Creator: Maingi, R.; Mahdavi, M. A. & Jernigan, T. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Recent H-mode density limit experiments on DIII-D

A vast body of tokamak data is in good agreement with the empirical density limit scalings proposed by Hugill and Greenwald. These scalings have common puzzling features of showing no dependence on either impurity concentration or heating power, since the density limit is frequently correlated with a rapid rise of the edge radiation. Despite the resiliency of these scalings, several machines under restrictive conditions have operated at densities well above the predictions of these scalings, albeit with pellet injection and at varying degrees of confinement degradation. Furthermore, data from several machines display a weak dependence on heating power. These results cast doubt on the universal validity of both of these scalings. Nevertheless the fact remains that access to densities above Hugill-Greenwald scaling is very difficult. A number of theories based on radiative power balance in the plasma boundary have explained some but not all features of tokamak density limit behavior, and as ITER design studies recently brought to focus, a satisfactory understanding of this phenomenon is lacking. Motivated by a need for better understanding of effects of density and fueling on tokamak plasmas in general, the authors have conducted a series of experiments designed to identify and isolate physical effects suspected to be directly or indirectly responsible for the density limit. The physical effects being considered include: divertor power balance, MARFE, poloidally symmetric radiative instabilities, MHD instabilities, and transport. In this paper they first present a brief summary of the experimental results up to the writing of this paper. The remainder of the paper is devoted to a comparison of this data at the onset of the MARFE instability with predictions of theory and the implications of the results on access to densities beyond the Hugill-Greenwald limit.

Date: June 1997

Creator: Mahdavi, M.A.; Maingi, R. & Hyatt, A.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Density limit studies on DIII-D

The authors have studied the processes limiting plasma density and successfully achieved discharges with density {approximately}50% above the empirical Greenwald density limit with H-mode confinement. This was accomplished by density profile control, enabled through pellet injection and divertor pumping. By examining carefully the criterion for MARFE formation, the authors have derived an edge density limit with scaling very similar to Greenwald scaling. Finally, they have looked in detail at the first and most common density limit process in DIII-D, total divertor detachment, and found that the local upstream separatrix density (n{sub e}{sup sep,det}) at detachment onset (partial detachment) increases with the scrape-off layer heating power, P{sub heat}, i.e., n{sub e}{sup sep,det} {approximately} P{sub heat}{sup 0.76}. This is in marked contrast to the line-average density at detachment which is insensitive to the heating power. The data are in reasonable agreement with the Borass model, which predicted that the upstream density at detachment would increase as P{sub heat}{sup 0.7}.

Date: August 1, 1998

Creator: Maingi, R.; Mahdavi, M.A. & Petrie, T.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Bias-sustained shield plasma

Divertor biasing may provide a method for density and impurity control by enhancing the shielding efficiency of the scrape-off layer. The idea is to make the scrape-off plasma denser and thicker by heating it with a bias-driven current, and by inducing a radial E [times] B drift. If the bias is applied to flux surfaces at the outer edge of the usual scrape-off layer, a new layer of plasma can be added which is sustained by the bias-supplied power. A simple theoretical model will be presented which shows that there is a threshold condition which must be satisfied in order for the bias-heated plasma to be self-sustaining. The bias-sustained plasma must also be opaque enough to neutrals in order for it to be fueled by a gas puff, which means that it win serve as a shield to the core plasma against neutral impurities and hydrogen. Experiments performed on DIII-D have demonstrated both a modification of the central nickel impurity concentration and an increase in the ionization of hydrogen within the scrape-off layer due to biasing.

Date: September 1, 1992

Creator: Staebler, G.M.; Hyatt, A.W.; Schaffer, M.J. & Mahdavi, M.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Conceptual design summary for modifying Doublet III to a large dee-shaped configuration

The Doublet III tokamak is to be reconfigured by replacing its indented (doublet) vacuum vessel with a larger one of a dee-shaped cross section. This change will permit significantly larger elongated plasmas than is presently possible and will allow higher plasma current (up to 5 MA) and anticipated longer confinement time. Reactor relevant values of stable beta and plasma pressure are predicted. This modification, while resulting in a significant change in capability, utilizes most of the existing coils, structure, systems and facility.

Date: May 1, 1983

Creator: Davis, L.G.; Gallix, R.; Luxon, J.L.; Mahdavi, M.A.; Puhn, F.A.; Rock, P.J. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Reduction of Carbon Influx by Minimizing Title Heating DIII-D

No abstract prepared.

Date: November 1, 1999

Creator: Friend, M. E.; Baxi, C. B.; Mahdavi, M. A. & O'Neill, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Direct measurement of divertor exhaust neo enrichment in DIII-D

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.

Item Type: Article

Partner: UNT Libraries Government Documents Department

THE ROLE OF NEUTRALS IN H-MODE PEDESTAL FORMATION

An analytic model, derived from coupled continuity equations for the electron and neutral deuterium densities, is consistent with many features of edge electron density profiles in the DIII-D tokamak. For an assumed constant particle diffusion coefficient, the model shows that particle transport and neutral fueling produce electron and neutral density profiles that have the same characteristic scale lengths at the plasma edge. For systematic variations of density in H-mode discharges, the model predicts that the width of the electron density transport barrier decreases and the maximum gradient increases, as observed in the experiments. The widths computed from the model agree quantitatively with the experimental widths for conditions in which the model is valid. These results support models of transport barrier formation in which the H-mode particle barrier is driven by the edge particle flux and the width of the barrier is approximately the neutral penetration length.

Date: November 1, 2001

Creator: GROEBNER, R.J.; MAHDAVI, M.A.; LEONARD, A.W.; OSBORNE, T.H.; PORTER, G.D.; COLCHIN, R.J. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Stability of a radiative mantle in ITER

We report results of a study to evaluate the efficacy of various impurities for heat dispersal by a radiative mantle and radiative divertor(including SOL). We have derived a stability criterion for the mantle radiation which favors low Z impurities and low ratios of edge to core thermal conductivities. Since on the other hand the relative strength of boundary line radiation to core bremsstrahlung favors high Z impurities, we find that for the ITER physics phase argon is the best gaseous impurity for mantle radiation. For the engineering phase of ITER, more detailed analysis is needed to select between krypton and argon.

Date: December 1, 1996

Creator: Mahdavi, M.A.; Staebler, G.M.; Wood, R.D.; Whyte, D.G. & West, W.P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Pumping Characteristics of the DIII-D Cryopump

Beginning in 1992, the first of the DIII-D divertor baffles and cryocondensation pumps was installed. This open divertor configuration, located on the outermost floor of the DIII-D vessel, includes a cryopump with a predicted pumping speed of 50,000 {ell}/s excluding obstructions such as support hardware. Taking the pump structural and support characteristics into consideration, the corrected pumping speed for D{sub 2} is 30,000 {ell}/s [1]. In 1996, the second divertor baffle and cryopump were installed. This closed divertor structure, located on the outermost ceiling of the DIII-D vessel, has a cryopump with a predicted pumping speed of 32,000 {ell}/s. In the fall of 1999, the third divertor baffle and cryopump will be installed. This divertor structure will be located on the 45{sup o} angled corner on the innermost ceiling of the DIII-D vessel, known as the private flux region of the plasma configuration. With hardware supports factored into the pumping speed calculation, the private flux cryopump is expected to have a pumping speed of 15,000 {ell}/s. There was question regarding the effectiveness of the private flux cryopump due to the close proximity of the private flux baffle. This led to a conductance calculation study of the impact of rotating the cryopump aperture by 180{sup o} to allow for greater particle and gas exhaust into the cryopump's helium panel. This study concluded that the cost and schedule impact of changing the private flux cryopump orientation and design did not warrant the possible 20% (3,000 {ell}/s) increase in pumping ability gained by rotating the cryopump aperture 180{sup o}. The comparison of pumping speed of the first two cryocondensation pumps with the measured results will be presented as well as the calculation of the pumping speed for the private flux cryopump now being installed.

Date: November 1, 1999

Creator: Bozek, A.S.; Baxi, C.B.; Callis, R.W.; Mahdavi, M.A.; O'Neill, R.C. & Reis, E.E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

High Density H-Mode Discharges with Gas Fueling and Good Confinement on DIII-D

H-mode operation at high density is an attractive regime for future reactor-grade tokamaks [1]. High density maximizes fusion power output while the high confinement of H-mode keeps the plasma energy loss below the alpha heating power. One concern though is the energy released due to individual ELMs must be kept small to protect the diverter target from excess ablation. We report on discharges in DIII-D with electron densities as high as 1.45 times the Greenwald density, n{sub GW}(10{sup 20}m{sup -3})=I{sub p}(MA)/[{pi}{sup 2}(m)], with good confinement, H{sub ITER89P}=1.9, and ELMs with energy amplitude small enough to protect the divertor. These results were achieved at low triangularity single-null divertor, {delta}{approx}0.0 with a plasma current of 1.2 MA, q{sub 95} {approx} 3-4, and moderate neutral beam heating power of 2-4 MW. The density was controlled by moderate gas puffing and private flux pumping. A typical discharge is shown in Fig. 1 where upon gas puffing the pedestal density, n{sub e,epd}, quickly rises to {approx}0.8 x n{sub GW}. The confinement initially drops with the gas puff, on a longer timescale the central density rises, peaking the profile and increasing the confinement until an MHD instability terminates the high density and high confinement phase of the discharge. In this report we describe in detail edge pedestal changes and its effect on confinement as the density is increased. We then describe peaking of the density profile that offsets degradation of the pedestal at high density and restores good confinement. Finally we describe the small benign ELMs that result at these high densities.

Date: August 1, 2000

Creator: Leonard, A.W.; Osborne, T.H.; Mahdavi, M.A.; Fenstermacher, M.E.; Lasnier, C.J.; Petrie, T.W. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Particle exhaust of helium plasmas with actively cooled outboard pump limiter on Tore Supra

The superconducting tokamak Tore Supra was designed for long-pulse (30-s) high input power operation. Here observations on the particle-handling characteristics of the actively cooled modular outboard pump limiter (OPL) are presented for helium discharges. The important experimental result was that a modest pumping speed (1 m{sup 3}/s) of the OPL turbomolecular pump (TMP) provided background helium exhaust. This result came about due to a well-conditioned vessel wall with helium discharges that caused no wall outgasing. The particle accountability in these helium discharges was excellent, and the well-conditioned wall did not play a significant role in the particle balance. The helium density control, 25% density drop with OPL exhaust efficiency of {approximately}1%, was possible with TMP although this may not be the case with reactive gases such as deuterium. The observed quadratic increase of the OPL neutral pressure with helium density was consistent with an improvement of the particle control with increasing plasma density.

Date: August 1, 1995

Creator: Uckan, T.; Mioduszewski, P.K.; Loarer, T.; Chatelier, M.; Guilhem, D.; Lutz, T. et al.

Item Type: Report

Partner: UNT Libraries Government Documents Department

Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements.

Date: July 1, 1995

Creator: Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

EFFECT OF PARTICLE SOURCES ON THE STRUCTURE OF THE H-MODE PEDESTAL

Techniques of dimensional analysis have been applied to deuterium and hydrogen plasmas in DIII-D to test the postulate that the edge particle source plays a role in forming the edge H-mode density profile. These experiments show that the pedestal density scale length is typically a factor of two to three larger in hydrogen plasmas than in deuterium plasmas with dimensionally similar ion parameters. These results are in agreement with the postulate [1,2] that the density scale length is primarily determined by the local particle source, rather than by the shape of a hypothetical particle transport barrier. The electron temperature scale length displays a similar trend, albeit with a weaker density dependence. Thus the pedestal pressure gradient scale length is larger in hydrogen. It is also observed that the frequency of a coherent mode, localized within the pedestal, increases with the local density (i.e. inversely with the local density scale length) irrespective of the working gas species. This frequency is a factor of two lower in a hydrogen discharge than in a dimensionally similar deuterium plasma, a result which cannot be explained solely in terms of plasma physics variables.

Date: July 1, 2002

Creator: MAHDAVI, M.A.; R.J.GROEBNER; LEONARD, A.W.; LUCE, T.C.; McKEE, G.R.; MOYER, R.A. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

TRANSPORT OF ELM ENERGY AND PARTICLES INTO THE SOL AND DIVERTOR OF DIII-D

We report on DIII-D data that reveal the underlying processes responsible for transport of energy and particles from the edge pedestal to the divertor target during edge-localized modes (ELMs). The separate convective and conductive transport of energy due to an ELM is determined by Thomson scattering measurements of electron density and temperature in the pedestal. Conductive transport is measured as a drop in pedestal temperature and decreases with increasing density. The convective transport of energy, measured as a loss of density from the pedestal, however, remains constant as a function of density. From the SOL ELM energy is quickly carried to the divertor target. An expected sheath limit to the ELM heat flux set by the slower arrival of pedestal ions is overcome by additional ionization of neutrals generated from the divertor target as evidenced by a fast, {approx}100 {micro}s, rise in divertor density. A large in/out asymmetry of the divertor ELM heat flux is observed at high density, but becomes nearly symmetric at low density.

Date: June 1, 2002

Creator: LEONARD, A.W.; BOEDO, J.A.; FENSTERMACHER, M.E.; GROEBNER, R.J.; GROTH, M.; LASNIER, C.J. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Helium transport and exhaust studies in enhanced confinement regimes in DIII-D

A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust ({tau}*{sub He}/{tau}{sub E} {approximately} 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump ({approximately}5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D{sub He}/X{sub eff}{approximately}1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system.

Date: February 1, 1995

Creator: Wade, M.R.; Hillis, D.L.; Hogan, J.T.; Mahdavi, M.A.; Maingi, R.; West, W.P. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance ...

Date: November 1, 1999

Creator: O'Neill, R.C.; Bozek, A.S.; Friend, M.E.; Baxi, C.B.; Reis, E.E.; Mahdavi, M.A. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Divertor characterization experiments

Recent DIII-D experiments with enhanced Scrape-off Layer (SOL) diagnostics permit detailed characterization of the SOL and divertor plasma under various operating conditions. We observe two distinct plasma modes: attached and detached divertor plasmas. Detached plasmas are characterized by plate temperatures of only 1 to 2 eV. Simulation of detached plasmas using the UEDGE code indicate that volume recombination and charge exchange play an important role in achieving detachment. When the power delivered to the plate is reduced by enhanced radiation to the point that recycled neutrals can no longer be efficiently ionized, the plate temperature drops from around 10 eV to 1-2 eV. The low temperature region extends further off the plate as the power continues to be reduced, and charge exchange processes remove momentum, reducing the plasma flow. Volume recombination becomes important when the plasma flow is reduced sufficiently to permit recombination to compete with flow to the plate.

Date: June 18, 1996

Creator: Porter, G.D.; Allen, S.; Fenstermacher, M.; Hill, D.; Brown, M.; Jong, R.A, et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Divertor particle exhaust and wall inventory on DIII-D

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.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Compatibility of the Radiating Divertor with High Performance Plasmas in DIII-D

A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade, et al., Proc. 20th IAEA Fusion Energy Conf., Vilamoura, (2004)]. Our technique included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n{sub AR}/n{sub e} {le}0.004 on axis), although the argon profile was more peaked than the electron profile.

Date: May 18, 2006

Creator: Petrie, T W; Wade, M R; Brooks, N H; Fenstermacher, M E; Groth, M; Hyatt, A W et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

Particle control in the DIII-D advanced divertor

A new, electrically biasable, semi-closed divertor was installed and operated in the D3-D lower outside divertor location. The semi-closed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. (The planned cryogenic pumping is not yet installed). Electrical bias controls the distribution of particle recycle between the inner and outer divertors by {rvec E} {times} {rvec B} drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduce the sensitivity of plenum pressure to separatrix position. In particular, {rvec E} {times} {rvec B} drifts in the D3-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density. 5 refs., 8 figs.

Date: November 1, 1991

Creator: Schaffer, M.J.; Lippmann, S.I.; Mahdavi, M.A.; Petrie, T.W.; Stambaugh, R.D. (General Atomics, San Diego, CA (United States)); Hogan, J. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

OVERVIEW OF H-MODE PEDESTAL RESEARCH ON DIII-D

Developing an understanding of the processes that control the H-mode transport barrier is motivated by the significant impact this small region (typically <2% of the minor radius) can have on overall plasma performance. Conditions at the inner edge of the H-mode transport barrier can strongly influence the overall energy confinement, and the maximum density, and therefore fusion power, that can be achieved with the typically flat H-mode density profiles [1,2]. The ELM instability, which usually regulates the pressure gradient in the H-mode edge, can result in large power loads to, and erosion of, the divertor targets in a reactor scale device [3]. The goal of H-mode pedestal research at DIII-D is to: (1) develop a physics based model that would allow prediction of the conditions at the top of the H-mode pedestal, (2) develop an understanding of processes which control Type I ELM effects in the core and divertor, and (3) explore alternatives to the Type I ELM regime.

Date: July 1, 2001

Creator: OSBORNE, T.H.; BURRELL, K.H.; CARLSTROM, T.N.; CHU, M.S.; DOYLE, E.J.; FERRON, J.R. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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