The introduction of a divertor Thomson scattering system in DID-D has enabled accurate determination of the plasma properties in the divertor region. We identify two plasma regimes; detached and attached. The electron temperature in the detached regime is about 2 eV, much lower than 5 to 10 eV determined earlier. We show that fluid models of the DIU-D scrape-off layer plasma are able to reproduce many of the features of these two plasma regimes, including the boundaries for transition between them. Detailed comparison between the results obtained from the fluid models and experiment suggest the models underestimate the spatial extent of the low temperature region associated the detached plasma mode. We suggest that atomic physics processes at the low electron temperatures reported here may account for this discrepancy. lGeneral Atomics, P 0 Box 85608, San Diego, CA 92186

In this paper we present measurements of the global power and particle balance in the high-performance phase of negative central magnetic shear (NCS) discharges and compare with reference VH-mode discharges. The principal differences observed are that NCS has a much lower fraction of the total input power flowing into the boundary, less core radiation, and larger rate of stored energy increase as a fraction of total power. Scrape-off layer (SOL) temperature and divertor heat flux profiles, and radiation profiles at the midplane, are similar to VH-mode. Due to the good core particle confinement and efficient fueling by neutral beam injection (NBI), with little gas puffing, the gas load on the walls and the recycling are very low during the NCS discharges. This results in a rate of density rise relative to beam fueling at the L to H transition time which is 1/3 of the value for VH transitions, which is in turn 1/2 that for L-to-ELMing-H-mode transitions. GENERAL ATOMICS REPORT GA-A22353 iii C. J. Lusnier, et al. CHARACTERISTICS OF THE SCRAPE-OFF LAYER IN DIII-D HIGH-PERFORMANCE NEGATIVE CENTRAL MAGNETIC SHEAR DISCHARGES

This paper reports on the progress in developing a library of code modules under the auspices of the National Transport Code Collaboration (NTCC). Code modules are high quality, fully documented software packages with a clearly defined interface. The modules provide a variety of functions, such as implementing numerical physics models; performing ancillary functions such as I/O or graphics; or providing tools for dealing with common issues in scientific programming such as portability of Fortran codes. Researchers in the plasma community submit code modules, and a review procedure is followed to insure adherence to programming and documentation standards. The review process is designed to provide added confidence with regard to the use of the modules and to allow users and independent reviews to validate the claims of the modules' authors. All modules include source code; clear instructions for compilation of binaries on a variety of target architectures; and test cases with well-documented input and output. All the NTCC modules and ancillary information, such as current standards and documentation, are available from the NTCC Module Library Website http://w3.pppl.gov/NTCC. The goal of the project is to develop a resource of value to builders of integrated modeling codes and to plasma physics researchers generally. Currently, there are more than 40 modules in the module library.

Toroidal plasmas created with negative magnetic shear in the core region offer advantages in terms of MHD stability properties. These plasmas, transiently created in several tokamaks, have exhibited high-performance as measured by normalized stored energy and neutron production rates. A critical issue with extending the duration of these plasmas is the need to maintain the off-axis-peaked current distribution required to support the minimum in the safety factor q at large radii. We present equilibrium and transport simulations that explore the use of electron cyclotron heating and current drive to maintain this negative shear configuration. Using parameters consistent with DIII-D tokamak operation (Strait E et al 1995 Phys. Rev. Lett. 75 4421, Rice B W et al 1996 Nucl. Fusion 36 1271), we find that with sufficiently high injected power, it is possible to achieve steady-state conditions employing well aligned electron cyclotron and bootstrap current drive in fully non-inductively current-driven configurations.

External impurity injection into L mode edge discharges in DIII-D has produced clear confinement improvement (a factor of 2 in energy confinement and neutron emission), reduction in all transport channels (particularly ion thermal diffusivity to the neoclassical level), and simultaneous reduction of long wavelength turbulence. Suppression of the long wavelength turbulence and transport reduction are attributed to synergistic effects of impurity induced enhancement of E × B shearing rate and reduction of toroidal drift wave turbulence growth rate. A prompt reduction of density fluctuations and local transport at the beginning of impurity injection appears to result from an increased gradient of toroidal rotation enhancing the E × B shearing. Transport simulations carried out using the National Transport Code Collaboration demonstration code with a gyro-Landau fluid model, GLF23, indicate that E × B shearing suppression is the dominant transport suppression mechanism.

The objective of this work is to develop realizable advanced tokamak scenarios based on experimental results from high performance negative central shear (shots 99411,103322) and QDB (quiescentelm free,double barrier,shot 103818) discharges in DIII-D. The study is carried out under a new collaboration between GA and LLNL in transport modelling, the Onetwo/Corsica project. The collaboration is based on the GA transport code ONETWO and LLNL's Corsica which results in increased modeling capabilites. Improvements in the resolution of magnetic reconstruction techniques in EFIT and refined accuracy in the DIII-D MSE diagnosticsignificantly enhance our ability to reliably design ECCD and beam current drive scenarios for DIII-D. An interesting application of ONETWO/Corsica is the use of time dependent EQDSK equilibrium data files from experiment (TDEM) for determination of non inductive current profiles. The TDEM method allows us to treat the total and ohmic currents as experimentally observed quantities thus making an experimental determination of the non-inductive current profile possible. In H-mode discharges the formation of an edge transport barrier leads to significant edge bootstrap current which can drive MHD instabilitites such as ELMs. In these cases the theoretically calculated edge bootstrap current can be significantly larger than TDEM methods indicate. This may be due to uncertainty in our ability to model such discharges accurately because of the kink unstable nature of the predicted edge current profile. The stability packages in the new ONETWO/Corsica will be used to evaluate the effects of ELMS on the predicted non-inductive current profiles. Control of the ITB and, in particular, its stability during the long time evolution of AT discharges is addressed in this work. With density profiles fixed in space and time we introdce dynamic transport barrier modeling with the GLF23 confinement model. This confinement model has had some success in modeling DIII-D discharges1 and consequently represents our best choice for DIII-D AT scenario development at this time. We are also exploring scenarios for use of ECH/ECCD to control the QDB mode. Using measurements of electron and ion transport coefficients the ability to modify and control the q profile in QDB discharges with various amounts of ECH/ECCD power are modeled using the ray tracing code TORAY-GA.

Major fusion experiments and modeling efforts rely on joint research of scientists from several locations around the world. A variety of software tools are in use to provide remote interactive access to facilities and data are routinely available over wide-area-network connections to researchers. Audio and video communications, monitoring of control room information and synchronization of remote sites with experimental operations all enhance participation during experiments. Remote distributed computing capabilities allow utilization of off-site computers that now help support the demands of control room analyses and plasma modeling. A collaborative software development project is currently using object technologies with CORBA-based communications to build a network executable transport code that further demonstrates the ability to utilize geographically dispersed resources. Development to extend these concepts with security and naming services and possible applications to instrumentation systems has been initiated. An Information Technology Initiative is deploying communication systems, ISDN (telephone) and IP (network) audio/video (A/V) and web browser-based, to build the infrastructure needed to support remote physics meetings, seminars and interactive discussions.

External impurity injection into L-mode edge discharges in DIII-D has produced clear increases in confinement (factor of 2 in energy confinement and neutron emission), reduction in all transport channels (particularly ion thermal diffusivity to the neoclassical level), and simultaneous reduction of long-wavelength turbulence. Suppression of the flux wavelength turbulence and transport reduction are attributed to synergistic effects of impurity-induced enhancement of ExB shearing rate and reduction of toroidal drift wave turbulence. A prompt reduction of density fluctuations and local transport at the beginning of impurity injection appears to result from an increased gradient of toroidal rotation enhancing the ExB shearing. Transport simulations carried out using the National Transport Code Collaboration Demonstration Code with a GyroLandau fluid model, GLF23, indicate ExB shearing suppression is the dominant transport suppression mechanism.

Recently a number of enhancements to the computer hardware infrastructure have been implemented at the DIII–D National Fusion Facility. Utilizing these improvements to the hardware infrastructure, software enhancements are focusing on streamlined analysis, automation, and graphical user interface (GUI) systems to enlarge the user base. The adoption of the load balancing software package LSF Suite by Platform Computing has dramatically increased the availability of CPU cycles and the efficiency of their use. Streamlined analysis has been aided by the adoption of the MDSplus system to provide a unified interface to analyzed DIII–D data. The majority of MDSplus data is made available in between pulses giving the researcher critical information before setting up the next pulse. Work on data viewing and analysis tools focuses on efficient GUI design with object–oriented programming (OOP) for maximum code flexibility. Work to enhance the computational infrastructure at DIII–D has included a significant effort to aid the remote collaborator since the DIII–D National Team consists of scientists from nine national laboratories, 19 foreign laboratories, 16 universities, and five industrial partnerships. As a result of this work, DIII–D data is available on a 24×7 basis from a set of viewing and analysis tools that can be run on either the collaborators’ or DIII–D's computer systems. Additionally, a web based data and code documentation system has been created to aid the novice and expert user alike.

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultraviolet spectroscopy and 2-D images of impurity emission show that carbon radiation near the X-point, and deuterium radiation near the target plates contribute to the reduction in Te. Uniformity of radiated power (Prad) (within a factor of 2) along the outer divertor leg, with peak heat flux on the divertor target reduced fourfold, was obtained. A comparison with 2-D fluid simulations shows good agreement when physical sputtering and an ad hoc chemical sputtering source (0.5%) from the private flux region surface are used.

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also,serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma.Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma-wall interaction. In this paper we will describe a set of experiments on DIII-D designed to provide detailed two-dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

Deuterium gas injected into ELMing H mode divertor discharges in the DIII-D tokamak typically reduced the total power at the divertor target ~2 times and the peak heat flux ~3 to 5 times with modest (<10%) degradation in plasma energy confinement. The parameter range for the discharges investigated was: Ip=1.0-2.0 MA, q95 approximately= 2.4-6.0 and total input power (≲20 MW. Most of this reduction in heat flux occurred at the sudden formation of a high density, highly radiating region located between the outboard divertor separatrix strike point and the X point. This divertor behaviour is associated with a `partially detached' divertor plasma condition, which is referred to in this paper as the partially detached divertor (PDD) regime. With the onset of the PDD, typically at a line averaged density of 0.6 to 0.7 times the Greenwald density limit, an abrupt reduction in plasma electron pressure (≳4 times) was observed at the outboard divertor separatrix strike point; at the same time, however, only a modest (≲30%) change in the electron pressure was observed upstream near the outboard midplane separatrix. The data suggest that significant plasma momentum loss occurred between the high density, highly radiative region and the (downstream) divertor separatrix target. Plasma performance showed little degradation with the onset of the PDD regime. Deuterium injection made only modest changes in the temperature and density profile shapes near the midplane separatrix of the main plasma. The PDD approach is shown to be compatible with discharges operating at low safety factor (i.e. q95 equivalent to 2.9) and to be effective in significantly reducing toroidal asymmetry in observed divertor plasma properties (e.g., heat flux). The potential for operating in a steady state has been demonstrated using feedback control of the neutral pressure outside the main plasma

This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (partially detached divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2D (R, Z) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. Dα emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. Lα and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.