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Improved operating scenarios of the DIII-D tokamak as a result of the addition of UNIX computer systems

Description: The increased use of UNIX based computer systems for machine control, data handling and analysis has greatly enhanced the operating scenarios and operating efficiency of the DRI-D tokamak. This paper will describe some of these UNIX systems and their specific uses. These include the plasma control system, the electron cyclotron heating control system, the analysis of electron temperature and density measurements and the general data acquisition system (which is collecting over 130 Mbytes of data). The speed and total capability of these systems has dramatically affected the ability to operate DIII-D. The improved operating scenarios include better plasma shape control due to the more thorough MHD calculations done between shots and the new ability to see the time dependence of profile data as it relates across different spatial locations in the tokamak. Other analysis which engenders improved operating abilities will be described.
Date: October 1, 1995
Creator: Henline, P.A.
Partner: UNT Libraries Government Documents Department

Evaluation of US demo helium-cooled blanket options

Description: A He-V-Li blanket design was developed as a candidate for the U.S. fusion demonstration power plant. This paper presents an 18 MPa helium-cooled, lithium breeder, V-alloy design that can be coupled to the Brayton cycle with a gross efficiency of 46%. The critical issue of designing to high gas pressure and the compatibility between helium impurities and V-alloy are addressed.
Date: October 1, 1995
Creator: Wong, C.P.C.; McQuillan, B.W. & Schleicher, R.W.
Partner: UNT Libraries Government Documents Department

Development of fast wave systems tolerant of time-varying loading

Description: A new approach to fast wave antenna array design based on the traveling wave antenna has been successfully demonstrated on the JFT-2M tokamak. A traveling wave antenna is powered though a single feed and the power flow from element to element is only via mutual reactive coupling. A combine is a particular type of traveling wave antenna, in which only the fed element and the element at the downstream end of the array are connected to vacuum feed troughs, while the intermediate elements are terminated with reactances inside the vacuum chamber. A twelve element combine for operation at 200 MHz was designed and fabricated at General Atomics, and installed and operated on the JFT-2M tokamak. The full output power of a single transmitter, 0.2 MW, was coupled to tokamak discharges with very little conditioning required. The input impedance of the combine was well matched to the transmission line impedance for all loading conditions, including vacuum (no plasma), Taylor discharge cleaning plasmas, and ohmic, L- and H-mode tokamak discharges with neutral beam heating without any adjustment of tuning elements.
Date: June 1, 1996
Creator: Pinsker, R.I.; Moeller, C.P.; Petty, C.C. & Phelps, D.A.
Partner: UNT Libraries Government Documents Department

Initial tests and operation of a 110 GHz, 1 MW gyrotron with evacuated waveguide system on the DIII-D tokamak

Description: A gyrotron producing nominally 1 MW at 110 GHz has been installed at the DIII-D tokamak and operated in a program of initial tests with a windowless evacuated transmission line. The alignment and first test operation were performed in an air environment at atmospheric pressure. Under these conditions, the tube produced rf output in excess of 800 kW for pulse lengths greater than 10 msec and power near 500 kW for pulse lengths of about 100 msec into a free space dummy load. The gyrotron was operated into evacuated corrugated waveguide in the full power parameter regime for pulse lengths of up to 500 msec injecting greater than 0.5 MW into DIII-D for a preliminary series of experiments. Generated powers greater than 900 kW were achieved. A parasitic oscillation at various frequencies between 20 and 100 MHz, which was generated during the pulsing of the gyrotron electron beam, was suppressed somewhat by a capacitive filter attached to the gyrotron itself. Addition of a magnetic shield intended to alter the magnetic field geometry below the cathode eliminated internal tube sparks. Rework of the external power and interlock circuitry to improve the immunity to electromagnetic interference was also done in parallel so that the fast interlock circuitry could be used. The latest results of the test program, the design of the free space load and other test hardware, and the transmission line will be presented.
Date: August 1, 1996
Creator: Lohr, J.; Ponce, D. & Tooker, J.F.
Partner: UNT Libraries Government Documents Department

A new technique to measure the neutralizer cell gas line density applied to a DIII-D neutral beamline

Description: The DIII-D tokamak employs eight ion sources for plasma heating. In order to obtain the maximum neutralization of energetic ions (providing maximum neutral beam power) and reduce the heat load on beamline internal components caused by residual energetic ions, sufficient neutral gas must be injected into the beamline neutralizer cell. The neutral gas flow rate must be optimized, however, since excessive gas will increase power losses due to neutral beam scattering and reionization. It is important, therefore, to be able to determine the neutralizer cell gas line density. A new technique which uses the ion source suppressor grid current to obtain the neutralizer cell gas line density has been developed. The technique uses the fact that slow ions produced by beam-gas interactions in the neutralizer cell during beam extraction are attracted to the negative potential applied to the suppressor grid, inducing current flow in the grid. By removing the dependence on beam energy and beam current a normalized suppressor grid current function can be formed which is dependent only on the gas line density. With this technique it is possible to infer the gas line density on a shot by shot basis.
Date: October 1, 1995
Creator: Kessler, D.N.; Hong, R.M. & Riggs, S.P.
Partner: UNT Libraries Government Documents Department

Manufacturing development of low activation vanadium alloys

Description: General Atomics is developing manufacturing methods for vanadium alloys as part of a program to encourage the development of low activation alloys for fusion use. The culmination of the program is the fabrication and installation of a vanadium alloy structure in the DIII-D tokamak as part of the Radiative Divertor modification. Water-cooled vanadium alloy components will comprise a portion of the new upper divertor structure. The first step, procuring the material for this program has been completed. The largest heat of vanadium alloy made to date, 1200 kg of V-4Cr-4Ti, has been produced and is being converted into various product forms. Results of many tests on the material during the manufacturing process are reported. Research into potential fabrication methods has been and continues to be performed along with the assessment of manufacturing processes particularly in the area of joining. Joining of vanadium alloys has been identified as the most critical fabrication issue for their use in the Radiative Divertor Program. Joining processes under evaluation include resistance seam, electrodischarge (stud), friction and electron beam welding. Results of welding tests are reported. Metallography and mechanical tests are used to evaluate the weld samples. The need for a protective atmosphere during different welding processes is also being determined. General Atomics has also designed, manufactured, and will be testing a helium-cooled, high heat flux component to assess the use of helium cooled vanadium alloy components for advanced tokamak systems. The component is made from vanadium alloy tubing, machined to enhance the heat transfer characteristics, and joined to end flanges to allow connection to the helium supply. Results are reported.
Date: October 1, 1996
Creator: Smith, J.P.; Johnson, W.R. & Baxi, C.B.
Partner: UNT Libraries Government Documents Department

Improving plasma shaping accuracy through consolidation of control model maintenance, diagnostic calibration, and hardware change control

Description: With the advent of more sophisticated techniques for control of tokamak plasmas comes the requirement for increasingly more accurate models of plasma processes and tokamak systems. Development of accurate models for DIII-D power systems, vessel, and poloidal coils is already complete, while work continues in development of general plasma response modeling techniques. Increased accuracy in estimates of parameters to be controlled is also required. It is important to ensure that errors in supporting systems such as diagnostic and command circuits do not limit the accuracy of plasma parameter estimates or inhibit the ability to derive accurate plasma/tokamak system models. To address this issue, we have developed more formal power systems change control and power system/magnetic diagnostics calibration procedures. This paper discusses our approach to consolidating the tasks in these closely related areas. This includes, for example, defining criteria for when diagnostics should be re-calibrated along with required calibration tolerances, and implementing methods for tracking power systems hardware modifications and the resultant changes to control models.
Date: December 1, 1995
Creator: Baggest, D.S.; Rothweil, D.A. & Pang, S.
Partner: UNT Libraries Government Documents Department

Inertial confinement fusion target component fabrication and technology development support: Annual report, October 1, 1995--September 30, 1996

Description: On December 30, 1990, the U.S. Department of Energy entered into a contract with General Atomics (GA) to be the Inertial Confinement Fusion (ICF) Target Component Fabrication and Technology Development Support contractor. In September 1995 this contract ended and a second contract was issued for us to continue this ICF target support work. This report documents the technical activities of the period October 1, 1995 through September 30, 1996. During this period, GA and our partners WJ Schafer Associates (WJSA) and Soane Technologies, Inc. (STI) were assigned 14 formal tasks in support of the Inertial Confinement Fusion program and its five laboratories. A portion of the effort on these tasks included providing direct {open_quotes}Onsite Support{close_quotes} at Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), and Sandia National Laboratory Albuquerque (SNLA). We fabricated and delivered over 800 gold-plated hohlraum mandrels to LLNL, LANL and SNLA. We produced nearly 1,200 glass and plastic target capsules for LLNL, LANL, SNLA and University of Rochester/Laboratory for Laser Energetics (UR/LLE). We also delivered over 100 flat foil targets for Naval Research Lab (NRL) and SNLA in FY96. This report describes these target fabrication activities and the target fabrication and characterization development activities that made the deliveries possible. The ICF program is anticipating experiments at the OMEGA laser and the National Ignition Facility (NIF) which will require capsules containing cryogenic layered D{sub 2} or deuterium-tritium (DT) fuel. We are part of the National Cryogenic Target Program to create and demonstrate viable ways to generate and characterize cryogenic layers. Substantial progress has been made on ways to both create and characterize viable layers. During FY96, significant progress was made in the design of the OMEGA Cryogenic Target System that will field cryogenic targets on OMEGA.
Date: February 1, 1997
Creator: Hoppe, M.
Partner: UNT Libraries Government Documents Department

Reduction of toroidal rotation by fast wave power in DIII-D

Description: The application of fast wave power in DIII-D has proven effective for both electron heating and current drive. Since the last RIF Conference FW power has been applied to advanced confinement regimes in DIII-D; negative central shear (NCS), VH- and H-modes, high {beta}{sub p}, and high-{ell}i. Typically these regimes show enhanced confinement of toroidal momentum exhibited by increased toroidal rotation velocity. Indeed, layers of large shear in toroidal velocity are associated with transport barriers. A rather common occurrence in these experiments is that the toroidal rotation velocity is decreased when the FW power is turned on, to lowest order independent of whether the antennas are phased for co or counter current drive. At present all the data is for co-injected beams. The central toroidal rotation can be reduced to 1/2 of the non-FW level. Here the authors describe the effect in NCS discharges with co-beam injection.
Date: April 1, 1997
Creator: Grassie, J.S. de; Baker, D.R. & Burrell, K.H.
Partner: UNT Libraries Government Documents Department

Upgrade of the DIII-D RF systems

Description: The DIII-D Advanced Tokamak Program requires the ability to modify the current density profile for extended time periods in order to achieve the improved plasma conditions now achieved with transient means. To support this requirement DIII-D has just completed a major addition to its ion cyclotron range of frequency (ICRF) systems. This upgrade project added two new fast wave current drive (FWCD) systems, with each system consisting of a 2 MW, 30 to 120 MHz transmitter, an all ceramic insulated transmission line, and water-cooled four-strap antenna. With this addition of 4 MW of FWCD power to the original 2 MW, 30 to 60 MHz capability, experiments can be performed with centrally localized current drive enhancement. For off-axis current modification, plans are in place to add 110 GHz electron cyclotron heating (ECH) power to DIII-D. Initially, 3 MW of power will be available with plans to increase the power to 6 MW and to 10 MW.
Date: October 1, 1995
Creator: Callis, R. W.; Cary, W. P. & O`Neill, R. C.
Partner: UNT Libraries Government Documents Department

Status of DIII-D plasma control

Description: A key component of the DIII-D Advanced Tokamak and Radiative Divertor Programs is the development and implementation of methods to actively control a large number of plasma parameters. These parameters include plasma shape and position, total stored energy, density, rf loading resistance, radiated power and more detailed control of the current profile. To support this research goal, a flexible and easily expanded digital control system has been developed and implemented. We have made parallel progress in modeling of the plasma, poloidal coils, vacuum vessel, and power system dynamics and in ensuring the integrity of diagnostic and command circuits used in control. Recent activity has focused on exploiting the mature digital control platform through the implementation of simple feedback controls of more exotic plasma parameters such as enhanced divertor radiation, neutral pressure and Marfe creation and more sophisticated identification and digital feedback control algorithms for plasma shape, vertical position, and safety factor on axis (q{sub 0}). A summary of recent progress in each of these areas will be presented.
Date: October 1, 1995
Creator: Walker, M. L.; Ferron, J. R. & Penaflor, B.
Partner: UNT Libraries Government Documents Department

Fast wave heating and current drive in DIII-D discharges with negative central shear

Description: The noninductive current driven by fast Alfven waves (FWCD) has been applied to discharges in DIII-D with negative central shear. Driven currents as high as 275 kA have been achieved with up to 3 MW of fast wave power with the efficiency and profile as predicted by theory-based modeling. When counter-current FWCD was applied to discharges with negative central shear, the negative shear was strengthened and prolonged, showing that FWCD can help to control the current profile in advanced tokamak discharges. Under some conditions in negative central shear, the plasma spontaneously makes a transition into a regime of improved performance, with a reduction in both the ion and the electron heat diffusivities. Up to 3 MW of fast wave power has been successfully coupled into H-mode discharges with large edge localized modes through use of an innovative decoupler/hybrid power splitter combination.
Date: October 1, 1996
Creator: Prater, R.; Austin, M.E. & Baity, F.W.
Partner: UNT Libraries Government Documents Department

Divertor plasma physics experiments on the DIII-D tokamak

Description: 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.
Partner: UNT Libraries Government Documents Department

Effects of ExB velocity shear and magnetic shear on turbulence and transport in magnetic confinement devices

Description: One of the scientific success stories of fusion research over the past decade is the development of the ExB shear stabilization model to explain the formation of transport barriers in magnetic confinement devices. This model was originally developed to explain the transport barrier formed at the plasma edge in tokamaks after the L (low) to H (high) transition. This concept has the universality needed to explain the edge transport barriers seen in limiter and divertor tokamaks, stellarators, and mirror machines. More recently, this model has been applied to explain the further confinement improvement from H (high)-mode to VH (very high)-mode seen in some tokamaks, where the edge transport barrier becomes wider. Most recently, this paradigm has been applied to the core transport barriers formed in plasmas with negative or low magnetic shear in the plasma core. These examples of confinement improvement are of considerable physical interest; it is not often that a system self-organizes to a higher energy state with reduced turbulence and transport when an additional source of free energy is applied to it. The transport decrease that is associated with ExB velocity shear effects also has significant practical consequences for fusion research. The fundamental physics involved in transport reduction is the effect of ExB shear on the growth, radial extent and phase correlation of turbulent eddies in the plasma. The same fundamental transport reduction process can be operational in various portions of the plasma because there are a number ways to change the radial electric field Er. An important theme in this area is the synergistic effect of ExB velocity shear and magnetic shear. Although the ExB velocity shear appears to have an effect on broader classes of microturbulence, magnetic shear can mitigate some potentially harmful effects of ExB velocity shear and facilitate turbulence stabilization.
Date: November 1, 1996
Creator: Burrell, K. H.
Partner: UNT Libraries Government Documents Department