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Fusion ignition research experiment

Description: Understanding the properties of high gain (alpha-dominated) fusion plasmas in an advanced toroidal configuration is the largest remaining open issue that must be addressed to provide the scientific foundation for an attractive magnetic fusion reactor. The critical parts of this science can be obtained in a compact high field tokamak which is also likely to provide the fastest and least expensive path to understanding alpha-dominated plasmas in advanced toroidal systems.
Date: July 18, 2000
Creator: Meade, Dale
Partner: UNT Libraries Government Documents Department

The Burning Plasma Experiment conventional facilities

Description: The Burning Program Plasma Experiment (BPX) is phased to start construction of conventional facilities in July 1994, in conjunction with the conclusion of the Tokamak Fusion Test Reactor (TFTR) project. This paper deals with the conceptual design of the BPX Conventional Facilities, for which Functional and Operational Requirements (F&ORs) were developed. Existing TFTR buildings and utilities will be adapted and used to satisfy the BPX Project F&ORs to the maximum extent possible. However, new conventional facilities will be required to support the BPX project. These facilities include: The BPX building; Site improvements and utilities; the Field Coil Power Conversion (FCPC) building; the TFTR modifications; the Motor Generation (MG) building; Liquid Nitrogen (LN{sub 2}) building; and the associated Instrumentation and Control (I&C) systems. The BPX building will provide for safe and efficient shielding, housing, operation, handling, maintenance and decontamination of the BPX and its support systems. Site improvements and utilities will feature a utility tunnel which will provide a space for utility services--including pulse power duct banks and liquid nitrogen coolant lines. The FCPC building will house eight additional power supplied for the Toroidal Field (TF) coils. The MG building will house the two MG sets larger than the existing TFTR MG sets. This paper also addresses the conventional facility cost estimating methodology and the rationale for the construction schedule developed. 6 figs., 1 tab.
Date: December 1, 1991
Creator: Commander, J. C.
Partner: UNT Libraries Government Documents Department

Compact tokamak reactors. Part 1 (analytic results)

Description: We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model.
Date: September 13, 1996
Creator: Wootton, A. J.; Wiley, J. C.; Edmonds, P. H. & Ross, D. W.
Partner: UNT Libraries Government Documents Department

Nondimensional transport experiments on DIII-D and projections to an ignition tokamak

Description: The concept of nondimensional scaling of transport makes it possible to determine the required size for an ignition device based upon data from a single machine and illuminates the underlying physics of anomalous transport. The scaling of cross-field heat transport with the relative gyroradius {rho}*, the gyroradius normalized to the plasma minor radius, is of particular interest since {rho}* is the only nondimensional parameter which will vary significantly between present day machines and an ignition device. These nondimensional scaling experiments are based upon theoretical considerations which indicate that the thermal heat diffusivity can be written in the form {chi} = {chi}{sub B}{rho}*{sup x{sub {rho}}} F({beta}, v*, q, R/a, {kappa}, T{sub e}/T{sub i},...), where {chi}{sub B} = cT/eB. As explained elsewhere, x{sub {rho}} = 1 is called gyro-Bohm scaling, x{sub {rho}} is Bohm scaling, x{sub {rho}} = {minus}1/2 is Goldston scaling, and x{sub {rho}} = {minus}1 is stochastic scaling. The DIII-D results reported in this paper cover three important aspects of nondimensional scaling experiments: the testing of the underlying assumption of the nondimensional scaling approach, the determination of the {rho}* scaling of heat transport for various plasma regimes, and the extrapolation of the energy confinement time to future ignition devices.
Date: July 1, 1996
Creator: Petty, C.C.; Luce, T.C.; Balet, B.; Christiansen, J.P. & Cordey, J.G.
Partner: UNT Libraries Government Documents Department

Heat transport in PBX-M high {beta}{sub p} plasmas

Description: PBX-M high beta poloidal discharges routinely transition into the H-mode regime: typically, a quiescent phase followed by an MHD active phase characterize the H-mode period. An analysis of the energy transport during these phases is conducted using the experimental data and the TRANSP code; effective diffusivities are computed to quantify the energy transport of the thermal component of the plasma. Compared to the L-mode, the quiescent H-phase is characterized by a decrease of the thermal ion energy transport and a flattening of the associated effective diffusivity profile. An error analysis is presented. Enhanced fast-ion losses are observed during the MHD active phase: particles in the lower end of the fast-ion energy spectrum with large perpendicular velocity component are predominantly affected. These losses must be taken into account in the analysis in order to reproduce the measured stored energy and time evolution of the neutron production rate during the MHD active phase.
Date: April 1, 1992
Creator: LeBlanc, B.; Kaye, S.; Bell, R.; Fishman, H.; Hatcher, R.; Kaita, R. et al.
Partner: UNT Libraries Government Documents Department

Midplane Faraday Rotation: A densitometer for BPX

Description: The density in a high field, high density tokamak such as BPX can be determined by measuring the Faraday rotation of a 10.6 {mu}m laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then n{sub e}(R) can be readily obtained with a simple Abel version about the center line of the tokamak. For BPX operated at full field and density, the rotation angle would be quite large -- about 75{degrees} per pass. A layout in which a single laser beam is fanned out in the horizontal midplane of the tokamak, with a set of retroreflectors on the far side of the vacuum vessel, would provide good spatial resolution, depending only upon the number of reflectors. With this proposed layout, only one window would be needed. Because the rotation angle is never more than 1 ``fringe,`` the data is always good, and it is also a continuous measurement in time. Faraday rotation is dependent only upon the plasma itself, and thus is not sensitive to vibration of the optical components. Simulations of the expected results show that BPX would be well served even at low densities by a Midplane Faraday Rotation densitometer of {approximately}64 channels. Both TFTR and PBX-M would be suitable test beds for the BPX system.
Date: February 1, 1992
Creator: Jobes, F. C. & Mansfield, D. K.
Partner: UNT Libraries Government Documents Department

PBX-M ion Bernstein wave heating overview

Description: A high power ion Bernstein wave heating system has been introduced on PBX-M for heating and for controlling the plasma pressure profile in an effort to achieve the stable high beta ``second stability`` regime. The pressure profile can be controlled through local bulk ion heating as well as density profile control. In bean-shaped plasmas with plasma currents range from 180 kA to 250 kA, good ion heating up to the highest, applied rf power, ({approx}700 kW) has been observed. The observed broadening of the ion temperature profile is consistent with localized off-axis bulk ion heating as predicted by IBW ray tracing calculations. Application of IBW also resulted in a greatly modified density profile. The ability for IBW to change the density profile appears to be particularly attractive for controlling the bootstrap current profile for advanced tokamaks. Many important IBWH-related edge physics results were also obtained, including ponderomotive edge plasma modification and parametric instability onset conditions. The experimental plan for the next IBW run includes investigation of synergy with LHCD, attainment of high bootstrap current fraction discharges utilizing the IBW density profile control, and exploration of high beta plasma regimes.
Date: April 1, 1993
Creator: Ono, M.; Chu, T. K.; Hermann, H.; LeBlanc, B.; Tighe, W.; Bell, R. et al.
Partner: UNT Libraries Government Documents Department

Spitzer or neoclassical resistivity: A comparison between measured and model poloidal field profiles on PBX-M

Description: Direct measurements of the radial profile of the magnetic field line pitch on PBX-M coupled with model predictions of these profiles allow a critical comparison with the Spitzer and neoclassical models of plasma parallel resistivity. The measurements of the magnetic field line pitch are made by Motional Stark Effect polarimetry, while the model profiles are determined by solving the poloidal field diffusion equation in the TRANSP transport code using measured plasma profiles and assuming either Spitzer or neoclassical resistivity. The measured field pitch profiles were available for only seven cases, and the model profiles were distinguishable from each other in only three of those cases due to finite resistive diffusion times. The data in two of these three were best matched by the Spitzer model, especially in the inner half of the plasma. Portions of the measured pitch profiles for these two cases and the full profiles for other cases, however, departed significantly from both the Spitzer and neoclassical models, indicating a plasma resistivity profile different from either model.
Date: January 1, 1992
Creator: Kaye, S. M.; Hatcher, R.; Kaita, R.; Kessel, C.; LeBlanc, B.; McCune, D. C. et al.
Partner: UNT Libraries Government Documents Department

Initial boronization of PBX-M using ablation of solid boronized probes

Description: The initial boronization of PBX-M was performed using the sequential ablation of two types of solid target probes. Probe-1 in a mushroom shape consisted of a 10.7% boronized 2-D C-C composite containing 3.6 g of boron in a B{sub 4}C binder. Probe-2 in a rectangular shape consisted of an 86% boronized graphite felt composite containing 19.5 g of 40 {mu} boron particles. After boronization with Probe-1, the loop voltage during 1 MW neutral beam heated plasmas decreased 27% and volt-sec consumption decreased 20%. Strong peripheral spectral lines from low-Z elements decreased by factors of about 5. The central oxygen density decreased 15--20%. The total radiated power during neutral beam injection decreased by 43%. Probe-2 boronization exhibited improved operating conditions similar to Probe-1, but for some parameters, a smaller percentage change occurred due to the residual boron from the previous boronization using Probe-1. The ablation rates of both probes were consistent with front face temperatures at or slightly above the boron melting point. These results confirm the effectiveness of the solid target boronization (STB) technique as a real-time impurity control method for replenishing boron depositions without the use of hazardous borane compounds.
Date: May 1, 1993
Creator: Kugel, H. W.; Hirooka, Y.; Kaita, R.; Kaye, S.; Khandagle, M.; Timberlake, J. et al.
Partner: UNT Libraries Government Documents Department

Control of the current density profile with lower hybrid current drive on PBX-M

Description: Lower hybrid current drive (LHCD) is being explored as a means to control the current density profile on PBX-M with the goal of raising the central safety factor q(O) to values of 1.5-2 to facilitate access to a full-volume second stable regime. Initial experiments have been conducted with up to 400 kW of 4.6 GHz LH power in circular and indented plasmas with modest parameters. A tangential-viewing two-dimensional hard x-ray imaging diagnostic has been used to observe the bremsstrahlung emission from the suprathermal electrons generated during LHCD. Hollow hard x-ray images have indicated off-axis localization of the driven current. A serious obstacle to the control of the current density profile with LHCD is the concomitant generation of MHD activity, which can seriously degrade the confinement of suprathermal electrons. By combining neutral beam injection with LHCD, an MHD-free condition has been obtained where q(O) is raised above 1.
Date: July 1, 1993
Creator: Bell, R. E.; Bernabei, S.; Chu, T. K.; Gettelfinger, G.; Greenough, N.; Hatcher, R. et al.
Partner: UNT Libraries Government Documents Department

The Burning Plasma Experiment conventional facilities

Description: The Burning Program Plasma Experiment (BPX) is phased to start construction of conventional facilities in July 1994, in conjunction with the conclusion of the Tokamak Fusion Test Reactor (TFTR) project. This paper deals with the conceptual design of the BPX Conventional Facilities, for which Functional and Operational Requirements (F ORs) were developed. Existing TFTR buildings and utilities will be adapted and used to satisfy the BPX Project F ORs to the maximum extent possible. However, new conventional facilities will be required to support the BPX project. These facilities include: The BPX building; Site improvements and utilities; the Field Coil Power Conversion (FCPC) building; the TFTR modifications; the Motor Generation (MG) building; Liquid Nitrogen (LN{sub 2}) building; and the associated Instrumentation and Control (I C) systems. The BPX building will provide for safe and efficient shielding, housing, operation, handling, maintenance and decontamination of the BPX and its support systems. Site improvements and utilities will feature a utility tunnel which will provide a space for utility services--including pulse power duct banks and liquid nitrogen coolant lines. The FCPC building will house eight additional power supplied for the Toroidal Field (TF) coils. The MG building will house the two MG sets larger than the existing TFTR MG sets. This paper also addresses the conventional facility cost estimating methodology and the rationale for the construction schedule developed. 6 figs., 1 tab.
Date: January 1, 1991
Creator: Commander, J.C.
Partner: UNT Libraries Government Documents Department

Program status 3. quarter -- FY 1990: Confinement systems programs

Description: Highlights of the DIII-D Research Operations task are: completed five weeks tokamak operations; initiated summer vent; achievement of 10.7% beta; carried out first dimensionless transport scaling experiment; completed IBW program; demonstrated divertor heat reduction with gas puffing; field task proposals presented to OFE; presentation of DIII-D program to FPAC; made presentation to Admiral Watkins; and SAN safety review. Summaries are given on research programs, operations, program development, hardware development, operations support and collaborative efforts. Brief summaries of progress on the International Cooperation task include: TORE SUPRA, ASDEX, JFT-2M, and JET. Funding for work on CIT physics was received this quarter. Several physics R and D planning tasks were initiated. Earlier in FY90, a poloidal field coil shaping system (PFC) was found for DIGNITOR. This quarter more detailed analysis has been done to optimize the design of the PFC system.
Date: July 24, 1990
Partner: UNT Libraries Government Documents Department

The configuration development of the compact ignition tokamak (CIT) device

Description: The paper consists of viewgraphs discussing different component design options for the Compact Ignition Tokamak (CIT). The author concludes that a CIT device configuration has been established which meets physics requirements, performance margins and engineering component design needs. The baseline (R/sub 0/ = 1.75 meters) machine appears reasonable based on the analysis performed to date. (LSP)
Date: January 1, 1987
Creator: Brown, T.G.
Partner: UNT Libraries Government Documents Department

CIT (Compact Ignition Tokamak) fueling

Description: A series of viewgraphs present issues related to the conceptual design of the Compact Ignition Tokamak. The presentation includes discussions of fueling issues, pellet injector technology, pellet ablation and penetration, particle confinement, and fueling scenarios. The author concludes that existing technology should be used for the basic injector while several options for enhanced velocity injectors are under development; adequate models exist for pellet ablation; and improvements in confinement models can come from the TFTR and JET programs. (DWL)
Date: January 1, 1988
Creator: Houlberg, W.A.
Partner: UNT Libraries Government Documents Department

The configuration development of the compact ignition tokamak device

Description: The Compact Ignition Tokamak (CIT) device is planned as the next major fusion device to be built at the Princeton Plasma Physics Laboratory to demonstrate ignition operations of a burning plasma. Stringent engineering requirements have been imposed on this device by physics necessities of high margins against ignition and by cost constraints in minimizing the overall cost of the project. A compact design has been developed under these design conditions incorporating many unique design features, including a hydraulic preload system to provide a compression load to the toroidal field (TF) inner leg and using a high-strength copper-Inconel composite material in the design of the TF coil and the ohmic heating solenoid. The device is inertially cooled by liquid nitrogen, and the vacuum vessel, coils, and supporting structure are contained in a thermally insulated cryostat. A close-in igloo shield surrounds the device to provide the capability for hands-on access within the test cell and also to minimize activation. Even with the compact nature of this device, there still remains the basic requirement of maximizing access to the plasma for diagnostics and heating components; access for electrical leads and coolant lines; and access to provide the capability of remotely maintaining all diagnostic and peripheral equipment that interfaces with the device. This paper describes the configurational development that has taken place during the conceptual design period of the CIT project, highlighting the major design integration features used to develop a functional device that meets the physics and component design requirements. 1 ref., 7 figs.
Date: January 1, 1988
Creator: Brown, T.G.
Partner: UNT Libraries Government Documents Department

Alpha-particle effects on high-n instabilities in tokamaks

Description: Hot ..cap alpha..-particles and thermalized helium ash particles in tokamaks can have significant effects on high toroidal mode number instabilities such as the trapped-electron drift mode and the kinetically calculated magnetohydrodynamic ballooning mode. In particular, the effects can be stabilizing, destabilizing, or negligible, depending on the parameters involved. In high-temperature tokamaks capable of producing significant numbers of hot ..cap alpha..-particles, the predominant interaction of the mode with the ..cap alpha..-particles is through resonances of various sorts. In turn, the modes can cause significant anomalous transport of the ..cap alpha..-particles and the helium ash. Here, results of comprehensive linear eigenfrequency-eigenfunction calculations are presented for relevant realistic cases to show these effects. 24 refs., 12 figs., 6 tabs.
Date: June 1, 1988
Creator: Rewoldt, G.
Partner: UNT Libraries Government Documents Department

Plasma diagnostics for the compact ignition tokamak

Description: The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10/sup 21/m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs.
Date: June 1, 1988
Creator: Medley, S. S. & Young, K. M.
Partner: UNT Libraries Government Documents Department

Simulation of a compact ignition tokamak discharge (CIT-2L)

Description: A reference simulation of a compact ignition tokamak (CIT) limiter discharge is carried out using the 1-1/2-D BALDUR transport code. The parameters for the discharge are as in the CIT-2L design; these parameters are described here, and the results of the simulation are discussed. It is found that plasma ignition can be reached and sustained within the specified constraints.
Date: February 1, 1987
Creator: Stotler, D.P. & Bateman, G.
Partner: UNT Libraries Government Documents Department

Recent progress on the Compact Ignition Tokamak (CIT)

Description: This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule.
Date: January 1, 1987
Creator: Ignat, D.W.
Partner: UNT Libraries Government Documents Department

Experimental and theoretical research in applied plasma physics. Technical progress report, October 15, 1990--October 14, 1993

Description: This report discusses research in the following areas: fusion theory and computations; theory of thermonuclear plasmas; user service center; high poloidal beta studies on PBX-M; fast ECE fluctuation diagnostic for balloning mode studies; x-ray imaging diagnostic; millimeter/submillimeter-wave fusion ion diagnostics; small scale turbulence and nonlinear dynamics in plasmas; plasma turbulence and transport; phase contrast interferometer diagnostic for long wavelength fluctuations in DIII-D; and charged and neutral fusion production for fusio plasmas.
Date: June 1, 1992
Creator: Porkolab, M.
Partner: UNT Libraries Government Documents Department

Fast electron current density profile and diffusion studies during LHCD in PBX-M

Description: Successful current profile control experiments using lower hybrid current drive (LCHD) clearly require knowledge of (1) the location of the driven fast electrons and (2) the ability to maintain that location from spreading due to radial diffusion. These issues can be addressed by examining the data from the hard x-ray camera on PBX-M, a unique diagnostic producing two-dimensional, time resolved tangential images of fast electron bremsstrahlung. Using modeling, these line-of-sight images are inverted to extract a radial fast electron current density profile. We note that ``hollow`` profiles have been observed, indicative of off-axis current drive. These profiles can then be used to calculate an upper bound for an effective fast electron diffusion constant: assuming an extremely radially narrow lower hybrid absorption profile and a transport model based on Rax and Moreau, a model fast electron current density profile is calculated and compared to the experimentally derived profile. The model diffusion constant is adjusted until a good match is found. Applied to steady-state quiescent modes on PBX-M, we obtain an upper limit for an effective diffusion constant of about D*=1.1 m{sup 2}/sec.
Date: August 1993
Creator: Jones, S. E.; Kesner, J.; Luckhardt, S.; Paoletti, F.; von Goeler, S.; Bernabei, S. et al.
Partner: UNT Libraries Government Documents Department

The effects of plasma deformability on the feedback stabilization of axisymmetric modes in tokamak plasmas

Description: The effects of plasma deformability on the feedback stabilization of axisymmetric modes of tokamak plasmas are studied. It is seen that plasmas with strongly shaped cross sections have unstable motion different from a rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the eigenfunction in a way that reduces the stabilizing eddy currents in these conductors. Passive feedback results using several equilibria of varying shape are presented. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the position of the flux loops which are used to determine plasma vertical position for the active feedback system. The variations of these non-rigid components of the eigenfunction always serve to reduce the stabilizing effect of the active feedback system by reducing the measurable poloidal flux at the flux-loop locations. Active feedback results are presented for the PBX-M tokamak configuration.
Date: January 1, 1992
Creator: Ward, D. J. & Jardin, S. C.
Partner: UNT Libraries Government Documents Department