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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

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

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

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

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

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

Dependence of CIT (Compact Ignition Tokamak) PF (poloidal field) coil currents on profile and shape parameters using the Control Matrix

Description: The plasma shaping flexibility of the Compact Ignition Tokamak (CIT) poloidal field (PF) coil set is demonstrated through MHD equilibrium calculations of optimal PF coil current distributions and their variation with poloidal beta, internal inductance, plasma 95% elongation, and 95% triangularity. Calculations of the magnetic stored energy are used to compare solutions associated with various plasma parameters. The Control Matrix (CM) equilibrium code, together with the nonlinear equation and numerical optimization software packages HYBRD, and VMCON, respectively, are used to find equilibrium coil current distributions for fixed divertor geometry, volt-seconds, and plasma profiles in order to isolate the dependence on individual parameters. A reference equilibrium and coil current distribution are chosen, and correction currents dI are determined using the CM equilibrium method to obtain other specified plasma shapes. The reference equilibrium is the {kappa} = 2 divertor at beginning of flattop (BOFT) with a minimum stored energy solution for the coil current distribution. The pressure profile function is fixed.
Date: January 1, 1990
Creator: Strickler, D.J.; Peng, Y-K.M. (Oak Ridge National Lab., TN (USA)); Jardin, S.C. & Pomphrey, N. (Princeton Univ., NJ (USA). Plasma Physics Lab.)
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

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

The CIT physics program

Description: The mission of the CIT device is to determine the physics behavior of self-heated fusion plasmas, and to demonstrate the production of substantial amounts of fusion power. In order to achieve this mission at minimum risk and cost, CIT is designed to be a high field, compact, copper-coil device, of modest pulse length. The best measure of extrapolation in confinement properties is the dimensionless parameter {omega}{sub c}{tau}{sub E}. CIT is projected to stand midway between JET and ITER in this parameter, and so represents a relatively modest step. Nonetheless, because nT{tau}{sub E} {alpha} B for dimensionlessly similar devices (aB{sup 4/5} = const.). CIT should have about the same nT{tau}{sub E} as ITER, {minus}10x that of JET. Standard assumptions on confinement, impurity levels, and profile shapes project to Q = 25, with 20 MW of heating power, corresponding to {beta} = 3% (=21/aB) and a total fusion output power of 500 MW. Even given pessimistic projection assumptions. CIT can achieve its basic mission to determine the confinement physics, operational limits, and {alpha}-particle dynamics of self-heated fusion plasmas with {alpha} power greater than auxiliary heating power, while producing more than 100 MW of fusion power. In order to reach these conditions CIT will also demonstrate heating, fueling, and plasma handling techniques necessary to produce reactorlike power density, self-heated fusion plasmas. 3 refs., 3 figs.
Date: January 1, 1990
Creator: Goldston, R.; Bateman, G.; Bell, M.; Colestock, P.; Jardin, S.; Medley, S. et al.
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

The CIT (compact ignition tokamak) pellet injection system: Description and supporting research and development

Description: The Compact Ignition Tokamak (CIT) will use an advance, high-velocity pellet injection system to achieve and maintain ignited plasmas. Two pellet injectors are provided: a moderate-velocity (1-to 1.5-km/s), single-stage pneumatic injector with high reliability and a high-velocity (4- to 5-km/s), two-stage pellet injector that uses frozen hydrogenic pellets encased in sabots. Both pellet injectors are qualified for operation with tritium feed gas. Issues such as performance, neutron activation of injector components, maintenance, design of the pellet injection vacuum line, gas loads to the reprocessing system, and equipment layout are discussed. Results and plans for supporting research and development (R and D) in the areas of tritium pellet fabrication and high-velocity, repetitive two-stage pneumatic injectors are presented. 7 refs., 4 figs., 2 tabs.
Date: January 1, 1989
Creator: Gouge, M.J.; Combs, S.K.; Fisher, P.W. & Milora, S.L.
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

Electromagnetic loads and structural response of the CIT (Compact Ignition Tokamak) vacuum vessel to plasma disruptions

Description: Studies of the electromagnetic loads produced by a variety of plasma disruptions, and the resulting structural effects on the compact Ignition Tokamak (CIT) vacuum vessel (VV), have been performed to help optimize the VV design. A series of stationary and moving plasmas, with disruption rates from 0.7--10.0 MA/ms, have been analyzed using the EMPRES code to compute eddy currents and electromagnetic pressures, and the NASTRAN code to evaluate the structural response of the vacuum vessel. Key factors contributing to the magnitude of EM forces and resulting stresses on the vessel have been found to include disruption rate, and direction and synchronization of plasma motion with the onset of plasma current decay. As a result of these analyses, a number of design changes have been made, and design margins for the present 1.75 meter design have been improved over the original CIT configuration. 1 ref., 10 figs., 4 tabs.
Date: January 1, 1987
Creator: Salem, S.L.; Listvinsky, G.; Lee, M.Y. & Bailey, C.
Partner: UNT Libraries Government Documents Department

Configuration development of a hydraulic press for preloading the toroidal field coils of the Compact Ignition Tokamak

Description: The Fusion Engineering Design Center (FEDC) is part of a national design team that is developing the conceptual design of the Compact Ignition Tokamak (CIT). To achieve a compact device with the minimum major radius, a vertical preload system is being developed to react the vertical separating force normally carried by the inboard leg of the toroidal field (TF) coils. The preload system is in the form of a hydraulic press. Challenges in the design include the development of hydraulic and structural systems for very large force requirements, which could interface with the CIT machine, while allowing maximum access to the top, bottom, and radial periphery of the machine. Maximum access is necessary for maintenance, diagnostics, instrumentation, and control systems. Materials used in the design must function in the nuclear environment and in the presence of high magnetic fields. The structural system developed is an arrangement in which the CIT device is installed in the jaws of the press. Large built-up beams above and below the CIT span the machine and deliver the vertical force to the center cylinder formed by the inboard legs of the TF coils. During the conceptual design study, the vertical force requirement has ranged between 25,000 and 52,000 t. The access requirement on top and bottom limits the width of the spanning beams. Nonmagnetic steel materials are also required because of operation in the high magnetic fields. In the hydraulic system design for the press, several options are being explored. These range from small-diameter jacks operating at very high pressure (228 MPa (33 ksi)) to large-diameter jacks operating at pressures up to 69 MPa (10 ksi). Configurations with various locations for the hydraulic cylinders have also been explored. The nuclear environment and maintenance requirements are factors that affect cylinder location. This paper presents the configuration ...
Date: January 1, 1987
Creator: Lee, V.D.
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

Design of a tritium-compatible vacuum pumping system for the Compact Ignition Tokamak

Description: The conceptual design for the Compact Ignition Tokamak (CIT) vacuum pumping system features high-speed, magnetic-bearing turbomolecular pumps (TMPs), metal-sealed scroll pumps for roughing and backing, and all-metal valves and flange seals. Because the plasma chamber exhaust is handled in a throughput instead of hold-up fashion with no organic seal or lubricating materials exposed to the vacuum stream, inventories of tritium, which are vulnerable to release during an accident and which inhibit maintenance of the vacuum pumping equipment, are minimized. To achieve an initial base pressure of 1.3 /times/ 10/sup /minus/6/ Pa in the plasma chamber, the design includes a large vacuum pumping duct and multiple high-speed TMPs arranged in two stages. The design studies discussed in this paper examine the feasibility and cost impact of providing a low-tritium-inventory, easily maintained vacuum pumping system for a deuterium-tritium (D-T) burning tokamak. 7 refs., 3 figs., 2 tabs.
Date: January 1, 1987
Creator: Haines, J.R.
Partner: UNT Libraries Government Documents Department

Ex-vessel remote maintenance design for the Compact Ignition Tokamak

Description: The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology to carry out maintenance operations on auxiliary machine components. These operations consist of removing and repairing components such as diagnostics and radio frequency (rf) heating modules using remotely operated maintenance equipment. The major equipment that is being developed to accomplish maintenance external to the plasma chamber includes the bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid rad-waste handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design have been addressed in fiscal year (FY) 1987. These included conceptual designs of manipulator systems, the start of a remote equipment research and development (R and D) program, and definition of the hot cell, decon, and equipment repair facility requirements. The manipulator work included investigating transporters and viewing/lighting subsystems. In each case, existing commercial units are being assessed initially, along with viable alternative approaches. R and D work also included demonstrations of remote handling operations on full-size, partial mock-ups of the CIT machine at the Oak Ridge National Laboratory (ORNL) Remote Operations and Maintenance Development Facility.
Date: January 1, 1987
Creator: Spampinato, P.T. & Macdonald, D.
Partner: UNT Libraries Government Documents Department

Electron cyclotron heating and current drive in toroidal geometry

Description: The Principal Investigator has continued to work on problems associated both with the deposition and with the emission of electron cyclotron power in toroidal plasmas. We have investigated the use of electron cyclotron resonance heating for bringing compact tokamaks (BPX) to ignition-like parameters. This requires that we continue to refine the modeling capability of the TORCH code linked with the BALDUR 1 {1/2} D transport code. Using this computational tool, we have examined the dependence of ignition on heating and transport employing both theoretical (multi-mode) and empirically based transport models. The work on current drive focused on the suppression of tearing modes near the q = 2 surface and sawteeth near the q = 1 surface. Electron cyclotron current drive in CIT near the q =2 surface was evaluated for a launch scenario where electron cyclotron power was launched near the equatorial plane. The work on suppression of sawteeth has been oriented toward understanding the suppression that has been observed in a number of tokamaks, in particular, in the WT-3 tokamak in Kyoto. To evaluate the changes in current profile (shear) near the q =1 surface, simulations have been carried out using the linked BALDUR-TORCH code. We consider effects on shear resulting both from wave-induced current as well as from changes in conductivity associated with changes in local temperature. Abstracts and a paper relating to this work is included in Appendix A.
Date: November 1, 1991
Creator: Kritz, A.H.
Partner: UNT Libraries Government Documents Department

Plasma shape control calculations for BPX divertor design

Description: The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs.
Date: January 1, 1991
Creator: Strickler, D.J.; Neilson, G.H. (Oak Ridge National Lab., TN (United States)); Jardin, S.C. & Pomphrey, N. (Princeton Univ., NJ (United States). Plasma Physics Lab.)
Partner: UNT Libraries Government Documents Department

Experimental and theoretical research in applied plasma physics

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: January 1, 1992
Creator: Porkolab, M.
Partner: UNT Libraries Government Documents Department