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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

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

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 (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

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 1. quarter -- FY 1989: Confinement systems programs

Description: Brief summaries are given for DIII-D Research Operations covering the following areas: beta and stability; confinement; boundary physics; electron cyclotron heating; ion Bernstein wave heating; current drive; tokamak operations; neutral beam operations; ECH operations; ICH operations; computer data systems; program development; and hardware development. The progress summaries on the International Cooperation task are given for the Tora Supra, HIDEX -- Nagoya Tokamak Experiment, ASDEX, JET, JFT-2M, CHS, and JT-60. Finally a brief summary of progress on the CIT physics task is given.
Date: January 20, 1989
Partner: UNT Libraries Government Documents Department

Princeton Plasma Physics Laboratory

Description: This report discusses the following topics: principal parameters achieved in experimental devices fiscal year 1990; tokamak fusion test reactor; compact ignition tokamak; Princeton beta experiment- modification; current drive experiment-upgrade; international collaboration; x-ray laser studies; spacecraft glow experiment; plasma processing: deposition and etching of thin films; theoretical studies; tokamak modeling; international thermonuclear experimental reactor; engineering department; project planning and safety office; quality assurance and reliability; technology transfer; administrative operations; PPPL patent invention disclosures for fiscal year 1990; graduate education; plasma physics; graduate education: plasma science and technology; science education program; and Princeton Plasma Physics Laboratory reports fiscal year 1990.
Date: January 1, 1990
Partner: UNT Libraries Government Documents Department

A CO/sub 2/ laser Thomson scattering diagnostic for fusion product alpha particle measurement

Description: A description of a CO/sub 2/ laser Thomson scattering diagnostic for fusion alpha particles is presented. Scattering calculations based on CIT plasma parameters are presented and compared to previous work based on TFTR parameters. Systems components are described and a proof-of-principle test in a nonburning plasma is discussed. 8 refs., 3 figs., 1 tab.
Date: January 1, 1988
Creator: Richards, R.K.; Bennett, C.A.; Fletcher, L.K.; Hunter, H.T. & Hutchinson, D.P.
Partner: UNT Libraries Government Documents Department

Physics guidelines for the Compact Ignition Tokamak

Description: The goal of the Compact Ignition Tokamak (CIT)d program is to provide a cost-effective route to the production of a burning deuterium-tritium plasma, so that alpha-particle effects may be studied. A key issue to be studied in the CIT is whether alpha power behaves like other power sources in affecting tokamak plasma confinement. The program is managed by the Princeton Physics Laboratory and includes broad community involvement. Guidelines for the preliminary design effort have been provided by the Ignition Technical Oversight Committee in discussion with the tokamak community. The reference design is a tokamak with a high filed (10 T), high current (10 MA), poloidal divertor, and liquid-nitrogen-cooled coils. It is a small, high-power-density device of the type proposed by Bruno Coppi (MIT). It has a major radius of 1.23 m, a minor radius of 0.43 m, and plasma elipticity of 1.8. This paper reviews the aims of the program and the basis for the physics guidelines. The role of the CIT in the longer-term tokamak program is briefly discussed. 23 refs., 9 figs., 1 tab.
Date: January 1, 1986
Creator: Sheffield, J.; Dory, R.A.; Houlberg, W.A.; Uckan, N.A.; Bell, M.; Colestock, P. et al.
Partner: UNT Libraries Government Documents Department

Heating the Compact Ignition Tokamak (CIT)

Description: The proposed CIT starts operation in the late 1990's with 20 MW of rf heating power. The tokamak and facility are to be designed to accommodate 50 MW auxiliary heating. The heating methods new being considered are ion cyclotron heating (ICH) and electron cyclotron heating (ECH). Aspects of these systems are described, and the choice of power level and type is discussed. 18 refs.
Date: November 1, 1989
Creator: Ignat, D.W.
Partner: UNT Libraries Government Documents Department

Improved numerical grid generation techniques for the B2 edge plasma code

Description: Techniques used to generate grids for edge fluid codes such as B2 from numerically computed equilibria are discussed. Fully orthogonal, numerically derived grids closely resembling analytically prescribed meshes can be obtained. But, the details of the poloidal field can vary, yielding significantly different plasma parameters in the simulations. The magnitude of these differences is consistent with the predictions of an analytic model of the scrape-off layer. Both numerical and analytic grids are insensitive to changes in their defining parameters. Methods for implementing nonorthogonal boundaries in these meshes are also presented; they differ slightly from those required for fully orthogonal grids.
Date: June 1, 1992
Creator: Stotler, D.P. & Coster, D.P.
Partner: UNT Libraries Government Documents Department

Ex-vessel remote maintenance development plans for the Burning Plasma Experiment

Description: Remote maintenance (RM) is fundamental to the basic design requirements of the Burning Plasma Experiment (BPX), and an extensive RM development and demonstration program is planned to meet these requirements. The program first draws from the experience base that exists in the fission community and Europe's Joint European Torus (JET) Project. Successful solutions are applied where possible and, in many cases, improved in order to achieve the performance demanded by a multiyear program that must be capable of efficiently executing RM procedures. Early, concurrent efforts in the design and fabrication of prototype remote handling (RH) equipment, remote tooling, and maintainable machine components will precede an extensive use of mock-up equipment in order to test, develop, and demonstrate the technology. 7 refs,. 5 figs.
Date: January 1, 1991
Creator: Burgess, T.W. & Davis, F.C.
Partner: UNT Libraries Government Documents Department

BPX commitment to total remote maintenance

Description: The Burning Plasma Experiment (BPX), to be located at Princeton Plasma Physics Laboratory, is the next major experimental machine in the US Fusion Program. It will be fueled with deuterium-tritium (D-T) that, when burned, will generate high-energy neutrons. This will activate the various materials used in construction of the machine, which will result in high levels of gamma radiation. Any subsequent maintenance activities on the machine or in the test cell area must be performed remotely. The initial criteria for BPX assumed that failure of toroidal field (TF) coil or poloidal field (PF) coil was an unlikely event. Therefore, no provisions were made for remote replacement. Expected failures were limited to the plasma-facing components and the external auxiliary equipment such as heating systems and diagnostics. Recent coil failures experienced at the Tokamak Fusion Test Reactor (TFTR), the Joint European Torus (JET), JT-60, and Tore Supra caused the BPX project staff to reconsider the need for remote replacement. A study was undertaken to investigate how the project would be affected if the capability to recover from a coil failure were required. Potential effects including configuration changes to the machine and facility, project cost, and project operation were considered. The study revealed that it is indeed feasible to design BPX for remote recovery from any coil failure. However, for this to be accomplished effectively, it is imperative to incorporate the necessary remote maintenance features of the components to be remotely replaced into the original design along with all of the other functional features. The remote maintenance capability cannot be retrofitted after the design is complete or the equipment is built. This paper discusses the impacts of the coil remote replacement study and the subsequent changes to the design. 4 figs., 1 tab.
Date: January 1, 1991
Creator: Davis, F.C. & Burgess, T.W.
Partner: UNT Libraries Government Documents Department

BPX insulation irradiation program test results

Description: The toroidal field coil insulation for the Burning Plasma Experiment (BPX) is expected to receive a radiation dose of nearly 10{sup 10} rad and to withstand significant mechanical stresses. An irradiation test program was performed at the Idaho National Engineering Laboratory (INEL) using the Advanced Technology Reactor (ATR) for irradiations to doses on the order of 3 {times} 10{sup 10} rad. The flexure and shear strength with compression of commercially procured sheet material were reported earlier. A second series of tests has been performed to slightly higher dose levels with vacuum impregnated materials, glass strand material, and Spaulrad-S sheet samples. Vacuum impregnation with a Shell 9405 resin and 9470 hardener was used to produce bonded copper squares and flexure samples of both pure resin and resin with S-glass. A new test fixture was developed to test the bonded samples in shear without applied compression. The Spaulrad-S flexure samples demonstrated a loss of strength with irradiation, similar to previous results. The pure resin lost nearly all flexibility, while the S-glass-reinforced samples retained between 30% and 40% of the initial flexure strength. The S-glass strands showed a 30% loss of strength at the higher dose level when tested in tension. The bonded copper squares had a low room-temperature shear strength of approximately 17 MPa before irradiation, which was unchanged in the irradiated samples. Shear testing of unirradiated bonded copper squares with ten different types of surface treatment revealed that the low shear strength resulted from the polyurethane primer used. In the later series of test, the epoxy-based primers and DZ-80 from Ciba-Geigy did much better, with shear strengths on the order of 40 MPa. These samples also demonstrated a resistance to cryogenic shock. One irradiated bonded sample was tested up 10 210 MPa in compression, the limit of the test fixture, without ...
Date: January 1, 1991
Creator: McManamy, T.J. (Oak Ridge National Lab., TN (United States)); Kanemoto, G. (EG and G Idaho, Inc., Idaho Falls, ID (United States)) & Snook, P.G. (Princeton Univ., NJ (United States). Plasma Physics Lab.)
Partner: UNT Libraries Government Documents Department

Overview of the compact ignition tokamak

Description: A national team has developed a baseline concept for a Compact Ignition Tokamak (CIT). The CIT mission is to achieve ignition and provide experimental capability to study the behavior of burning plasma. The design uses large magnetic fields on axis (about 10 T) and large plasma currents (about 9-10 MA). The magnet structure derives high strength from the use of a copper-Inconel composite plate design in the nose of region of the toroidal field (TF) coil and in the ohmic heating solenoid. Inertial cooling is used;liquid nitrogen temperatures are established at the beginning of each pulse. Capability is provided to operate either with a divertor or limiter based plasma. The design is very compact (1.32-m major radius, 0.43-m plasma radius), has 16 TF coils, and has 16 major horizontal access ports, about 30 cm by 80 cm, located between TF coils. The schedule is for a construction project to be authorized for the period FY 1988-93.
Date: January 1, 1986
Creator: Flanagan, C.A.
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

FY90 milestone report for the CIT (Compact Ignition Tokamak) project: Localizability of electron-cyclotron heating power

Description: Estimates of the localizability of electron-cyclotron heating power are made for the Compact Ignition Tokamak. A particular heating scenario is examined, namely, the fundamental O-mode, injected nearly perpendicular to the toroidal magnetic field. The absorption depth due to finite T{sub e} is very small, about 1 cm, near the q = 2 surface. Absorption is even better localized near q = 1. Several issues that might lead to degraded localizability are reviewed. Use of an intense, pulsed microwave source is the only issue with a possibly significant impact. 3 refs.
Date: October 25, 1990
Creator: Smith, G.R.
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

Damping of electron cyclotron waves in dense plasmas of a compact ignition tokamak

Description: Absorption of electromagnetic waves by hot and dense plasmas is investigated in the electron cyclotron range of frequency. It is shown that the strong reduction of the damping of the extraordinary mode, caused by finite Larmor radius effects on waves propagating perpendicularly to the magnetic field, becomes insignificant at large values of the parallel component of the refractive index. With an appropriate form of the relativistic dispersion relation which includes high order Larmor radius terms, heating of dense plasmas in a Compact Ignition Tokamak is investigated. It is shown that by using the extraordinary mode with oblique propagation and frequency of 190 GHz it is possible to bring to thermonuclear ignition a dense ohmic plasma with a toroidal magnetic field of 105 kG and a central density of 1 x 10/sup 15/ cm/sup -3/. 11 refs., 11 figs.
Date: June 1, 1987
Creator: Mazzucato, E.; Fidone, I. & Granata, G.
Partner: UNT Libraries Government Documents Department

Operating conditions of the BPX divertor

Description: In this paper we discuss the expected operating conditions at the divertor of the BPX tokamak (Burning Plasma Experiment), the next- step US tokamak proposed for the study of self-heated plasmas at Q {approx equal} 5 to ignition. In this double-null device ({kappa} {approx equal} 2), the predicted first-wall loading is high because of is compact size (R = 2.6m, {alpha} = 0.8m, I{sup p} = 10.6 MA, and B{sub T}) and its high projected fusion power output (100--500 MW with up to 20 MW of ICRH). Present designs call for inertially cooled carbon-based target plate material and X-point sweeping to handle the divertor heat flux during the 3--5 s flat-top at full power. The X-point is maintained about 15--20 cm off the target plates (a distance of {approximately}5m along field lines), which represents a reasonable compromise between lowering the divertor electron temperature (T{sub e,d}) by increasing the connection length, and lowering the peak divertor heat flux ({cflx q}{sub d}) by increasing the magnetic flux expansion (which is about 15--20 in this case). It is planned for the BPX device to operate with H-mode confinement; ELMs are expected because of the relatively high power flow through the edge plasma (P{sub sep} {approx equal} 0.6 MW/m{sup 2} for P{sub fus} = 500 MW). The ELMs will help reduce the impurity concentration in the core plasma (Z{sub eff} {approx equal} 1.7) and keep the density down, but should not add significantly to the divertor heat flux since their measured contribution to the global power balance drops with increasing input power.
Date: January 1, 1991
Creator: Hill, D.N.; Milovich, J.; Rognlien, T. (Lawrence Livermore National Lab., CA (USA)); Braams, B.J. (New York Univ., NY (USA)); Brooks, J.N. (Argonne National Lab., IL (USA)); Campbell, R. (Sandia National Labs., Albuquerque, NM (USA)) et al.
Partner: UNT Libraries Government Documents Department