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Sample records for ignition tokamak design

  1. Ignition probabilities for Compact Ignition Tokamak designs

    International Nuclear Information System (INIS)

    Stotler, D.P.; Goldston, R.J.

    1989-09-01

    A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs

  2. Configuration and structural design of Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Brown, T.G.

    1985-01-01

    Viewgraphs are presented on the configuration and structural design of the Compact Ignition Tokamak, originally presented to the US/Japan Workshop on Next Step Machine Design. Items discussed in this presentation include: PPPL 0424 ref design; MIT LITE ref design; IGNITOR 1.01 M ref design; and IGNITOR 1.08 M press configuration

  3. Pulse length assessment of compact ignition tokamak designs

    International Nuclear Information System (INIS)

    Stotler, D.P.; Pomphrey, N.

    1989-07-01

    A time-dependent zero-dimensional code has been developed to assess the pulse length and auxiliary heating requirements of Compact Ignition Tokamak (CIT) designs. By taking a global approach to the calculation, parametric studies can be easily performed. The accuracy of the procedure is tested by comparing with the Tokamak Simulation Code which uses theory-based thermal diffusivities. A series of runs is carried out at various levels of energy confinement for each of three possible CIT configurations. It is found that for cases of interest, ignition or an energy multiplication factor Q /approxreverse arrowgt/ 7 can be attained within the first half of the planned five-second flattop with 10--40 MW of auxiliary heating. These results are supported by analytic calculations. 18 refs., 7 figs., 2 tabs

  4. TIBER: Tokamak Ignition/Burn Experimental Research. Final design report

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Barr, W.L.

    1985-01-01

    The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs

  5. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

  6. Recent progress on the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Ignat, D.W.

    1987-01-01

    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

  7. Tokamak and RFP ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

  8. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

  9. Plasma transport in a compact ignition tokamak

    International Nuclear Information System (INIS)

    Singer, C.E.; Ku, L.P; Bateman, G.

    1987-02-01

    Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the 3 He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved

  10. Options for an ignited tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed

  11. Ignition experiment in a single-turn-coil tokamak

    International Nuclear Information System (INIS)

    Carrera, R.; Driga, M.; Gully, J.H.

    1989-01-01

    A novel concept for a fusion ignition experiment, IGNITEX proposed along the lines of previous ideas for a compact thermonuclear device is analyzed. A single-turn-coil tokamak is analyzed. A single-turn-coil tokamak supplied by homopolar generators can ohmically heat a DT plasma to ignition conditions and maintain a thermally stable ignited phase for about ten energy confinement times. The IGNITEX experiment can provide a simple and relatively inexpensive way to produce and control ignited plasmas for scientific study

  12. Compact Ignition Tokamak conventional facilities optimization

    International Nuclear Information System (INIS)

    Commander, J.C.; Spang, N.W.

    1987-01-01

    A high-field ignition machine with liquid-nitrogen-cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed for the next phase of the United States magnetically confined fusion program. A team of national laboratory, university, and industrial participants completed the conceptual design for the CIT machine, support systems and conventional facilities. Following conceptual design, optimization studies were conducted with the goal of improving machine performance, support systems design, and conventional facilities configuration. This paper deals primarily with the conceptual design configuration of the CIT conventional facilities, the changes that evolved during optimization studies, and the revised changes resulting from functional and operational requirements (F and ORs). The CIT conventional facilities conceptual design is based on two premises: (1) satisfaction of the F and ORs developed in the CIT building and utilities requirements document, and (2) the assumption that the CIT project will be sited at the Princeton Plasma Physics Laboratory (PPPL) in order that maximum utilization can be made of existing Tokamak Fusion Test Reactor (TFTR) buildings and utilities. The optimization studies required reevaluation of the F and ORs and a second look at TFTR buildings and utilities. Some of the high-cost-impact optimization studies are discussed, including the evaluation criteria for a change from the conceptual design baseline configuration. The revised conventional facilities configuration are described and the estimated cost impact is summarized

  13. Structural features of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Citrolo, J.; Brown, G.; Rogoff, P.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is undergoing preliminary structural design and definitions. It will be relatively inexpensive with ignition capabilities. During the definition phase it was concluded that the TF coil should be assembled from the laminate copper-Inconel plates since copper alone cannot sustain the expected magnetic and thermal loads. An extensive test program is being initiated to investigate the various materials, and their elastic and inelastic response and to develop the constitutive equations required for the selection of design criteria and for the stress analysis of this device. Finite element analysis nonlinear material capabilities are being used to study, predict and correlate the machine behavior

  14. Physics aspects of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Post, D.; Bateman, G.; Houlberg, W.

    1986-11-01

    The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha-particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve the condition of ntau/sub E/ ∼ 2 x 10 20 sec m -3 required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k ∼ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided

  15. TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Gilleland, J.R.

    1988-01-01

    The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

  16. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    Johner, J.

    1989-04-01

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  17. Ignited tokamak devices with ohmic-heating dominated startup

    International Nuclear Information System (INIS)

    Cohn, D.R.; Bromberg, L.; Jassby, D.L.

    1986-01-01

    Startup of tokamaks such that the auxiliary heating power is significantly less than the ohmic heating power at all times during heating to ignition can be referred to as ''Ohmic-heating dominated startup.'' Operation in this mode could increase the certainty of heating to ignition since energy confinement during startup may be described by present scaling laws for ohmic heating. It could also reduce substantially the auxiliary heating power (the required power may be quite large for auxiliary-heating dominated startup). These advantages might be realized without the potentially demanding requirements for pure ohmic heating to ignition. In this paper the authors discuss the requirements for ohmic-heating dominated startup and present illustrative design parameters for compact experiment ignition devices that use high performance copper magnets

  18. 'Crescent'-shaped tokamak for compact ignition

    International Nuclear Information System (INIS)

    Yamazaki, K.; Reiersen, W.T.

    1985-12-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross-section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design dose not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  19. 'Crescent'-shaped tokamak for compact ignition

    International Nuclear Information System (INIS)

    Yamazaki, K.; Reiersen, W.T.

    1986-01-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design does not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  20. Rotating shield ceiling for the compact ignition tokamak test cell

    International Nuclear Information System (INIS)

    Commander, J.C.

    1986-01-01

    For the next phase of the United States fusion program, a compact, high-field, toroidal ignition machine with liquid nitrogen cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed. The CIT machine will be housed in a test cell with design features developed during preconceptual design. Configured as a right cylinder, the selected test cell design features: a test cell and basement with thick concrete shielding walls, and floor; leak tight tritium seals; and operational characteristics well suited to the circular CIT machine configuration and radially oriented ancillary equipment and systems

  1. Overview of tritium systems for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Bartlit, J.R.; Gruetzmacher, K.M.; Fleming, R.B.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is being designed at several laboratories to produce and study fully ignited plasma discharges. The tritium systems which will be needed for CIT include fueling systems and radiation monitoring and safety systems. Design of the tritium systems is the responsibility of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. Major new tritium systems for CIT include a pellet injector, an air detritiation system and a glovebox atmosphere detritiation system. The pellet injector is being developed at Oak Ridge National Laboratory. 7 refs., 2 figs

  2. System studies of compact ignition tokamaks

    International Nuclear Information System (INIS)

    Galambos, J.D.; Peng, Y.K.M.; Blackfield, D.T.

    1986-01-01

    A new version of the FEDC Tokamak System Code (TSC) has been developed to analyze the Compact Ignition Tokamak (CIT). These proposed experiments have small (major radius F 1.5m) and high magnetic fields (B J 10T), and are characterized by reduced cost. Key design constraints of CIT include limits to the high stress levels in the magnetic coils, limits to the large temperature rises in the coils and on the first wall or divertor plate, minimizing power supply requirements, and assuring adequate plasma performance in fusion ignition and burn time consistent with the latest physics understanding. We present systems code level studies of CIT parameter space here for a range of design options with various design constraints. The present version of the TSC incorporates new models for key components of CIT. For example, new algorithms have been incorporated for calculating stress levels in the TFC and ohmic solenoid, temperature rise in the magnetic coils, peak power requirements, plasma MHD equilibrium and volt-second capability. The code also incorporates a numerical optimizer to find combinations of engineering quantities (device size, coil sizes, coil current densities etc.) and physics quantities (plasma density temperature, and beta, etc.) which satisfy all the constraints and can minimize or maximize a figure of merit (e.g., the major radius). This method was recently used in a mirror reactor system code (3) for the Minimara concept development

  3. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

    1985-01-01

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/ 2 /q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs

  4. Confinement scaling and ignition in tokamaks

    International Nuclear Information System (INIS)

    Perkins, F.W.; Sun, Y.C.

    1985-10-01

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10 15 cm -3 , high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition

  5. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  6. Whist code calculations of ignition margin in an ignition tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    A simple global model was developed to determine the ignition margin of tokamaks including electron and ion conduction losses. A comparison of this model with results from a 1 1/2 dimensional Whist code is made

  7. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  8. Maintenance concept development for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Macdonald, D.

    1988-01-01

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs

  9. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  10. Plasma diagnostics for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Medley, S.S.; Young, K.M.

    1988-06-01

    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 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

  11. Probability of ignition - a better approach than ignition margin

    International Nuclear Information System (INIS)

    Ho, S.K.; Perkins, L.J.

    1989-01-01

    The use of a figure of merit - the probability of ignition - is proposed for the characterization of the ignition performance of projected ignition tokamaks. Monte Carlo and analytic models have been developed to compute the uncertainty distribution function for ignition of a given tokamak design, in terms of the uncertainties inherent in the tokamak physics database. A sample analysis with this method indicates that the risks of not achieving ignition may be unacceptably high unless the accepted margins for ignition are increased. (author). Letter-to-the-editor. 12 refs, 2 figs, 2 tabs

  12. Design of an ion cyclotron resonance heating system for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Yugo, J.J.; Goranson, P.L.; Swain, D.W.; Baity, F.W.; Vesey, R.

    1987-01-01

    The Compact Ignition Tokamak (CIT) requires 10-20 MW of ion cyclotron resonance heating (ICRH) power to raise the plasma temperature to ignition. The initial ICRH system will provide 10 MW of power to the plasma, utilizing a total of six rf power units feeding six current straps in three ports. The systems may be expanded to 20 MW with additional rf power units, antennas, and ports. Plasma heating will be achieved through coupling to the fundamental ion cyclotron resonance of a 3 He minority species (also the second harmonic of tritium). The proposed antenna is a resonant double loop (RDL) structure with vacuum, shorted stubs at each end for tuning and impedance matching. The antennas are of modular, compact construction for installation and removal through the midplane port. Remote maintainability and the reactorlike operating environment have a major impact on the design of the launcher for this machine. 6 refs., 7 figs., 5 tabs

  13. Physics parameter space of tokamak ignition devices

    International Nuclear Information System (INIS)

    Selcow, E.C.; Peng, Y.K.M.; Uckan, N.A.; Houlberg, W.A.

    1985-01-01

    This paper describes the results of a study to explore the physics parameter space of tokamak ignition experiments. A new physics systems code has been developed to perform the study. This code performs a global plasma analysis using steady-state, two-fluid, energy-transport models. In this paper, we discuss the models used in the code and their application to the analysis of compact ignition experiments. 8 refs., 8 figs., 1 tab

  14. Ex-vessel remote maintenance design for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology for ex-vessel maintenance operations on auxiliary machine components. These operations consist of repairing and replacing components such as diagnostic, radio-frequency (rf) heating, and fueling systems using remotely operated maintenance equipment in the test cell. In addition, ex-vessel maintenance design also includes developing hot cell facilities for equipment decontamination, repair, and solid radioactive waste handling. The test cell maintenance philosophy is markedly influenced by the neutron/gamma shield surrounding the machine that allows personal access into the test cell one day after shutdown. Hence, maintenance operations can be performed hands-on in the test cell with the shield intact and must be remotely performed when the shield is disassembled for machine access. The constricted access to the auxiliary components of the machine affect the design requirements for the maintenance equipment and impose major spatial constraints. Several major areas of the maintenance system design are being addressed in fiscal year 1987. These include conceptual design of the manipulator system, preliminary remote equipment research and development, and definition of the hot cell, decontamination, and equipment repair facility requirements. The manipulator work includes investigating transporters and viewing/lighting subsystems. 2 figs

  15. Elements of a method to scale ignition reactor Tokamak

    International Nuclear Information System (INIS)

    Cotsaftis, M.

    1984-08-01

    Due to unavoidable uncertainties from present scaling laws when projected to thermonuclear regime, a method is proposed to minimize these uncertainties in order to figure out the main parameters of ignited tokamak. The method mainly consists in searching, if any, a domain in adapted parameters space which allows Ignition, but is the least sensitive to possible change in scaling laws. In other words, Ignition domain is researched which is the intersection of all possible Ignition domains corresponding to all possible scaling laws produced by all possible transports

  16. Four ignition TNS tokamak reactor systems: design summary

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1977-10-01

    Principal TNS objectives assumed included: (1) demonstration of ignition and burning dynamics; and (2) reactor technology forcing. The selection of an overall design approach for TNS required an early quantitative assessment of the most important design issues; namely, choice of ignition plasma design conditions (principally size and confining field of axis), and choice of toroidal field coil technology (resistive or superconducting windings). The design space investigated in this study ranged from ignited plasmas (elongated) with minor radii varying between 0.8 m (TFTR-like) and approximately 2.0 m (EPR-like). Four TF coil types were examined; these included copper, NbTi, Nb 3 Sn, and a hybrid design employing nested coils of copper and NbTi. A final step involved a further comparison of the four reference concepts using decision modeling techniques as a mechanism for selecting a preferred design approach for the TNS mission. Section 3.0 describes the TNS study process. Section 4.0 presents a summary of the parameters for the four reference point designs. Finally, Section 5.0 presents a brief description of the design features of many of the systems comprising the TNS design

  17. Ohmic ignition of Neo-Alcator tokamak with adiabatic compression

    International Nuclear Information System (INIS)

    Inoue, Nobuyuki; Ogawa, Yuichi

    1992-01-01

    Ohmic ignition condition on axis of the DT tokamak plasma heated by minor radius and major radius adiabatic compression is studied assuming parabolic profiles for plasma parameters, elliptic plasma cross section, and Neo-Alcator confinement scaling. It is noticeable that magnetic compression reduces the necessary total plasma current for Ohmic ignition device. Typically in compact ignition tokamak of the minor radius of 0.47 m, major radius of 1.5 m and on-axis toroidal field of 20 T, the plasma current of 6.8 MA is sufficient for compression plasma, while that of 11.7 MA is for no compression plasma. Another example with larger major radius is also described. In such a device the large flux swing of Ohmic transformer is available for long burn. Application of magnetic compression saves the flux swing and thereby extends the burn time. (author)

  18. Engineering parameters for four ignition TNS tokamak reactor systems

    International Nuclear Information System (INIS)

    Varljen, T.C.; Gibson, G.; French, J.W.; Heck, F.M.

    1977-01-01

    The ORNL/Westinghouse program for The Next Step (TNS) tokamak beyond TFTR has examined a large number of potential configurations for D-T burning ignition tokamak systems. An objective of this work has been to quantify the trade-offs associated with the assumption of certain plasma physics criteria and toroidal field coil technologies. Four tokamak system point designs are described, each representative of the TF coil technologies considered, to illustrate the engineering features associated with each concept. Point designs, such as the ones discussed herein, have been used to develop component size, performance and cost scaling relationships which have been incorporated in a digital computer code to facilitate an examination of the total design and cost impact of candidate design approaches. The point designs which are described are typical, however, they have not been individually optimized. The options are distinguished by the TF coil technology chosen and include: (1) a high field water-cooled copper TF system, (2) a moderate field NbTi superconducting TF system, (3) a high field Nb 3 Sn superconducting TF system, and (4) a high field hybrid TF system with outer NbTi superconducting windings and inner water-cooled copper windings. Descriptions are provided for the major device components and all major support systems including power supplies, vacuum systems, fuel systems, heat transport and facility systems

  19. Ex-vessel remote maintenance design for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    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

  20. Self-organized ignition of a tokamak plasma

    International Nuclear Information System (INIS)

    Schoepf, K.

    2007-01-01

    The continuous progress in the attainment of plasma parameters required for establishing nuclear fusion in magnetically confined plasmas as well as the prospect of feasible steady-state operation has instigated the interest in the physics of burning plasmas [1]. Aside from the required plasma current drive, fusion energy production with tokamaks demands particular attention to confinement and fuelling regimes in order to maintain the plasma density n and temperature T at favourable values matching with specific requirements such as the triple product nτ E T, where τ E represents the plasma energy confinement time. The identification of state and parameter space regions capable of ignited fusion plasma operation is evidently crucial if significant energy gains are to be realized over longer periods. Examining the time-evolving state of tokamak fusion plasma in a parameter space spanned by the densities of plasma constituents and their temperatures has led to the formation of an ignition criterion [2] fundamentally different from the commonly used static patterns. The incorporation of non-stationary particle and energy balances into the analysis here, the application of a 'soft' Troyon beta limit [3], the consideration of actual fusion power deposition [4,5] and its effect of reducing τ E are seen to significantly influence the fusion burn dynamics and to shape the ignition conditions. The presented investigation refers to a somewhat upgraded (to achieve ignition) ITER-like tokamak plasma and uses volume averages of locally varying quantities and processes. The resulting ignition criterion accounts for the dynamic evolution of a reacting plasma controlled by heating and fuel feeding. Interestingly, also self-organized ignition can be observed: a fusion plasma possessing a density and temperature above a distinct separatrix in the considered parameter phase space is seen to evolve - without external heating and hence practically by itself - towards an ignited

  1. A carbon-carbon panel design concept for the inboard limiter of the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Mantz, H.C.; Bowers, D.A.; Williams, F.R.; Witten, M.A.

    1989-01-01

    The inboard limiter of the Compact Ignition Tokamak (CIT) must protect the vacuum vessel from the plasma energy. This limiter region must withstand nominal heat fluxes in excess of 10 MW/m 2 and in addition it must be designed to be remotely maintained. Carbon-carbon composite material was selected over bulk graphite materials for the limiter design because of its ability to meet the thermal and structural requirements. The structural design concept consists of carbon-carbon composite panels attached to the vacuum vessel by a hinged rod/retainer concept. Results of the preliminary design study to define this inboard limiter are presented. The design concept is described along with the analyses of the thermal and structural response during nominal plasma operation and during plasma disruption events. 2 refs., 8 figs

  2. Design of Tokamak plasma with high Tc superconducting coils

    International Nuclear Information System (INIS)

    Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.

    1999-01-01

    This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)

  3. Relative merits of size, field, and current on ignited tokamak performance

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1988-01-01

    A simple global analysis is developed to examine the relative merits of size (L = a or R/sub 0 /), field (B/sub 0 /), and current (I) on ignition regimes of tokamaks under various confinement scaling laws. Scalings of key parameters with L, B/sub 0 /, and I are presented at several operating points, including (a) optimal path to ignition (saddle point), (b) ignition at minimum beta, (c) ignition at 10 keV, and (d) maximum performance at the limits of density and beta. Expressions for the saddle point and the minimum conditions needed for ohmic ignition are derived analytically for any confinement model of the form tau/sub E/ ∼ n/sup x/T/sup y/. For a wide range of confinement models, the ''figure of merit'' parameters and I are found to give a good indication of the relative performance of the devices where q* is the cylindrical safety factor. As an illustration, the results are applied to representative ''CIT'' (as a class of compact, high-field ignition tokamaks) and ''Super-JETs'' [a class of large-size (few x JET), low-field, high-current (≥20-MA) devices.

  4. Time-dependent simulations of a Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Stotler, D.P.; Bateman, G.

    1988-05-01

    Detailed simulations of the Compact Ignition Tokamak are carried out using a 1-1/2-D transport code. The calculations include time-varying densities, fields, and plasma shape. It is shown that ignition can be achieved in this device if somewhat better than L-mode energy confinement time scaling is possible. We also conclude that the performance of such a compact, short-pulse device can depend greatly on how the plasma is evolved to its flat-top parameters. Furthermore, in cases such as the ones discussed here, where there is not a great deal of ignition margin and the electron density is held constant, ignition ends if the helium ash is not removed. In general, control of the deuterium--tritium density is equivalent to burn control. 48 refs., 15 figs

  5. Energy confinement time and tokamak ignition - a theoretical viewpoint

    International Nuclear Information System (INIS)

    Sigmar, D.J.; Hsu, C.T.

    1989-01-01

    A rigorous approach developed earlier to obtain the global energy confinement time from theory based local thermal transport coefficients is applied to investigate the approach to thermonuclear fusion in the tokamak. Theory leads naturally to a generalized 'offset' form for the global energy confinement time. The limitations of the theoretical paradigm presented here are discussed and specific ignition contour results for a large 5 Tesla and 12 Tesla ignition experiment are given. (orig.) [de

  6. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    Troyon, F.

    1990-10-01

    There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab

  7. Development of a remotely maintainable radio-frequency module for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Snider, J.D.

    1988-01-01

    The Compact Ignition Tokamak (CIT) will require reliable remote handling (RH) systems to overcome failures in diagnostic and operational equipment. Oak Ridge National laboratory (ORNL) is responsible for the ex-vessel remote maintenance systems for the CIT. Part of this effort is performing remote maintenance demonstrations on replicas of various CIT equipment. To ensure successful RH, the machine must be designed with proven remote maintenance features. In the demonstrations, critical remote maintenance features are tested before actual CIT equipment designs are finalized. Designs and procedures required to remotely remove and install a radio-frequency (rf) module from a modplane port on the tokamak were recently demonstrated at ORNL. This testing identified both successful design features for remote maintenance of the rf module and areas that require further development. 1 ref., 11 figs

  8. Radiation analysis of the CIT (Compact Ignition Tokamak) pellet injector system and its impact on personnel access

    Energy Technology Data Exchange (ETDEWEB)

    Selcow, E.C.; Stevens, P.N.; Gomes, I.C.; Gomes, L.M.

    1987-01-01

    Conceptual design of the Compact Ignition Tokamak (CIT) is near completion. This short-pulse ignition experiment is planned to follow the operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The high neutron wall loadings, /approximately/4-5 MW/m/sup 2/, associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of device components and personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure, and the entire experiment is housed in a circular test cell facility with a radius of /approximately/12 m. The most critical radiation concern in the CIT design process relates to the numerous penetrations in the device. This paper discusses the impact of a major penetration on the design and operations of the CIT pellet injection system. The pellet injector is a major component, which has a line-of-sight penetration through the igloo and test cell wall. All current options for maintenance of the injector require personnel access. A nuclear analysis has been performed to determine the feasibility of hands-on access. Results indicate that personnel access to the pellet injector glovebox is possible. 10 refs., 3 figs., 3 tabs.

  9. TSC [Tokamak Simulation Code] disruption scenarios and CIT [Compact Ignition Tokamak] vacuum vessel force evolution

    International Nuclear Information System (INIS)

    Sayer, R.O.; Peng, Y.K.M.; Strickler, D.J.; Jardin, S.C.

    1990-01-01

    The Tokamak Simulation Code and the TWIR postprocessor code have been used to develop credible plasma disruption scenarios for the Compact Ignition Tokamak (CIT) in order to predict the evolution of forces on CIT conducting structures and to provide results required for detailed structural design analysis. The extreme values of net radial and vertical vacuum vessel (VV) forces were found to be F R =-12.0 MN/rad and F Z =-3.0 MN/rad, respectively, for the CIT 2.1-m, 11-MA design. Net VV force evolution was found to be altered significantly by two mechanisms not noted previously. The first, due to poloidal VV currents arising from increased plasma paramagnetism during thermal quench, reduces the magnitude of the extreme F R by 15-50% and modifies the distribution of forces substantially. The second effect is that slower plasma current decay rates give more severe net vertical VV loads because the current decay occurs when the plasma has moved farther from midplane than is the case for faster decay rates. 7 refs., 9 figs., 1 tab

  10. Remote maintenance design activities and research and development accomplishments for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1988-01-01

    The use of deuterium-tritium (D-T) fuel for the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations. The remote operations consist of removing and replacing such components as first wall armor protection tiles, radio-frequency (rf) heating modules, and diagnostic modules. The major pieces of equipment being developed for maintenance activities internal to the vacuum vessel include an articulated boom manipulator (ABM), an inspection manipulator, and special tooling. For activities external to the vessel, the equipment includes a bridge-mounted manipulator system, decontamination equipment, hot cell equipment, and solid radiation-waste (rad-waste) handling and packaging equipment. The CIT Project is completing the conceptual design phase; research and development (R and D) activities, which include demonstrations of remote maintenance operations on full-size partial mock-ups are under way. 5 figs

  11. Remote maintenance design activities and research and development accomplishments for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.

    1989-01-01

    The use of deuterium-tritium (D-T) fuel for the Compact Ignition Tokamak (CIT) requires the use of remote handling technology in order to carry out maintenance operations. The remote operations consist of removing and replacing such components as first wall armor protection tiles, radio-frequency (RF) heating modules, and diagnostic modules. The major pieces of equipment being developed for maintenance operations internal to the vacuum vessel include an articulated boom manipulator (ABM), an inspection manipulator, and special tooling. For operations external to the vessel, the equipment includes a bridge-mounted manipulator system, decontamination equipment, hot cell equipment, and solid radioactive waste (rad-waste) handling and packaging equipment. The CIT Project is completing the conceptual design phase; research and development (R and D) activities, which include demonstrations of remote maintenance operations on full-size partial mock-ups are under way. (orig.)

  12. Reaching ignition in the tokamak

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-06-01

    This review covers the following areas: (1) the physics of burning plasmas, (2) plasma physics requirements for reaching ignition, (3) design studies for ignition devices, and (4) prospects for an ignition project

  13. Plasma position control in a tokamak reactor around ignition

    International Nuclear Information System (INIS)

    Carretta, U.; Minardi, E.; Bacelli, N.

    1986-01-01

    Plasma position control in a tokamak reactor in the phase approaching ignition is closely related to burn control. If ignited burn corresponds to a thermally unstable situation the plasma becomes sensitive to the thermal instability already in the phase when ignition is approached so that the trajectory in the position-pressure (R,p) space becomes effectively unpredictable. For example, schemes involving closed cycles around ignition can be unstable in the heating-cooling phases, and the deviations may be cumulative in time. Reliable plasma control in pressure-position (p, R) space is achieved by beforehand constraining the p, R trajectory rigidly with suitable feedback vertical field stabilization, which is to be established already below ignition. A scheme in which ignition is approached in a stable and automatic way by feedback stabilization on the vertical field is proposed and studied in detail. The values of the gain coefficient ensuring stabilization and the associated p and R excursions are discussed both analytically, with a 0-D approximation including non-linear effects, and numerically with a 1-D code in cylindrical geometry. Profile effects increase the excursions, in particular above ignition. (author)

  14. Effect of sawteeth on alpha power deposition and ignition in Tokamaks

    International Nuclear Information System (INIS)

    Kolesnichenko, Ya.I.; Lutsenko, V.V.; Yakovenko, Yu.V.

    1992-01-01

    The main features of the alpha particle heating, ignition and thermonuclear burn in a tokamak plasma with sawtooth oscillations are revealed. The sensitivity of results against the various model of sawteeth and characteristics of the safety factor q(r) is investigated. Analysis of ignition is best applicable to the case T ε r ,T ε being the alpha particle energy loss time, T r period of sawtooth oscillations

  15. Poloidal magnetics of a divertor compact ignition tokamak

    International Nuclear Information System (INIS)

    Strickler, D.J.; Peng, Y.K.M.; Jardin, S.C.

    1987-10-01

    A technique is presented for calculating bounds on the poloidal field (PF) coil currents required to constrain critical plasma shape parameters when plasma pressure and current density profiles are changed. Such considerations are important in the conceptual design of the PF coils for the Compact Ignition Tokamak (CIT) and their electrical power systems in view of the uncertainty in plasma profiles and operating scenarios. Four relatively independent coil groups are sufficient to find a coil current distribution and equilibrium satisfying a prescribed plasma major radius, minor radius, and divertor strike point coordinates. The variation in the coil current distribution with plasma profiles tends to be large for external PF systems and provides a measure by which coil configurations may be compared. 6 refs., 7 figs., 4 tabs

  16. Implications of rf current drive theory for next step steady-state tokamak design

    International Nuclear Information System (INIS)

    Schultz, J.H.

    1985-06-01

    Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor

  17. Compact Ignition Tokamak Program: R and D needs

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1985-01-01

    This report on the Compact Ignition Tokamak Program supplies information concerning: segmented vacuum vessel joint development; first wall tile attachments; first wall/tile development - composite materials; vacuum leak detection; high frequency rf sources; Faraday shield development; design and testing of rf launchers for high power, ling pulse operation; radiation hardened, low loss, dielectric windows for rf, IR, visible, UV and X-rays, mirrors for changing direction and focusing IR, visible and UV radiation; radiation resistant optical dielectric wave guides; radiation resistant HV insulation for diagnostic magnetic pickup coils; compact radiation and/or magnetic shielding for in-vault diagnostics that need some attenuation to reduce S/N ratio; radiation hardened line-of-sight sensors such as bolometers, UV and soft X-ray detectors, neutral particle analyzers, torus pressure gauges; special maintenance fixtures and tools; material properties - design data base - all materials; and insulation - electrical/thermal and mechanical properties

  18. Liquid nitrogen cooling considerations of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Dabiri, A.E.

    1986-01-01

    An analytical procedure was developed to estimate the cooldown time between pulses of the Compact Ignition Tokamak (CIT) utilizing liquid nitrogen. Fairly good agreement was obtained between the analysis results and those measured in the early fusion experimental devices. The cooldown time between pulses in the CIT is controlled by the energy disposition in the inner leg of the TF coil. A cooldown time of less than one hour is feasible for the CIT if fins are used in the cooling channels. An R and D experimental program is proposed to determine the actual cooldown time between pulses since this would be considered an issue in the conceptual design of the CIT

  19. Equilibrium system analysis in a tokamak ignition experiment

    International Nuclear Information System (INIS)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies

  20. Equilibrium system analysis in a tokamak ignition experiment

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  1. A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

    International Nuclear Information System (INIS)

    Slack, D.S.; Kerns, J.A.

    1987-01-01

    Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

  2. Maintenance features of the Compact Ignition Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Hager, E.R.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs

  3. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  4. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb 3 Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered

  5. Radiation analysis of the CIT [Compact Ignition Tokamak] pellet injector system and its impact on personnel access

    International Nuclear Information System (INIS)

    Selcow, E.C.; Stevens, P.N.; Gomes, I.C.; Gomes, L.M.

    1988-08-01

    The conceptual design of the Compact Ignition Tokamak (CIT) is nearing completion. The CIT is a short-pulse ignition experiment, which is planned to follow the operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). The high neutron wall loadings, 4--5 MW/m 2 , associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of device components as well as personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure, and the entire experiment is housed in a circular test cell facility that has a radius of 12 m. The most critical radiation concerns in the CIT design process relate to the numerous penetrations in the device. This report discusses the impact of a major penetration on the design and operation of the pellet injection system in the CIT. The pellet injector is a major component, and it has a line-of-sight penetration through the igloo and test cell wall. All current options for maintenance of the injector require hands-on-access. A nuclear analysis has been performed to establish the feasibility of hands-on-access. A coupled Monte Carlo/discrete-ordinates methodology was used to perform the analysis. This problem is characterized by deep penetration and streaming with very large length-to-diameter ratios. Results from this study indicate that personnel access to the pellet injector glovebox is possible. 14 refs., 3 figs., 3 tabs

  6. Updated tokamak systems code and applications to high-field ignition devices

    International Nuclear Information System (INIS)

    Reid, R.L.; Galambos, J.D.; Peng, Y-K.M.; Strickler, D.J.; Selcow, E.C.

    1985-01-01

    This paper describes revisions made to the Tokamak Systems Code to more accurately model high-field copper ignition devices. The major areas of revision were in the plasma physics model, the toroidal field (TF) coil model, and the poloidal field (PF) coil/MHD model. Also included in this paper are results obtained from applying the revised code to a study for a high-field copper ignition device to determine the impact of magnetic field on axis, (at the major radius), on performance, and on cost

  7. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  8. Delaying sawtooth oscillations in the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Bateman, G.

    1986-09-01

    A combination of pellets, off-axis heating, and current ramp is used to delay the onset of sawtooth oscillations for 3.4 seconds and achieve ignition with less than 0.2-second confinement time in a 1-1/2-D BALDUR simulation of the Compact Ignition Tokamak. Deuterium and tritium pellets are injected into an initially cold, relatively low density plasma, where they cool the center and produce a very centrally peaked density profile. A centrally peaked density profile (n/sub e0// = 4.0) is subsequently maintained by an inward particle pinch. Twenty megawatts of auxiliary heating is applied halfway between the magnetic axis and the edge of the plasma for 2 seconds after the pellets are injected. The plasma ignites and then burns from the time the auxiliary heating is turned off until the first large sawtooth crash occurs at 3.4 seconds. The burn would be expected to continue after that only if the sawtooth period is sufficiently long (roughly 0.3 seconds or longer)

  9. The CIT [compact ignition tokamak] pellet injection system: Description and supporting research and development

    International Nuclear Information System (INIS)

    Gouge, M.J.; Combs, S.K.; Fisher, P.W.; Milora, S.L.

    1989-01-01

    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

  10. Maintainability features of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Bushnell, C.W.

    1986-01-01

    The Compact Ignition Tokamak (CIT) is a deuterium-tritium (D-T) device envisaged to be the next experimental reactor in the US Fusion Program. The reactor will initially operate in a nonactivated hydrogen phase for approximately two years. This will permit verification of the integrity of the total system and allow hands-on repair to equipment which has experienced shakedown and early operation failures. Once D-T operations commence, reactor maintenance will require remote handling techniques. An evaluation has been completed to determine what maintenance operations must be performed on the CIT. A maintenance philosophy has been developed which is based upon the use of manipulator systems and robotics in the test cell. Replacement of life-limited equipment will be accomplished using a modular design approach for components, with simple remotely operable interfaces. Examples of operations to be done remotely include: (1) replacing of rf antennae and Faraday shields, (2) uncoupling diagnostic and fueling penetrations, (3) removing of all port covers, and (4) replacing first wall armor tiles, optical mirrors, and vacuum windows

  11. Remote maintenance concepts for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Davis, F.C.; Hager, E.R.

    1988-01-01

    Because deuterium-tritium fuel will be used in the Compact Ignition Tokamak (CIT), remote handling technology is needed to carry out some maintenance operations on the machine. In keeping with the compact, low-cost nature of CIT, remote maintenance is provided only for systems with the highest probability of failure. Remote operations include removing, repairing (if feasible), and replacing such components as thermal protection tiles on the first wall, radio-frequency (rf) heating modules, and diagnostic modules. For maintenance inside the vacuum vessel, major pieces of equipment under development include an articulated boom manipulator with servomanipulators, an inspection manipulator, and special tooling. For maintenance outside the cryostat, remote equipment includes a bridge-mounted manipulator system, equipment for decontamination and hot cell activities, and for handling and packaging solid radioactive waste. The conceptual design phase of the CIT project is nearing completion; research and development activities in support of the project include demonstrations of remote maintenance operations on full-size partial mock-ups. 9 figs

  12. Elastic-plastic analysis of the toroidal field coil inner leg of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Horie, T.

    1987-07-01

    Elastic-plastic analyses were made for the inner leg of the Compact Ignition Tokamak toroidal field (TF) coil, which is made of copper-Inconel composite material. From the result of the elastic-plastic analysis, the effective Young's moduli of the inner leg were determined by the analytical equations. These Young's moduli are useful for the three-dimensional, elastic, overall TF coil analysis. Comparison among the results of the baseline design (R = 1.324 m), the bucked pressless design, the 1.527-m major radius design, and the 1.6-m major radius design was also made, based on the elastic-plastic TF coil inner leg analyses

  13. Plans for the CIT [Compact Ignition Tokamak] instrumentation and control system

    International Nuclear Information System (INIS)

    Preckshot, G.G.

    1987-01-01

    Extensive experience with previous fusion experiments (TFTR, MFTF-B and others) is driving the design of the Instrumentation and Control System (I and C) for the Compact Ignition Tokamak (CIT) to be built at Princeton. The new design will reuse much equipment from TFTR and will be subdivided into six major parts: machine control, machine data acquisition, plasma diagnostic instrument control and instrument data acquisition, the database, shot sequencing and safety interlocks. In a major departure from previous fusion experiment control systems, the CIT machine control system will be a commercial process control system. Since the machine control system will be purchased as a completely functional product, we will be able to concentrate development manpower in plasma diagnostic instrument control, data acquisition, data processing and analysis, and database systems. We will discuss the issues driving the design, give a design overview and state the requirements upon any prospective commercial process control system

  14. Microinstability-based models for confinement properties and ignition criteria in tokamaks

    International Nuclear Information System (INIS)

    Tang, W.M.; Bishop, C.M.; Coppi, B.; Kaye, S.M.; Perkins, F.W.; Redi, M.H.; Rewoldt, G.

    1987-02-01

    This paper reports on results of theoretical studies dealing with: (1) the use of microinstability-based thermal transport models to interpret the anomalous confinement properties observed in key tokamak experiments such as TFTR and (2) the likely consequences of the presence of such instabilities for future ignition devices. Transport code simulations using profile-consistent forms of anomalous thermal diffusivities due to drift-type instabilities have yielded good agreement with the confinement times and temperatures observed in TFTR under a large variety of operating conditions including pellet-fuelling in both ohmic- and neutral-beam-heated discharges. With regard to achieving an optimal ignition margin, the adverse temperature scaling of anomalous losses caused by drift modes leads to the conclusion that it is best to operate at the maximum allowable density while holding the temperature close to the minimum value required for ignition

  15. Equilibrium system analysis in a tokamak ignition experiment. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  16. Potential off-normal events and associated radiological source terms for the compact ignition tokamak: Fusion Safety Program

    International Nuclear Information System (INIS)

    Holland, D.F.; Lyon, R.E.

    1987-10-01

    The Compact Ignition Tokamak (CIT), the latest step in the United States program to develop the commercial application of fusion power, is designed as the first fusion device to achieve ignition conditions. It is to be constructed near Princeton, New Jersey on the site of the existing Tokamak Fusion Test Reactor (TFTR). To address the environmental impact and public safety concerns, a preliminary analysis was performed of potential off-normal radiological releases. Operational occurrences, natural phenomena, accidents with external origins, and accidents external to the PPPL site were considered as potential sources for off-normal events. Based on an initial screening, events were selected for preliminary analysis. Included in these events were tritium releases from the tritium delivery and recovery system, tritium releases from the torus, releases of activated nitrogen from the test cell or cryostat, seismic events, and shipping accidents. In each case, the design considerations related to the event were reviewed and the release scenarios discussed. Because of the complexity of some of the proposed safety systems, in some cases event trees were used to describe the accident scenarios. For each scenario, the probability was estimated as well as the release magnitude, isotope, chemical form, and release mode. 10 refs., 17 figs., 5 tabs

  17. Ex-vessel remote maintenance for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist in removing and repairing such components as diagnostic modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the plasma chamber includes a bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid-radiation-waste-handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design were addressed in fiscal year (FY) 1987, including conceptual designs of manipulator systems, the start of a remote equipment research and development (RandD) program, and definition of the hot cell, decon, and equipment repair facility requirements. R and D work included preliminary 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 (ROMD) Facility. 1 ref., 6 figs

  18. Accessibility and replacement as prime constraints in the design of large experimental tokamaks

    International Nuclear Information System (INIS)

    Challender, R.S.; Reynolds, P.

    1976-01-01

    An attempt is made to bring together those design features of large, experimental Tokamaks, which would lead to better accessibility during non-active operation and, in particular, permit replacement and repair after activation, thereby making possible an extended period of experimental operation into the ignition phase

  19. Instrumentation and controls of an ignited tokamak

    International Nuclear Information System (INIS)

    Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega, R.J.; Raju, G.V.S.; Stone, R.S.

    1980-10-01

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented

  20. Neutral-beam requirements for compression-boosted ignited tokamak plasmas

    International Nuclear Information System (INIS)

    Cohn, D.R.; Jassby, D.L.; Kreischer, K.

    1977-12-01

    Neutral-beam energies of 200 to 500-keV D 0 may be required to insure adequate penetration into the center of ignition-sized tokamak plasmas. However, the beam energy requirement can be reduced by using a start-up scenario in which the final plasma is formed by major-radius compression of a beam-heated plasma whose density-radius product, na, is determined by satisfactory neutral-beam penetration. ''Compression boosting'' is attractive only for plasmas in which ntau/sub E/ increases with na, because a major-radius compression C increases na by C 3 / 2 . The dependence on C of beam energy and beam power for plasmas which obey ''empirical scaling laws'' of the type ntau/sub E/ varies as (na) 2 is analyzed. The dependences on C of stored magnetic energy and TF-coil power dissipation are also determined. It is found that a compression ratio of 1.5 to attain the ignited plasma permits adequate penetration by 150-keV D 0 beams

  1. A minimum-size tokamak concept for conditions near ignition

    International Nuclear Information System (INIS)

    Lehnert, B.

    1983-01-01

    Based on a combination of Alcator scaling and a recent theory on the Murakami density limit, a minimum-size tokamak concept (Minitor) is proposed. Even if this concept does not aim at alpha particle containment, it has the important goal of reaching plasma core temperatures and Lawson parameter values required for ignition, by ohmic heating alone and under macroscopically stable conditions. The minimized size, and the associated enhancement of the plasma current density, are found to favour high plasma temperatues, average densities, and beta values. The goal of this concept appears to be realizable by relatively modest technical means. (author)

  2. Ignition and burn control characteristics of thermonuclear plasmas

    International Nuclear Information System (INIS)

    Chaniotakis, E.A.

    1990-01-01

    Achieving the long sought goal of fusion energy requires the attainment of an ignited and controlled thermonuclear plasma. Obtaining an ignited plasma in a tokamak device requires consideration of both the physics of the plasma and the engineering of the machine. With the aide of completely analytical procedure optimized and ignited tokamaks are obtained under various physics assumptions. These designs show the possible advantage of tokamaks characterized by high (∼4.5) aspect ratio, and high (∼15 T) toroidal magnetic field. The control of an ignited plasma is investigated by using auxiliary power modulation. With auxiliary power stable operating points can be created with Q ∼50. Recognizing the need for a fast 1 1/2-D transport model for studying profile effects the plasma transport equations are solved using variational methods. A computer model based on the variational method has been developed. This model solves the 1 1/2-D transport equation very fast with little loss of accuracy. 74 refs., 70 figs., 8 tabs

  3. Ignition and time-dependent fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-01-01

    The eventual utilization of a tokamak fusion reactor for commercial power necessitates a thorough understanding of the operational requirements at full and fractional power levels and during transitions from one operating level to another. In this study we examine the role of burn control in maintaining the reactor plasma at equilibrium to avoid thermal runaway during fractional power operation. Because these requirements rely so heavily on the assumptions that govern the plasma transport, this study focuses on time-dependent analyses and a comparison of ignition requirements using a range of energy confinement

  4. Remote maintenance of Compact Ignition Tokamak ex-vessel systems

    International Nuclear Information System (INIS)

    DePew, R.E.; Macdonald, D.

    1989-01-01

    The use of deuterium-tritium (D-T) fuel in the Compact Ignition Tokamak (CIT) will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion device's auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. Throughout the CIT remote maintenance (RM) studies conducted to date, computer modeling has been used extensively to investigate manipulator access in these complex, tightly packed, and cluttered surroundings. A recent refinement of computer modeling involves the use of an intelligent engineering work station for realtime interactive display of task simulations. This paper discusses the use of three-dimensional (3-D) kinematic computer models of the CIT machines that are proving to be powerful tools in our efforts to evaluate RM requirements. This presentation includes a video-taped simulation of remote replacement of a plasma viewing assembly. The simulation illustrates some of the constraints associated with typical RM activities and the ways in which computer modeling enhances the design process. 1 ref., 3 figs

  5. Ignition in the next step tokamak

    International Nuclear Information System (INIS)

    Johner, J.

    1990-07-01

    A 1/2-D model for thermal equilibrium of a thermonuclear plasma with transport described by an empirical global energy confinement time is described. Ignition in NET and ITER is studied for a number of energy confinement time scaling expressions. Ignited operation of these machines at the design value of the neutron wall load is shown to satisfy both beta and density constraints. The value of the confinement time enhancement factor required for such operation is found to be lower for the more recently proposed scaling expressions than it is for the oldest ones. With such new scalings, ignition could be obtained in H-mode in NET and ITER even with relatively flat density profiles

  6. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  7. First wall thermomechanical stress analysis in a fusion ignition experiment

    International Nuclear Information System (INIS)

    Rodin, G.; Carrera, R.; Howell, J.; Hwang, Y.L.; Montalvo, E.; Ordonez, C.; Dong, J.Q.

    1990-01-01

    The fusion ignition experiment IGNITEX + has been proposed as a low cost means of producing and controlling fusion ignited plasmas for scientific study. A single-turn-coil tokamak plasmas for scientific study. A single-turn-coil tokamak cryogenically precooled at liquid nitrogen temperature is used to produce 20 T fields and 12 MA plasma currents so that high-density ohmic ignition is possible. The high-field, high-density operation should maintain the plasma relatively free of wall impurities. In order to minimize plasma cooling, a low-Z first wall is considered for IGNITEX. The IGNITEX design philosophy emphasizes simplicity and low cost. A limiterless, smooth first will without files and plates is proposed. A low-Z material is applied by plasma jet techniques over a resistive vacuum vessel. This design is thought to be adequate for a magnetic fusion ignition experiment. Maintenance and operation of the first wall system is significantly simplified when compared to conventional designs

  8. Accessibility and replacement as prime constraints in the design of large experimental Tokamaks

    International Nuclear Information System (INIS)

    Challender, R.S.; Reynolds, P.

    1976-01-01

    Many of the designs being developed for large Tokamak experiments are based on the classical geometry of the small machines in laboratories throughout the world: a circular array of coils interlinked by an inner toroidal vessel. An attempt is made to bring together those design features which would lead to better accessibility during non-active operation and in particular permit replacement and repair after activation, thereby making possible an extended period of experimental operation into the ignition phase

  9. Reversed field pinch ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    Plasma models are described and used to calculated numerically the transport confinement (nτ E ) requirements and steady state operation points for both the reversed field pinch (RFP) and the tokamak. The models are used to examine the CIT tokamak ignition conditions and the RFP experimental and ignition conditions. Physics differences between RFPs and tokamaks and their consequences for a D-T ignition machine are discussed. Compared with a tokamak, the ignition RFP has many physics advantages, including Ohmic heating to ignition (no need for auxiliary heating systems), higher beta, lower ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits) and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic field, larger aspect ratios and smaller plasma cross-sections, translate to significant cost reductions for both ignition and reactor applications. The primary drawback of the RFP is the uncertainty that the present scaling will extrapolate to reactor regimes. Devices that are under construction should go a long way toward resolving this scaling uncertainty. The 4 MA ZTH is expected to extend the nτ E transport scaling data by three orders of magnitude above the results of ZT-40M, and, if the present scaling holds, ZTH is expected to achieve a D-T equivalent scientific energy breakeven, Q = 1. A base case RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. (author). 19 refs, 11 figs, 3 tabs

  10. Topics in stability and transport in tokamaks: Dynamic transition to second stability with auxiliary heating; stability of global Alfven waves in an ignited plasma

    International Nuclear Information System (INIS)

    Fu, G.

    1988-01-01

    The problem of access to the high-beta ballooning second-stability regime by means of auxiliary heating and the problem of the stability of global-shear Alfven waves in an ignited tokamak plasma are theoretically investigated. These two problems are related to the confinement of both the bulk plasma as well as the fusion-product alpha particles an dare fundamentally important to the operation of ignited tokamak plasma. First, a model that incorporates both transport and ballooning mode stability was developed in order to estimate the auxiliary heating power required for tokamak plasma to evolve in time self-consistently into a high-beta, globally self-stabilized equilibrium. The critical heating power needed for access to second stability is found to be proportional to the square root of the anomalous diffusivity induced by the ballooning instability. Next, the full effects of toroidicity are retained in a theoretical description of global-type-shear Alfven modes whose stability can be modified by the fusion-product alpha particles that will present in an ignited tokamak plasma. Toroidicity is found to induce mode coupling and to stabilize the so-called Global Alfven Eigenmodes (GAE)

  11. Neutral pumping rates for a next step tokamak ignition device

    International Nuclear Information System (INIS)

    Galambos, J.D.; Peng, Y.K.M.; Heifetz, D.

    1985-01-01

    Neutral pumping rates are calculated for pump-limiter and divertor options of a next step tokamak ignition device using a method that accounts for the coupled effects of neutral transport and plasma transport. For both pump limiters and divertors the plasma flow into the channel surrounding the neutralizer plate is greatly reduced by the neutral recycling. The fraction of this flow that is pumped can be large (>50%) but in general is dependent on the particular geometry and plasma conditions. It is estimated that pumping speeds greater than or approximately 10 5 L/s are adequate for the exhaust requirements in the pump-limiter and the divertor cases

  12. RF-driven tokamak reactor with sub-ignited, thermally stable operation

    International Nuclear Information System (INIS)

    Harten, L.P.; Bers, A.; Fuchs, V.; Shoucri, M.M.

    1981-02-01

    A Radio-Frequency Driven Tokamak Reactor (RFDTR) can use RF-power, programmed by a delayed temperature measurement, to thermally stabilize a power equilibrium below ignition, and to drive a steady state current. We propose the parameters for such a device generating approx. = 1600 MW thermal power and operating with Q approx. = 40 (= power out/power in). A one temperature zero-dimensional model allows simple analytical formulation of the problem. The relevance of injected impurities for locating the equilibrium is discussed. We present the results of a one-dimensional (radial) code which includes the deposition of the supplementary power, and compare with our zero-dimensional model

  13. Radial effects in heating and thermal stability of a sub-ignited tokamak

    International Nuclear Information System (INIS)

    Fuchs, V.; Shoucri, M.M.; Thibaudeau, G.; Harten, L.; Bers, A.

    1982-02-01

    The existence of thermally stable sub-ignited equilibria of a tokamak reactor, sustained in operation by a feedback-controlled supplementary heating source, is demonstrated. The establishment of stability depends on a number of radially non-uniform, nonlinear processes whose effect is analyzed. One-dimensional (radial) stability analyses of model transport equations, together with numerical results from a 1-D transport code, are used in studying the heating of DT-plasmas in the thermonuclear regime. Plasma core supplementary heating is found to be a thermally more stable process than bulk heating. In the presence of impurity line radiation, however, core-heated temperature profiles may collapse, contracting inward from the limiter, the result of an instability caused by the increasing nature of the radiative cooling rate, with decreasing temperature. Conditions are established for the realization of a sub-ignited high-Q, toroidal reactor plasma with appreciable output power

  14. Liquid nitrogen cooling for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Fleming, R.B.; Martin, G.D.; Lyon, R.E.

    1989-01-01

    The Compact Ignition Tokamak (CIT), which is currently being designed, will have toroidal and poloidal magnetic field coils pre-cooled by liquid nitrogen to a temperature near 80 degree K prior to each plasma pulse. The purpose is to gain the advantage of lower copper resistivity at reduced temperature. To maintain this temperature, the field coils, vacuum vessel, and surrounding structure will be enclosed within a cryostat. During a full-power D-T pulse, nuclear and resistive heating will impart a heat load of 11.0 GJ to the coils, which will raise the temperature of certain areas of the coils to near room temperature. The cryogenic system will supply 60,000 kg (19,500 gallons) of liquid nitrogen to remove this heat within a 60-minute cool-down period between pulses. A primary design consideration is that the nitrogen gas within the cryostat during a pulse will be activated by neutrons, producing nitrogen-13, which has a half-life of 10 minutes. This gas cannot be released into the environment without a sufficient decay period. The coolant nitrogen will therefore be contained within a closed (primary) circuit, and will be condensed in a heat exchanger. Liquid nitrogen from the supply dewars will be evaporated on the other side of the exchanger (the secondary side), and released to the atmosphere via a roof vent. Other operating modes (standby operation and initial cool-down from room temperature) are described in the paper. A safety analysis indicates that the cryogenic system will meet all applicable environmental requirements. 1 ref., 1 fig., 1 tab

  15. A polar-drive shock-ignition design for the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, K. S.; McKenty, P. W.; Collins, T. J. B.; Craxton, R. S.; Delettrez, J. A.; Marozas, J. A.; Skupsky, S.; Shvydky, A. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Betti, R. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Fusion Science Center, University of Rochester, Rochester, New York 14623 (United States); Departments of Mechanical Engineering and Physics, University of Rochester, Rochester, New York 14627 (United States); Hohenberger, M.; Theobald, W.; Lafon, M.; Nora, R. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Fusion Science Center, University of Rochester, Rochester, New York 14623 (United States)

    2013-05-15

    Shock ignition [R. Betti et al., Phys. Rev. Lett. 98, 155001 (2007)] is being pursued as a viable option to achieve ignition on the National Ignition Facility (NIF). Shock-ignition target designs use a high-intensity laser spike at the end of a low-adiabat assembly pulse to launch a spherically convergent strong shock to ignite the hot spot of an imploding capsule. A shock-ignition target design for the NIF is presented. One-dimensional simulations indicate an ignition threshold factor of 4.1 with a gain of 58. A polar-drive beam-pointing configuration for shock-ignition experiments on the NIF at 750 kJ is proposed. The capsule design is shown to be robust to the various one- and two-dimensional effects and nonuniformities anticipated on the NIF. The target is predicted to ignite with a gain of 38 when including all anticipated levels of nonuniformity and system uncertainty.

  16. Electromagnetic loads and structural response of the CIT [Compact Ignition Tokamak] vacuum vessel to plasma disruptions

    International Nuclear Information System (INIS)

    Salem, S.L.; Listvinsky, G.; Lee, M.Y.; Bailey, C.

    1987-01-01

    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

  17. Ignition experiment - alternatives

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1979-10-01

    This report comprises three short papers on cost estimates, integral burn time and alternative versions of Tokamak ignition experiments. These papers were discussed at the ZEPHYR workshop with participants from IPP Garching, MIT Cambridge and PPPL Princeton (Garching July 30 - August 2 1979) (Chapters A, B, C). It is shown, that starting from a practical parameter independent minimum integral burn time of Tokamak ignition experiments (some 10 3 s) by adding a shield for protection of the magnet insulation (permitted neutron dose 10 9 rad) an integral burn time of some 10 4 s can be achieved for only about 30% more outlay. For a substantially longer integral burn time the outlay approaches rather quickly that for a Tokamak reactor. Some examples for alternatives to ZEPHYR are being given, including some with low or no compression. In a further chapter D some early results of evaluating an ignition experiment on the basis of the energy confinement scaling put forward by Coppi and Mazzucato are presented. As opposed to the case of the Alcator scaling used in chapters A through C the minimum integral burn time of Tokamak ignition experiments here depends on the plasma current. Provided neutral injectors up to about 160 keV are available compression boosting is not required with this scaling. The results presented have been obtained neglecting the effects of the toroidal field ripple. (orig.) 891 HT/orig. 892 RKD [de

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

    International Nuclear Information System (INIS)

    Lee, V.D.

    1987-01-01

    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. This paper presents the configuration development of the hydraulic press used to vertically preload the CIT device

  19. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  20. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  1. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  2. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

  3. Tokamak Physics Experiment (TPX) design

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1995-01-01

    TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

  4. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  5. Tokamak building-design considerations for a large tokamak device

    International Nuclear Information System (INIS)

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

  6. Stability of the Global Alfven Eigenmode in the presence of fusion alpha particles in an ignited tokamak plasma

    International Nuclear Information System (INIS)

    Fu, G.Y.; Van Dam, J.W.

    1989-05-01

    The stability of the Global Alfven Eigenmodes is investigated in the presence of super-Alfvenic energetic particles, such as the fusion-product alpha particles in an ignited deuterium-tritium tokamak plasma. Alpha particles tend to destabilize these modes when ω *α > ω A , where ω A is the shear-Alfven modal frequency and ω *α is the alpha particle diamagnetic drift frequency. This destabilization due to alpha particles is found to be significantly enhanced when the alpha particles are modeled with a slowing-down distribution function rather than with a Maxwellian. However, previously neglected electron damping due to the magnetic curvature drift is found to be comparable in magnitude to the destabilizing alpha particle term. Furthermore, the effects of toroidicity are also found to be stabilizing, since the intrinsic toroidicity induces poloidal mode coupling, which enhances the parallel electron damping from the sideband shear-Alfven Landau resonance. In particular, for the parameters of the proposed Compact Ignition Tokamak, the Global Alfven Eigenmodes are found to be completely stabilized by either the electron damping that enters through the magnetic curvature drift or the damping introduced by finite toroidicity. 29 refs., 8 figs., 1 tab

  7. Approach to ignition of tokamak reactors

    International Nuclear Information System (INIS)

    Sigmar, D.J.

    1981-02-01

    Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup

  8. Overview of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1986-01-01

    The mission of CIT is to realize, study, and optimize fully ignited plasma discharges. The physics requirements have been established to provide reasonable assurance that the mission will be achieved: (1) plasma confinement guidelines consider all present scaling laws; ohmic, auxiliary heated L-mode, auxiliary heated H-mode, (2) figure-of-merit established: X = aB/q; this is proportional to ignition margin; X must be >25, (3) burn pulse duration set at ten times tau-E; an additional two times tau-E specified to heat to ignition, and (4) capability to operate both in limiter and divertor mode

  9. Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

    International Nuclear Information System (INIS)

    Gillen, T.J.

    1980-11-01

    A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

  10. Progress in ICRF heating technology and designs for future large tokamak heating systems

    International Nuclear Information System (INIS)

    Baity, F.W.; Swain, D.W.; Hoffman, D.J.; Becraft, W.R.; Bryan, W.E.; Mayberry, M.J.; Owens, T.L.; Yugo, J.J.

    1986-01-01

    The problem of advancing the technology of heating with the ion cyclotron range of frequencies (ICRF) for successful application to ignited plasmas is being addressed at Oak Ridge National Laboratory (ORNL) with the collaboration of several laboratories in the United States and Europe. The needs of experiments such as the Compact Ignition Tokamak (CIT) have been evaluated and conceptual approaches identified. These concepts and their components are examined in the laboratory and applied to present-day machines. The status of this program is presented

  11. Design of ignition targets for the National Ignition Facility

    International Nuclear Information System (INIS)

    Haan, S.W.; Dittrich, T.R.; Marinak, M.M.; Hinkel, D.E.

    1999-01-01

    This is a brief update on the work being done to design ignition targets for the National Ignition Facility. Updates are presented on three areas of current activity : improvements in modeling, work on a variety of targets spanning the parameter space of possible ignition targets ; and the setting of specifications for target fabrication and diagnostics. Highlights of recent activity include : a simulation of the Rayleigh-Taylor instability growth on an imploding capsule, done in 3D on a 72degree by 72degree wedge, with enough zones to resolve modes out to 100 ; and designs of targets at 250eV and 350eV, as well as the baseline 300 eV ; and variation of the central DT gas density, which influences both the Rayleigh-Taylor growth and the smoothness of the DT ice layer

  12. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1983-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

  13. Mission and design of the Fusion Ignition Research Experiment (FIRE)

    International Nuclear Information System (INIS)

    Meade, D.M.; Jardin, S.C.; Schmidt, J.

    2001-01-01

    Experiments are needed to test and extend present understanding of confinement, macroscopic stability, alpha-driven instabilities, and particle/power exhaust in plasmas dominated by alpha heating. A key issue is to what extent pressure profile evolution driven by strong alpha heating will act to self-organize advanced configurations with large bootstrap current fractions and internal transport barriers. A design study of a Fusion Ignition Research Experiment (FIRE) is underway to assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis is on understanding the behavior of fusion plasmas dominated by alpha heating (Q≥5) that are sustained for durations comparable to the characteristic plasma time scales (≥20 τ E and ∼τ skin , where τ skin is the time for the plasma current profile to redistribute at fixed current). The programmatic mission of FIRE is to attain, explore, understand and optimize alpha-dominated plasmas to provide knowledge for the design of attractive magnetic fusion energy systems. The programmatic strategy is to access the alpha-heating-dominated regime with confidence using the present advanced tokamak data base (e.g., Elmy-H-mode, ≤0.75 Greenwald density) while maintaining the flexibility for accessing and exploring other advanced tokamak modes (e. g., reversed shear, pellet enhanced performance) at lower magnetic fields and fusion power for longer durations in later stages of the experimental program. A major goal is to develop a design concept that could meet these physics objectives with a construction cost in the range of $1B. (author)

  14. Remote replacement of TF [toroidal field] and PF [poloidal field] coils for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Macdonald, D.; Watkin, D.C.; Hollis, M.J.; DePew, R.E.; Kuban, D.P.

    1990-01-01

    The use of deuterium-tritium fuel in the Compact Ignition Tokamak will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion devices auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. In-vessel remote maintenance included replacement of divertor and first wall hardware, faraday shields, and for an in-vessel inspection system. Provision for remote replacement of a vacuum vessel sector, toroidal field coil or poloidal field ring coil was not included in the project baseline. As a result of recent coil failures experienced at a number of facilities, the CIT project decided to reconsider the question of remote recovery from a coil failure and, in January of 1990, initiated a coil replacement study. This study focused on the technical requirements and impact on fusion machine design associated with remote recovery from any coil failure

  15. Thermal and mechanical analysis of the Faraday shield for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Vesey, R.A.

    1988-02-01

    The antenna for the ion cyclotron resonance heating (ICRH) system of the Compact Ignition Tokamak (CIT) is protected from the plasma environment by a Faraday shield, an array of gas-cooled metallic tubes. The plasma side of the tubes is armored with graphite tiles, which can be either brazed or mechanically attached to the tube. The Faraday shield has been analyzed using finite element codes to model thermal and mechanical responses to typical CIT heating and disruption loads. Four representative materials (Inconel 718, tantalum-10 tungsten, copper alloy C17510, and molybdenum alloy TZM) and several combinations of tube and armor thicknesses were used in the thermal analysis, which revealed that maximum allowable temperatures were not exceeded for any of the four materials considered. The two-dimensional thermal stress analysis indicated Von Mises stresses greater than twice the yield stress for a tube constructed of Inconel 718 (the original design material) for the brazed-graphite design. Analysis of stresses caused by plasma disruption (/rvec J/ /times/ /rvec B/) loads eliminated the copper and molybdenum alloys as candidate tube materials. Of the four materials considered, tantalum-10 tungsten performed the best for a brazed graphite design, showing acceptable thermal stresses (69% of yield) and disruption stresses (42% of yield). A preliminary thermal analysis of the mechanically attached graphite scheme predicts minimal thermal stresses in the tube. The survivability of the graphite tubes in this scheme is yet to be analyzed. 8 refs., 19 figs., 2 tabs

  16. Mathematical modeling plasma transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1997-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 1020/m3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

  17. Mathematical modeling plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Quiang, Ji

    1995-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

  18. Effects of enhanced elongation and paramagnetism on the parameter space of the ignition spherical torus

    International Nuclear Information System (INIS)

    Strickler, D.J.; Peng, Y-K.M.; Borowski, S.K.; Selcow, E.C.; Miller, J.B.

    1985-01-01

    The Ignition Spherical Torus (IST) is a small aspect ratio device retaining only indispensable components along the major axis of a tokamak plasma, such as a cooled, normal conductor producing a toroidal magnetic field. The IST is expected to be a cost-effective approach to ignition by taking advantage of low field, large natural plasma elongation, high plasma current, high beta, and tokamak confinement. These result in compact, high-performance devices with relatively simple magnetic systems as compared with ignition tokamaks of larger aspect ratio. The plasma enhancement of the toroidal field on axis, or plasma paramagnetism, is significant in the IST. The use of this plasma-enhanced field in conventional tokamak beta and density limits leads to increased plasma pressure and performance and therefore smaller device size for a given ignition margin

  19. Energy transport requirements for tokamak reactors in the second ballooning stability regime

    International Nuclear Information System (INIS)

    Potok, R.E.; Bromberg, L.; Cohn, D.R.

    1986-01-01

    The authors present an analysis of ignition confinement constraints on a tokamak reactor operating in the second regime of ballooning stability. This regime is characterized by flat plasma pressure profiles, with a sharp pressure gradient near a conducting first wall at the plasma edge. The energy confinement time is determined by transport processes across the pressure gradient region. The authors have found that the required transport needed for ignition in the edge region is very close to the value predicted by neoclassical ion conductivity scaling. Only by carefully tailoring the conductivity scaling across the flux coordinate were the authors able to match both the kink stability and ignition requirements. With optimistic assumptions, R/sub o/ ≅ 7 m appears to be the minimum major radius for an economical tokamak reactor in the second ballooning stability regime. This paper presents a base design case at R/sub o/ = 7 m, and shows how the reactor design varies with changes in major radius, ion transport scaling, and electron transport scaling

  20. Simulations of beam-fueled supershot-like plasmas near ignition

    International Nuclear Information System (INIS)

    Budny, R.V.; Grisham, L.; Jassby, D.L.

    1992-01-01

    In certain conditions, neutral beam injection (NBI) and low recycling result in supershot plasmas. These are characterized by peaked density profiles and high central ion temperatures. We discuss the potential advantages of NBI fueled supershot-like plasmas in tokamaks operating near ignition. The goal is to investigate the feasibility of these plasmas to aid in the design of future advanced tokamaks. NBI has been very successful in advancing tokamak plasmas close to ignition conditions. The primary benefits of NBI are heating and particle fueling, but the plasma currents generated by the beam ions are also of considerable interest. The optimal injection energy E inj for the beam ions depends on the desired role of the NBI. For central particle fueling, E inj should be low to maximize the particle current at fixed P B , but high enough to penetrate to the center. For heating and current drive, higher E inj is preferable for deepest penetration. With the standard positive ion beam technology, the neutralization efficiency becomes too low for useful power densities if E inj is significantly greater than about 120 keV. Negative ion beam sources would be useful for heating and current drive at very high E inj (500 keV or more), but the fueling rate of NBI is too low to be practical. It seems generally accepted that future tokamaks which operate closer to ignition will have to be fueled and heated by means other than NBI since it is argued that the beams with low E inj cannot penetrate deeply into the dense plasmas of interest. (author) 3 refs., 4 figs

  1. Neutral beam system for an ignition tokamak

    International Nuclear Information System (INIS)

    Fasolo, J.; Fuja, R.; Jung, J.; Moenich, J.; Norem, J.; Praeg, W.; Stevens, H.

    1978-01-01

    We have attempted to make detailed designs of several neutral beam systems which would be applicable to a large machine, e.g. an ITR (Ignition Test Reactor), EPR (Experimental Power Reactor), or reactor. Detailed studies of beam transport to the reactor and neutron transport from the reactor have been made. We have also considered constraints imposed by the neutron radiation environment in the injectors, and the resulting shielding, radiation-damage, and maintenance problems. The effects of neutron heat loads on cryopanels and ZrAl getter panels have been considered. Design studies of power supplies, vacuum systems, bending magnets, and injector layouts are in progress and will be discussed

  2. The use of scaling laws for the design of high beta tokamaks

    International Nuclear Information System (INIS)

    Mauel, M.E.

    1987-01-01

    Several different empirical scaling laws for the tokamak energy confinement time are used to estimate the auxiliary heating power required for a laboratory experiment capable of testing tokamak confinement at high beta and techniques to access the second stability regime. Since operating experience in the second stability regime does not yet exist, these laws predict a wide range of possible power requirements, especially at large aspect ratios. However, by examining a model DT fusion power reactor with reasonable restrictions on the fusion island weight, neutron loading, and maximum magnetic field of the external coils, only a limited range of operating conditions are found for both first and second regime tokamaks, and only a subset of the scaling laws predict ignition. These particular scaling laws are then used to set confinement goals which if demonstrated by the laboratory experiment would indicate favourable scaling to a reactor. (author)

  3. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  4. Prediction of density limits in tokamaks: Theory, comparison with experiment, and application to the proposed Fusion Ignition Research Experiment

    International Nuclear Information System (INIS)

    Stacey, Weston M.

    2002-01-01

    A framework for the predictive calculation of density limits in future tokamaks is proposed. Theoretical models for different density limit phenomena are summarized, and the requirements for additional models are identified. These theoretical density limit models have been incorporated into a relatively simple, but phenomenologically comprehensive, integrated numerical calculation of the core, edge, and divertor plasmas and of the recycling neutrals, in order to obtain plasma parameters needed for the evaluation of the theoretical models. A comparison of these theoretical predictions with observed density limits in current experiments is summarized. A model for the calculation of edge pedestal parameters, which is needed in order to apply the density limit predictions to future tokamaks, is summarized. An application to predict the proximity to density limits and the edge pedestal parameters of the proposed Fusion Ignition Research Experiment is described

  5. Ignition of an overheated, underdense, fusioning tokamak plasma

    International Nuclear Information System (INIS)

    Singer, C.E.; Jassby, D.L.; Hovey, J.

    1979-08-01

    Methods of igniting an overheated but underdense D-T plasma core with a cold plasma blanket are investigated using a simple two-zone model with a variety of transport scaling laws, and also using a one-dimensional transport code. The power consumption of neutral-beam injectors required to produce ignition can be reduced significantly if the underdense core plasma is heated to temperatures much higher than the final equilibrium ignition values, followed by fueling from a cold plasma blanket. It is also found that the allowed impurity concentration in the initial hot core can be greater than normally permitted for ignition provided that the blanket is free from impurities

  6. Mechanical configuration for a superconducting ignition tokamak (TIBER)

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Johnston, B.M.

    1985-10-01

    The Lawrence Livermore National Laboratory is evaluating the engineering feasibility and economics of a superconducting ignition tokamak. Two major operational requirements had to be satisfied: (1) the conductive heat leak to the refrigerated structure had to be minimized, and (2) assembly and maintenance of the entire experiment had to be possible with remotely operated tools. The middle poloidal ''push coil'' must have many annular disks to transfer the TF-coil inward force to the post without crushing superconductor. The toppling moment on the TF-coil vertical legs is huge. A method of keying together the TF-coil cases has been developed. This forms an integrated structure that resists torque. The joining technique permits linear motion for simple assembly/disassembly. the topping moment on the outer vertical legs of the TF coils is very large. To react that moment and avoid great coil-case bulk, we have developed a method that allows the PF coil support structure to assist the TF case structure. Finite element techniques were used to determine the ability of the coal case and conductor to react the magnetic loads. The entire cold-coil structure is mounted on a circular plate that is suspended by several large tension rods, similar to the MFTF-B yin-yang support rods. The vacuum vessel is all at room temperature and is configured like a bell jar with sixteen side doors, one for each shield module

  7. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

  8. Design aspects of low activation fusion ignition experiments

    International Nuclear Information System (INIS)

    Cheng, E.T.; Creedon, R.L.; Hopkins, G.R.; Trester, P.W.; Wong, C.P.C.; Schultz, K.R.

    1986-01-01

    Preliminary design studies have been done exploring (1) materials selection, (2) shutdown biological dose rates, (3) mechanical design and (4) thermal design of a fusion ignition experiment made of low activation materials. From the results of these preliminary design studies it appears that an ignition experiment could be built of low activation materials, and that this design would allow hands-on access for maintenance

  9. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

    Purcell, J.R.; Chen, W.; Thomas, R.

    1976-01-01

    Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

  10. Conceptual design of a commercial tokamak reactor using resistive magnets

    International Nuclear Information System (INIS)

    LeClaire, R.J. Jr.

    1988-01-01

    The future of the tokamak approach to controlled thermonuclear fusion depends in part on its potential as a commercial electricity-producing device. This potential is continually being evaluated in the fusion community using parametric, system, and conceptual studies of various approaches to improving tokamak reactor design. The potential of tokamaks using resistive magnets as commercial electricity-producing reactors is explored. Parametric studies have been performed to examine the major trade-offs of the system and to identify the most promising configurations for a tokamak using resistive magnets. In addition, a number of engineering issues have been examined including magnet design, blanket/first-wall design, and maintenance. The study indicates that attractive design space does exist and presents a conceptual design for the Resistive Magnet Commercial Tokamak Reactor (RCTR). No issue has been identified, including recirculating power, that would make the overall cost of electricity of RCTR significantly different from that of a comparably sized superconducting tokamak. However, RCTR may have reliability and maintenance advantages over commercial superconducting magnet devices

  11. Direct-drive–ignition designs with mid-Z ablators

    Energy Technology Data Exchange (ETDEWEB)

    Lafon, M.; Betti, R. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Laboratory for Laser Energetics and Fusion Science Center, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Anderson, K. S.; Collins, T. J. B.; Epstein, R.; McKenty, P. W.; Myatt, J. F.; Shvydky, A.; Skupsky, S. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States)

    2015-03-15

    Achieving thermonuclear ignition using direct laser illumination relies on the capability to accelerate spherical shells to high implosion velocities while maintaining shell integrity. Ablator materials of moderate atomic number Z reduce the detrimental effects of laser–plasma instabilities in direct-drive implosions. To validate the physics of moderate-Z ablator materials for ignition target designs on the National Ignition Facility (NIF), hydro-equivalent targets are designed using pure plastic (CH), high-density carbon, and glass (SiO{sub 2}) ablators. The hydrodynamic stability of these targets is investigated through two-dimensional (2D) single-mode and multimode simulations. The overall stability of these targets to laser-imprint perturbations and low-mode asymmetries makes it possible to design high-gain targets. Designs using polar-drive illumination are developed within the NIF laser system specifications. Mid-Z ablator targets are an attractive candidate for direct-drive ignition since they present better overall performance than plastic ablator targets through reduced laser–plasma instabilities and a similar hydrodynamic stability.

  12. Optimization of the NIF ignition point design hohlraum

    International Nuclear Information System (INIS)

    Callahan, D A; Hinkel, D E; Berger, R L; Divol, L; Dixit, S N; Edwards, M J; Haan, S W; Jones, O S; Lindl, J D; Meezan, N B; Michel, P A; Pollaine, S M; Suter, L J; Town, R P J; Bradley, P A

    2008-01-01

    In preparation for the start of NIF ignition experiments, we have designed a porfolio of targets that span the temperature range that is consistent with initial NIF operations: 300 eV, 285 eV, and 270 eV. Because these targets are quite complicated, we have developed a plan for choosing the optimum hohlraum for the first ignition attempt that is based on this portfolio of designs coupled with early NIF experiements using 96 beams. These early experiments will measure the laser plasma instabilities of the candidate designs and will demonstrate our ability to tune symmetry in these designs. These experimental results, coupled with the theory and simulations that went into the designs, will allow us to choose the optimal hohlraum for the first NIF ignition attempt

  13. Optimization of the NIF ignition point design hohlraum

    Science.gov (United States)

    Callahan, D. A.; Hinkel, D. E.; Berger, R. L.; Divol, L.; Dixit, S. N.; Edwards, M. J.; Haan, S. W.; Jones, O. S.; Lindl, J. D.; Meezan, N. B.; Michel, P. A.; Pollaine, S. M.; Suter, L. J.; Town, R. P. J.; Bradley, P. A.

    2008-05-01

    In preparation for the start of NIF ignition experiments, we have designed a porfolio of targets that span the temperature range that is consistent with initial NIF operations: 300 eV, 285 eV, and 270 eV. Because these targets are quite complicated, we have developed a plan for choosing the optimum hohlraum for the first ignition attempt that is based on this portfolio of designs coupled with early NIF experiements using 96 beams. These early experiments will measure the laser plasma instabilities of the candidate designs and will demonstrate our ability to tune symmetry in these designs. These experimental results, coupled with the theory and simulations that went into the designs, will allow us to choose the optimal hohlraum for the first NIF ignition attempt.

  14. Design studies of Tokamak power reactor in JAERI

    International Nuclear Information System (INIS)

    Tone, T.; Nishikawa, M.; Tanaka, Y.

    1985-01-01

    Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

  15. Conceptual Design of Alborz Tokamak Poloidal Coils System

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.

    2013-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.

  16. Evaluation of the Revolver Ignition Design at the National Ignition Facility Using Polar-Direct-Drive Illumination

    Science.gov (United States)

    McKenty, P. W.; Collins, T. J. B.; Marozas, J. A.; Campbell, E. M.; Molvig, K.; Schmitt, M.

    2017-10-01

    The direct-drive ignition design Revolver employs a triple-shell target using a beryllium ablator, a copper driver, and an eventual gold pusher. Symmetric numerical calculations indicate that each of the three shells exhibit low convergence ( 3to 5) resulting in a modest gain (G 4) for 1.7 MJ of incident laser energy. Studies are now underway to evaluate the robustness of this design employing polar direct drive (PDD) at the National Ignition Facility. Integral to these calculations is the leveraging of illumination conditioning afforded by research done to demonstrate ignition for a traditional PDD hot-spot target design. Two-dimensional simulation results, employing nonlocal electron-thermal transport and cross-beam energy transport, will be presented that indicate ignition using PDD. A study of the allowed levels of long-wavelength perturbations (target offset and power imbalance) not precluding ignition will also be examined. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  17. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2015-10-15

    Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

  18. ZTI: Preliminary characterization of an ignition class reversed-field pinch

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Werley, K.A.

    1990-01-01

    A preliminary cost-optimized conceptual design of an intermediate-step, ignition-class RFP device (ZTI) for the study of alpha-particle physics in a DT plasma is reported. The ZTI design reflects potentially significant cost savings relative to similar ignition-class tokamaks for device parameters that reside on the path to a viable commercial RFP reactor. Reductions in both device costs and number of steps to commercialization portend a significantly reduced development cost for fusion. The methodology and result and coupling realistic physics, engineering, and cost models through a multi-dimensional optimizer are reported for ZTI, which is a device that would follow the 2--4 MA ZTH on a approx-gt 1996--98 timescale. 15 refs., 7 figs., 2 tabs

  19. Safety implications of a graphite oxidation accident in the compact ignition tokamak device

    International Nuclear Information System (INIS)

    Merrill, B.J.; O'Brien, M.H.

    1989-01-01

    This paper addresses the possible safety consequences of an air ingress accident for the Compact Ignition Tokamak (CIT) device. An experimental program was undertaken to determine oxidation rates of four nuclear grade graphites in air at temperatures ranging from 800 to 1800 C and flow velocities from 3 to 7 m/s. On the basis of these test results, an analytic model was developed to assess the extent of first wall/divertor protective tile oxidation and the amount of energy released from this oxidation. For CIT, a significant restriction to vacuum vessel air inflow will be provided by the air seals and walls of the surrounding test cells. Under these conditions, the graphite oxidation reaction inside the vacuum vessel will become oxygen starved within minutes of the onset of this event. Since significant oxidation rates were not achieved, the heat release did not elevate structural temperatures to levels of concern with regard to activated material release. 7 refs., 9 figs

  20. Target designs for energetics experiments on the National Ignition Facility

    International Nuclear Information System (INIS)

    Meezan, N B; Glenzer, S H; Suter, L J

    2008-01-01

    The goal of the first hohlraum energetics experiments on the National Ignition Facility (NIF) [G. H. Miller et al, Optical Eng. 43, 2841 (2004)] is to select the hohlraum design for the first ignition experiments. Sub-scale hohlraums heated by 96 of the 192 laser beams on the NIF are used to emulate the laser-plasma interaction behavior of ignition hohlraums. These 'plasma emulator' targets are 70% scale versions of the 1.05 MJ, 300 eV ignition hohlraum and have the same energy-density as the full-scale ignition designs. Radiation-hydrodynamics simulations show that the sub-scale target is a good emulator of plasma conditions inside the ignition hohlraum, reproducing density n e within 10% and temperature T e within 15% along a laser beam path. Linear backscatter gain analysis shows the backscatter risk to be comparable to that of the ignition target. A successful energetics campaign will allow the National Ignition Campaign to focus its efforts on optimizing ignition hohlraums with efficient laser coupling

  1. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

    Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

    2000-01-01

    The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

  2. Full power in the European tokamak

    International Nuclear Information System (INIS)

    Lallia, P.P.; Hugon, M.

    1987-01-01

    A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

  3. An interim report on the materials and selection criteria analysis for the Compact Ignition Tokamak toroidal field coil turn-to-turn insulation system

    International Nuclear Information System (INIS)

    Campbell, V.W.; Dooley, J.B.; Hubrig, J.G.; Janke, C.J.; McManamy, T.J.; Welch, D.E.

    1990-01-01

    This document contains photographs showing the results of laboratory testing of the combinations of epoxy resins, hardeners, and cures undertaken as part of the Compact Ignition Tokamak Insulation Screening Program. Cryogenic shock and soak to equilibrium proved to be the most demanding condition for these materials. The degree of damage to the basic materials when a poor candidate is selected is shown to be quite dramatic. 34 figs

  4. Socioeconomic information, Plainsboro area, New Jersey: Supplementary documentation for an environmental assessment for the CIT [Compact Ignition Tokamak] at PPPL

    International Nuclear Information System (INIS)

    Bentz, L.K.; Bender, D.S.

    1987-07-01

    This report contains socioeconomic information on the Plainsboro, New Jersey, area, the proposed location of the Compact Ignition Tokamak (CIT) facility. It was prepared as supplemental information for an environmental assessment for the CIT at Princeton Plasma Physics Laboratory (PPPL). The report contains descriptions of the demographic, economic, and community resource characteristics, and, based on information available in early 1987, considers the socioeconomic effect of the proposed facility. In all areas examined, the anticipated socioeconomic impacts of the proposed CIT facility at PPPL are negligible or minimal. 29 refs., 8 figs., 24 tabs

  5. Methods for the design and optimization of shaped tokamaks

    International Nuclear Information System (INIS)

    Haney, S.W.

    1988-05-01

    Two major questions associated with the design and optimization of shaped tokamaks are considered. How do physics and engineering constraints affect the design of shaped tokamaks? How can the process of designing shaped tokamaks be improved? The first question is addressed with the aid of a completely analytical procedure for optimizing the design of a resistive-magnet tokamak reactor. It is shown that physics constraints---particularly the MHD beta limits and the Murakami density limit---have an enormous, and sometimes, unexpected effect on the final design. The second question is addressed through the development of a series of computer models for calculating plasma equilibria, estimating poloidal field coil currents, and analyzing axisymmetric MHD stability in the presence of resistive conductors and feedback. The models offer potential advantages over conventional methods since they are characterized by extremely fast computer execution times, simplicity, and robustness. Furthermore, evidence is presented that suggests that very little loss of accuracy is required to achieve these desirable features. 94 refs., 66 figs., 14 tabs

  6. Conceptual design of the steady state tokamak reactor (SSTR)

    International Nuclear Information System (INIS)

    Oikawa, A.; Kikuchi, M.; Seki, Y.; Nishio, S.; Ando, T.; Ohara, Y.; Takizuka, Tani, K.; Ozeki, T.; Koizumi, K.; Ikeda, B.; Suzuki, Y.; Ueda, N.; Kageyama, T.; Yamada, M.; Mizoguchi, T.; Iida, F.; Ozawa, Y.; Mori, S.; Yamazaki, S.; Kobayashi, T.; Adachi, H.J.; Shinya, K.; Ozaki, A.; Asahara, M.; Konishi, K.; Yokogawa, N.

    1992-01-01

    This paper reports that on the basis of a high bootstrap current fraction observation with JT-60, the concept of steady state tokamak reactor , the SSTR, was conceived and was evolved with the design activity of the SSTR at JAERI. Also results of ITER/FER design activities has enhanced the SSTR design. Moreover the remarkable progress of R and D for fusion reactor engineering, especially in the development of superconducting coils and negative ion based NBI at JAERI have promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor

  7. Super high field ohmically heated tokamak operation

    International Nuclear Information System (INIS)

    Cohn, D.R.; Bromberg, L.; Leclaire, R.J.; Potok, R.E.; Jassby, D.L.

    1986-01-01

    The authors discuss a super high field mode of tokamak operation that uses ohmic heating or near ohmic heating to ignition. The super high field mode of operation uses very high values of Β/sup 2/α, where Β is the magnetic field and a is the minor radius (Β/sup 2/α > 100 T/sup 2/m). We analyze copper magnet devices with major radii from 1.7 to 3.0 meters. Minimizing or eliminating the need for auxiliary heating has the potential advantages of reducing uncertainty in extrapolating the energy confinement time of current tokamak devices, and reducing engineering problems associated with large auxiliary heating requirements. It may be possible to heat relatively short pulse, inertially cooled tokamaks to ignition with ohmic power alone. However, there may be advantages in using a very small amount of auxiliary power (less than the ohmic heating power) to boost the ohmic heating and provide a faster start-up, expecially in relatively compact devices

  8. Conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR), September 1978

    International Nuclear Information System (INIS)

    Kelley, J.L.

    1978-12-01

    The flexibility of the fusion hybrid reactor to function as a fuel production facility, power plant, waste disposal burner or combinations of all of these, as well as the reactor's ability to use proliferation resistant fuel cycles, has provided the incentive to assess the feasibility of a near-term demonstration plant. The goals for a Demonstration Tokamak Hybrid Reactor (DTHR) were established and an initial conceptual design was selected. Reactor performance and economics were evaluated and key developmental issues were assessed. The study has shown that a DTHR is feasible in the late 1980's, a significant quantity of fissile fuel could be produced from fertile thorium using present day fission reactor blanket technology, and a large number of commercially prototypical components and systems could be developed and operationally verified. The DTHR concept would not only serve as proof-of-principle for hybrid technology, but could be operated in the ignited mode and provide major advancements for pure fusion technology

  9. The design of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.; Kim, J.; Hwang, S.M.

    1999-01-01

    The Korea superconducting tokamak advanced research (KSTAR) project is the major effort of the Korean national fusion program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 MA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron-cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002. (orig.)

  10. Cryogenic system design for a compact tokamak reactor

    International Nuclear Information System (INIS)

    Slack, D.S.; Kerns, J.A.; Miller, J.R.

    1988-01-01

    The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

  11. Application of internally cooled superconductors to tokamak fusion reactors

    International Nuclear Information System (INIS)

    Materna, P.A.

    1986-01-01

    Recent proposals for ignition tokamaks containing superconductors are reviewed. As the funding prospects for the U.S. fusion program have worsened, the proposed designs have been shrinking to smaller machines with less ambitious goals. The most recent proposal, the Tokamak Fusion Core Experiment (TFCX), was based on internally cooled cabled Nb 3 Sn conductors for the options which used superconductors. Internally cooled conductors are particularly advantageous in their electrical insulating properties and in the similarity of their winding procedures to those of conventional copper coils. Epoxy impregnation is possible and is advantageous both structurally and electrically. The allowable current density for this type of conductor was shown to be larger than the current density for more conventional superconducting technology. The TFCX effort identified research and development needed in advance of TFCX or any other large ignition machine. These topics include the metal used for the conduit; nuclear effects on materials; properties of electrical and thermal insulators; extension of superconducting technology to the sizes of coils envisioned and to the field level envisioned; pulsed coil superconducting technology; joints and insulating breaks in conductors; heat removal or flow path length limitations; mechanical behavior of potted conductor bundles; instrumentation; and fault modes and various questions integrated with overall machine design

  12. Optimization design for SST-1 Tokamak insulators

    International Nuclear Information System (INIS)

    Zhang Yuanbin; Pan Wanjiang

    2012-01-01

    With the help of ANSYS FEA technique, high voltage and cryogenic proper- ties of the SST-1 Tokamak insulators were obtained, and the structure of the insulators was designed and modified by taking into account the simulation results. The simulation results indicate that the optimization structure has better high voltage insulating property and cryogenic mechanics property, and also can fulfill the qualification criteria of the SST-1 Tokamak insulators. (authors)

  13. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1981-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and USSR. The Zero-Phase of the INTOR Workshop, which was conducted during 1979, assessed the technical data base that would support the construction of the next major device in the tokamak program to operate in the early 1990s and defined the objectives and characteristics of this device. The INTOR workshop was extended into phase-1, the Definition Phase, in early 1980. The objective of the Phase-1 Workshop was to develop a conceptual design of the INTOR experiment. The purpose of this paper is to give an overview of the work of the Phase-1 INTOR Workshop (January 1980-June 1981, with emphasis upon the conceptual design

  14. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  15. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  16. Conceptual design of Remote Control System for EAST tokamak

    International Nuclear Information System (INIS)

    Sun, X.Y.; Wang, F.; Wang, Y.; Li, S.

    2014-01-01

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

  17. Conceptual design of Remote Control System for EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

    2014-05-15

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

  18. TIBER II: an upgraded tokamak igntion/burn experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Perkins, L.J.

    1986-01-01

    We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

  19. Impact of maximum TF magnetic field on performance and cost of an advanced physics tokamak

    International Nuclear Information System (INIS)

    Reid, R.L.

    1983-01-01

    Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of variation in the maximum value of the field at the toroidal field (TF) coils on the performance and cost of a low q/sub psi/, quasi-steady-state tokamak. Marginal ignition, inductive current startup plus 100 s of inductive burn, and a constant value of epsilon (inverse aspect ratio) times beta poloidal were global conditions imposed on this study. A maximum TF field of approximately 10 T was found to be appropriate for this device

  20. Physics design requirements for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Porkolab, M.; Ulrickson, M.

    1993-01-01

    The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust

  1. Engineering design of TFTR and it's impact on future tokamaks

    International Nuclear Information System (INIS)

    Sabado, M.M.

    1981-01-01

    TFTR is a second generation tokamak whose key objective is scientific break-even. TFTR is expected to be the first machine to demonstrate proper combination of plasma confinement time, density, and temperature to obtain this objective. A summary of major TFTR design parameters, including TFM, is presented, and their potential impact on future tokamaks discussed. Details of the updated engineering design and analysis of components are described. Status of major hardware fabrication, assembly installation and test are reviewed. TFTR features, technology, predicted performance and their potential implication for future tokamaks are summarized

  2. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

    During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

  3. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1996-01-01

    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

  4. The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

    International Nuclear Information System (INIS)

    Atzeni, S.; Vlad, G.

    1985-01-01

    The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D- 3 He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D- 3 He tokamak plasma

  5. Short-wavelength and three-dimensional instability evolution in National Ignition Facility ignition capsule designs

    International Nuclear Information System (INIS)

    Clark, D. S.; Haan, S. W.; Cook, A. W.; Edwards, M. J.; Hammel, B. A.; Koning, J. M.; Marinak, M. M.

    2011-01-01

    Ignition capsule designs for the National Ignition Facility (NIF) [G. H. Miller, E. I. Moses, and C. R. Wuest, Opt. Eng. 443, 2841 (2004)] have continued to evolve in light of improved physical data inputs, improving simulation techniques, and, most recently, experimental data from a growing number of NIF sub-ignition experiments. This paper summarizes a number of recent changes to the cryogenic capsule design and some of our latest techniques in simulating its performance. Specifically, recent experimental results indicated harder x-ray drive spectra in NIF hohlraums than were predicted and used in previous capsule optimization studies. To accommodate this harder drive spectrum, a series of high-resolution 2-D simulations, resolving Legendre mode numbers as high as 2000, were run and the germanium dopant concentration and ablator shell thicknesses re-optimized accordingly. Simultaneously, the possibility of cooperative or nonlinear interaction between neighboring ablator surface defects has motivated a series of fully 3-D simulations run with the massively parallel HYDRA code. These last simulations include perturbations seeded on all capsule interfaces and can use actual measured shell surfaces as initial conditions. 3-D simulations resolving Legendre modes up to 200 on large capsule sectors have run through ignition and burn, and higher resolution simulations resolving as high as mode 1200 have been run to benchmark high-resolution 2-D runs. Finally, highly resolved 3-D simulations have also been run of the jet-type perturbation caused by the fill tube fitted to the capsule. These 3-D simulations compare well with the more typical 2-D simulations used in assessing the fill tube's impact on ignition. Coupled with the latest experimental inputs from NIF, our improving simulation capability yields a fuller and more accurate picture of NIF ignition capsule performance.

  6. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

    1993-01-01

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  7. Design of a deuterium and tritium-ablator shock ignition target for the National Ignition Facility

    International Nuclear Information System (INIS)

    Terry, Matthew R.; Perkins, L. John; Sepke, Scott M.

    2012-01-01

    Shock ignition presents a viable path to ignition and high gain on the National Ignition Facility (NIF). In this paper, we describe the development of the 1D design of 0.5 MJ class, all-deuterium and tritium (fuel and ablator) shock ignition target that should be reasonably robust to Rayleigh-Taylor fluid instabilities, mistiming, and hot electron preheat. The target assumes “day one” NIF hardware and produces a yield of 31 MJ with reasonable allowances for laser backscatter, absorption efficiency, and polar drive power variation. The energetics of polar drive laser absorption require a beam configuration with half of the NIF quads dedicated to launching the ignitor shock, while the remaining quads drive the target compression. Hydrodynamic scaling of the target suggests that gains of 75 and yields 70 MJ may be possible.

  8. Design of the ITER tokamak assembly tools

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyunki [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)], E-mail: hkpark@nfri.re.kr; Lee, Jaehyuk; Kim, Taehyung [SFA Engineering Corp., 42-7 Palyong-dong, Changwon-si, Gyeongsangnam-do 641-847 (Korea, Republic of); Song, Yunju [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of); Im, Kihak [ITER Organization, CEA Cadarasche, 13108 Saint Paul-lez-Durance (France); Kim, Byungchul; Lee, Hyeongon; Jung, Ki-Jung [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)

    2008-12-15

    ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values.

  9. Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

    International Nuclear Information System (INIS)

    Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

    1990-05-01

    This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

  10. Design of the ITER Tokamak Assembly Tools

    International Nuclear Information System (INIS)

    Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

    2006-01-01

    ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

  11. Design, simulation and construction of the Taban tokamak

    Science.gov (United States)

    H, R. MIRZAEI; R, AMROLLAHI

    2018-04-01

    This paper describes the design and construction of the Taban tokamak, which is located in Amirkabir University of Technology, Tehran, Iran. The Taban tokamak was designed for plasma investigation. The design, simulation and construction of essential parts of the Taban tokamak such as the toroidal field (TF) system, ohmic heating (OH) system and equilibrium field system and their power supplies are presented. For the Taban tokamak, the toroidal magnetic coil was designed to produce a maximum field of 0.7 T at R = 0.45 m. The power supply of the TF was a 130 kJ, 0–10 kV capacitor bank. Ripples of toroidal magnetic field at the plasma edge and plasma center are 0.2% and 0.014%, respectively. For the OH system with 3 kA current, the stray field in the plasma region is less than 40 G over 80% of the plasma volume. The power supply of the OH system consists of two stages, as follows. The fast bank stage is a 120 μF, 0–5 kV capacitor that produces 2.5 kA in 400 μs and the slow bank stage is 93 mF, 600 V that can produce a maximum of 3 kA. The equilibrium system can produce uniform magnetic field at plasma volume. This system’s power supply, like the OH system, consists of two stages, so that the fast bank stage is 500 μF, 800 V and the slow bank stage is 110 mF, 200 V.

  12. U.S.S.R. eyes role in U.S. compact tokamak ignition experiment

    International Nuclear Information System (INIS)

    Crawford, M.

    1987-01-01

    Physicists working on nuclear fusion in the Soviet Union want to participate in America's next experiment, the Compact Ignition Tokamak (CIT). Soviet scientists first mentioned the idea to Department of Energy (DOE) officials last spring but have not put forth any formal proposal. Department officials, however, say the Soviets are serious about working with the US on the project. From the outset, however, efforts to work more closely with the Soviets on fusion research have been stifled by concerns within the Administration that sensitive Western technology will be transferred to the Soviet Union's military establishment. The key to Soviet participation in CIT may depend on what they can contribute materially, as well as intellectually to the US project. Soviet scientists are currently defining how their country might contribute to the CIT, and they could submit a proposal to the US within a year. Indeed, the technology transfer issue has made DOE and State Department officials reluctant to discuss the possibility of the Soviet Union taking a role in the CIT. They fear that the concept will be rejected on ideological grounds before the merits of a Soviet proposal can be considered

  13. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  14. Plasma engineering assessments of compact ignition experiments

    International Nuclear Information System (INIS)

    Houlberg, W.A.

    1985-01-01

    Confinement, startup sequences, and fast-alpha particle effects are assessed for a class of compact tokamak ignition experiments having high toroidal magnetic fields (8 to 12 T) and high toroidal currents (7 to 10 MA). The uncertainties in confinement scaling are spanned through examples of performance with an optimistic model based on ohmically heated plasmas and a pessimistic model that includes confinement degradation by both auxiliary and alpha heating. The roles of neoclassical resistivity enhancement and sawtooth behavior are also evaluated. Copper toroidal field coils place restrictions on pulse lengths due to resistive heating, so a simultaneous rampup of the toroidal field and plasma current is proposed as a means of compressing the startup phase and lengthening the burn phase. If the ignition window is small, fast-alpha particle physics is restricted to the high-density regime where a short slowing-down time leads to low fast-particle density and pressure contributions. Under more optimistic confinement, a larger ignition margin broadens the range of alpha particle physics that can be addressed. These issues are illustrated through examples of transport simulations for a set of machine parameters called BRAND-X, which typify the designs under study

  15. Plasma engineering assessments of compact ignition experiments

    International Nuclear Information System (INIS)

    Houlberg, W.A.

    1986-01-01

    Confinement, startup sequences, and fast-alpha particle effects are assessed for a class of compact tokamak ignition experiments having high toroidal magnetic fields (8-12 T) and high toroidal currents (7-10 MA). The uncertainties in confinement scaling are spanned through examples of performance with an optimistic model based on ohmically heated plasmas and a pessimistic model that includes confinement degradation by both auxiliary and alpha heating. The roles of neoclassical resistivity enhancement and sawtooth behavior are also evaluated. Copper toroidal field coils place restrictions on pulse lengths due to resistive heating, so a simultaneous rampup of the toroidal field and plasma current is proposed as a means of compressing the startup phase and lengthening the burn phase. If the ignition window is small, fast-alpha particle physics is restricted to the high-density regime where a short slowing-down time leads to low fast-particle density and pressure contributions. Under more optimistic confinement, a larger ignition margin broadens the range of alpha particle physics that can be addressed. These issues are illustrated through examples of transport simulations for a set of machine parameters called BRAND-X, which typify the designs under study

  16. National Ignition Facility Title II Design Plan

    International Nuclear Information System (INIS)

    Kumpan, S

    1997-01-01

    This National Ignition Facility (NIF) Title II Design Plan defines the work to be performed by the NIF Project Team between November 1996, when the U.S. Department of Energy (DOE) reviewed Title I design and authorized the initiation of Title H design and specific long-lead procurements, and September 1998, when Title 11 design will be completed

  17. Particle injection into the Castor tokamak by electric arcs

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

    1989-01-01

    The influence of arcing on the tokamak discharge was investigated in the Castor tokamak. A special calibrated gun which emitted tantalum by artificially ignited electric arcs, was used to study the transport of the injected tantalum ions, neutrals and droplets. The injection of tantalum led to an increase in electron density and to a change of plasma position only if the transported charge was higher than 0.01 C. As the naturally occurring arcs are well below this limit, the arcing in tokamaks is rather the consequence than the reason of instabilities. (J.U.)

  18. Design and construction of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.

    2001-01-01

    The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and 'advanced tokamak' physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long-pulse current-drive and heating systems. The 'advanced tokamak' aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. (author)

  19. Insulation irradiation test programme for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    McManamy, T.J.; Kanemoto, G.; Snook, P.

    1991-01-01

    In a programme to evaluate the effects of radiation exposure on the electrical insulation for the toroidal field coils of the Compact Ignition Tokamak, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1 and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to ≅ 5 x 10 9 and 3 x 10 10 rad with 35-40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was performed by cycling the shear load for up to 30000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths of the order of 120 MPa were measured. The behaviour of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed almost identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. No swelling was measured; however, the epoxy samples did twist slightly. (author)

  20. Design approach of seismic interface for cryoline with Tokamak building for ITER

    International Nuclear Information System (INIS)

    Badgujar, S.; Sarkar, B.; Vaghela, H.; Shah, N.; Naik, H.B.

    2012-01-01

    ITER Tokamak building is designed with seismic isolation pads to protect the Tokamak components from seismic events. Two main cryolines, designated as cryolines between buildings (Mg and CP), runs from interconnection box in cryoplant building to the Tokamak building. The lines outside Tokamak building are supported by seismically non-isolated supports. The cryoline design at the interface between seismically isolated and non-isolated support systems needs to be studied to fulfill the functional requirements. One of the options for interface, universal expansion joint has been modeled in CATIA with actual thickness of each ply and inter-ply distance, analyzed in ANSYS using contact definition, as a part of the preliminary study. The bellows have been checked by design calculation as per EJMA standard for the specified movements. The paper will present approach for conceptual design of interface, problem definition and boundary conditions, methodology for analysis and preliminary results of stress pattern for expansion joints. (author)

  1. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of inductively driven long pulse tokamak reactors: IDLT

    International Nuclear Information System (INIS)

    Ogawa, Yuichi

    1998-01-01

    Based on scientific data based adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R and D program, the scientific feasibility of inductively-driven tokamak fusion reactors is studied. A low wall-loading DEMO fusion reactor is designed, which utilizes an austenitic stainless steel in conjunction with significant data bases and operating experiences, since we have given high priority to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the DEMO reactor with the relatively large volume (i.e., major radius of 10 m) is employed, plasma ignition is achievable with a low fusion power of 0.8 GW, and an operation period of 4 - 5 hours is available only with inductive current drive. Disadvantages of pulsed operation in commercial fusion reactors include fatigue in structural materials and the necessity of an energy storage system to compensate the electric power during the dwell time. To overcome these disadvantages, a pulse length is prolonged up to about 10 hours, resulting in the remarkable reduction of the total cycle number to 10 4 during the life of the fusion plant. (author)

  2. Interim report on the assessment of engineering issues for compact high-field ignition devices

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1986-04-01

    The engineering issues addressed at the workshop included the overall configuration, layout, and assembly; limiter and first-wall energy removal; magnet system structure design; fabricability; repairability; and costs. In performing the assessment, the primary features and characteristics of each concept under study were reviewed as representative of this class of ignition device. The emphasis was to understand the key engineering areas of concern for this class of device and deliberately not attempt to define an optimum design or to choose a best approach. The assessment concluded that compact ignition tokamaks, as represented by the three concepts under study, are feasible. A number of critical engineering issues were identified, and all appear to have tractable solutions. The engineering issues appear quite challenging, and to obtain increased confidence in the apparent design solutions requires completion of the next level of design detail, complemented by appropriate development programs and testing

  3. Recent developments in engineering and technology concepts for prospective tokamak fusion reactors

    International Nuclear Information System (INIS)

    Ford, G.W.K.

    1987-01-01

    The tokamak has become the most developed magnetic fusion system and it appears likely that break-even and possibly ignition will first be demonstrated in existing machines of this type. Yet larger tokamaks could also demonstrate the essential technologies for the production of useful power. World-wide, well over a hundred tritium-breeder/heat-removal blanket concepts have been devised and preliminary engineering design studies undertaken, but the effort deployed on breeding and power recovery systems has been very small compared with that assigned to plasma research and development. The European Communities' NET (Next European Torus) project may offer an opportunity to redress this imbalance. The NET pre-design stage now in progress for some three years has selected many of the best features of plasma and nuclear design from the world's total efforts in these fields, and the NET concept is described in this paper as exemplifying where magnetic fusion power reactor technology stands today. It is concluded that although there are numerous more advanced types of magnetic confinement fusion reactor at early stages of their physics development, the tokamak offers the best opportunity for the early demonstration of fusion power

  4. Design considerations for the TF center conductor post for the Ignition Spherical Torus (IST)

    International Nuclear Information System (INIS)

    Dalton, G.R.; Haines, J.R.

    1986-01-01

    A trade-off study has been carried out to compare the differential costs of using high-strength alloy copper versus oxygen-free, high-conductivity (OFHC) copper for the center legs of the toroidal field (TF) coils of an Ignition Spherical Torus (IST). The electrical heating, temperatures, stresses, cooling requirements, material costs, pump costs, and power to drive the TF coils and pumps are all assessed for both materials for a range of compact tokamak reactors. The alloy copper material is found to result in a more compact reactor and to allow use of current densities of up to 170 MA/m 2 versus 40 MA/m 2 for the OFHC copper. The OFHC conductor system with high current density is $24 million less expensive than more conventional copper systems with 30 MA/m 2 . The alloy copper system costs $32 million less than conventional systems. Therefore, the alloy system offers a net savings of $8 million compared to the 50% cold-worked OFHC copper system. Although the savings are a significant fraction of the center conductor post cost, they are relatively insignificant in terms of the total device cost. It is concluded that the use of alloy copper contributes very little to the economic or technical viability of the compact IST. It is recommended that a similar systematic approach be applied to evaluating coil material and current density trade-offs for other compact copper-TF-coil tokamak designs. 9 refs., 13 figs., 13 tabs

  5. Revised design for the Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1977-03-01

    A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost

  6. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    Marinak, M.

    1988-01-01

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

  7. Possibilities for breakeven and ignition of D-3He fusion fuel in a near term tokamak

    International Nuclear Information System (INIS)

    Emmert, G.A.; El-Guebaly, L.; Kulcinski, G.L.; Santarius, J.F.; Scharer, J.E.; Sviatoslavsky, I.N.; Walstrom, P.L.; Klinghoefer, R.; Wittenberg, J.L.

    1988-09-01

    The recent realization that the moon contains a large amount of the isotope 3 He has rekindled interest in the D- 3 He fuel cycle. In this study we consider the feasibility of investigating D- 3 He reactor plasma conditions in a tokamak of the NET/INTOR class. We have found that, depending on the energy confinement scaling law, energy breakeven may be achieved without significant modification to the NET design. The best results are for the more optimistic ASDEX H-mode scaling law. Kaye-Goldston scaling with a modest improvement due to the H-mode is more pessimistic and makes achieving breakeven more difficult. Significant improvement in Q (ratio of the fusion power to the injected power), or the ignition margin, can be achieved by taking advantage of the much reduced neutron production of the D- 3 He fuel cycle. Removal of the tritium producing blanket and replacing the inboard neutron shield by a thinner shield optimized for the neutron spectrum in D- 3 He allows the plasma to be increased without changing the magnetic field at the toroidal field magnet. This allows the plasma to achieve higher beta and Q values up to about 3. The implications of D- 3 He operation for fast ion loss, neutron shielding, heat loads on the first wall and divertor, plasma refuelling, changes to the poloidal field coil system, and pumping of the helium from the vacuum chamber are considered in the report. (orig.)

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

    International Nuclear Information System (INIS)

    Lee, V.D.

    1987-01-01

    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 development of the hydraulic press used to vertically preload the CIT device

  9. A conceptual design of a negative-ion-grounded advanced tokamak reactor

    International Nuclear Information System (INIS)

    Yamamoto, Shin; Ohara, Yoshihiro; Tani, Keiji

    1988-05-01

    The NAVIGATOR concept is based on the negative-ion-grounded 500 keV 20 MW neutral beam injection system (NBI system), which has been proposed and studied at JAERI. The NAVIGATOR concept contains two categories; one is the NAVIGATOR machine as a tokamak reactor, and the other is the NAVIGATOR philosophy as a guiding principle in fusion research. The NAVIGATOR machine implies an NBI heated and full inductive ramped-up reactor. The NAVIGATOR concept should be applied in a phased approach to and beyond the operating goal for the FER (Fusion Experimental Reactor, the next generation tokamak machine in Japan). The mission of the FER is to realize self-ignition and a long controlled burn of about 800 seconds and to develop and test fusion technologies, including the tritium fuel cycle, superconducting magnet, remote maintenance and breeding blanket test modules. The NAVIGATOR concept is composed of three major elements, that is, reliable operation scenarios, reliable maintenability and sufficient flexibility of the reactor. The NAVIGATOR concept well supports the ideas of phased operation and phased construction of the FER, which will result in the reduction of technological risk. The NAVIGATOR concept is expected to bring forth the fruits growing up in the present large tokamak machines in the form of next generation machines. In addition, the NAVIGATOR concept will supply many required databases for the DEMO reactor. The details of the NAVIGATOR concept is described in this paper, and the concept may indicate a feasible strategy for developing fusion research. (author)

  10. Transients and burn dynamics in advanced tokamak fusion reactors

    International Nuclear Information System (INIS)

    Mantsinen, M.J.; Salomaa, R.R.E.

    1994-01-01

    Transient behavior of D 3 He-tokamak reactors is investigated numerically using a zero-dimensional code with prescribed profiles. Pure D 3 He start-up is compared to DT-assisted and DT-ignited start-ups. We have considered two categories of transients which could extinguish steady fusion burn: fuelling interruptions and sudden confinement changes similar to the L → H transients occurring in present-day tokamaks. Shutdown with various current and density ramp-down scenarios are studied, too. (author)

  11. Design and construction of electronic components for a ''Novillo'' Tokamak

    International Nuclear Information System (INIS)

    Lopez C, R.

    1986-07-01

    The goal of this effort was to design, construct and make functional the electronic components for a ''Novillo'' Tokamak currently being experimentally investigated at the National Institute of Nuclear Research in Mexico. The problem was to develop programmable electronic switches capable of discharging high voltage kilowatt energies stored in capacitator banks onto the coils of the Tokamak. (author)

  12. Neutronics design of the next tokamak. (Swimming pool type)

    International Nuclear Information System (INIS)

    Seki, Y.; Iida, H.; Kitamura, K.; Minato, A.; Sako, K.; Mori, S.; Nishida, H.

    1983-01-01

    A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

  13. Target design for shock ignition

    International Nuclear Information System (INIS)

    Schurtz, G; Ribeyre, X; Lafon, M

    2010-01-01

    The conventional approach of laser driven inertial fusion involves the implosion of cryogenic shells of deuterium-tritium ice. At sufficiently high implosion velocities, the fuel ignites by itself from a central hot spot. In order to reduce the risks of hydrodynamic instabilities inherent to large implosion velocities, it was proposed to compress the fuel at low velocity, and ignite the compressed fuel by means of a convergent shock wave driven by an intense spike at the end of the laser pulse. This scheme, known as shock ignition, reduces the risks of shell break-up during the acceleration phase, but it may be impeded by a low coupling efficiency of the laser pulse with plasma at high intensities. This work provides a relationship between the implosion velocity and the laser intensity required to ignite the target by a shock. The operating domain of shock ignition at different energies is described.

  14. Neutron monitoring measurements for the CIT [Compact Ignition Tokamak] materials irradiations in the ATR I1 position

    International Nuclear Information System (INIS)

    Rogers, J.W.; Anderl, R.A.

    1989-12-01

    Measurements were performed to help characterize the neutron environments in which the Compact Ignition Tokamak (CIT) materials were irradiated. These materials were irradiated in a lead shield plug assembly at the ATR I1 position. Neutron monitor materials were placed in the capsules in proximity with the CIT specimens. The neutron monitors sensed the neutrons through reactions that have different neutron energy region responses. By measuring the radioactivity of the neutron monitors it was possible to determine the neutron fluence rates (n/cm 2 /sec) and fluences (n/cm 2 ) at the locations of the monitors. It was also possible to determine the axial and radial gradients of the neutron environments near the specimens. This report presents the results obtained from these measurements for both the CIT number-sign 1 (ORNL 64-2) and CIT number-sign 2 (ORNL 64-1) capsules. In general, ASTM methods and procedures were used in all neutron monitoring associated activities. 7 refs., 9 figs., 10 tabs

  15. Configuration and engineering design of the ARIES-RS tokamak power plant

    International Nuclear Information System (INIS)

    Tillack, M.S.; Malang, S.; Waganer, L.; Wang, X.R.; Sze, D.K.; El-Guebaly, L.; Wong, C.P.C.; Crowell, J.A.; Mau, T.K.; Bromberg, L.

    1997-01-01

    ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems. (orig.)

  16. Comments on thermal runaway experiments in sub-ignition tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.

    1982-09-01

    Justification of deuterium-tritium operations is discussed from the physics viewpoint and optimal thermal runaway experiments in high-field, high-density compact tokamaks are suggested within the minimization of the induced radioactivation. (author)

  17. Electron cyclotron heating studies of the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Porkolab, M.; Bonoli, P.T.; Englade, R.; Myer, R.; Smith, G.R.; Kritz, A.H.

    1989-01-01

    The Compact Ignition Tokamak (CIT) operating scenario calls for ramping the toroidal magnetic field from B/sub T/ = 7.0 (8.0) to 10.0 Tesla in a few seconds, followed by a burn cycle and a ramp-down cycle. Simultaneously, the plasma must be heated from an initial low beta equilibrium (/bar /beta// ≅ 0.44% at 7.0 to 8.0 Tesla) to a final burn equilibrium (/bar /beta// = 2.8%) having 10.0 Tesla on the magnetic axis. Since the toroidal plasma current will be ramped at the same time and since the available time for flat-top magnetic field must be reserved for the burn cycle, it is imperative that densification and heating be carried out as the magnetic field is ramped. Here we examine an approach which is applicable to ECR heating. The frequency remains constant, while the angle of injection is varied by simply rotating a reflecting mirror placed in the path of the incident microwave beam. The rotating mirror permits one to launch waves with sufficiently high N/sub /parallel// so that the Doppler broadened resonance of particles on the magnetic axis with f = 280 GHz and B/sub T/ = 7.0--8.0 Tesla can provide adequate absorption. As the resonance layer moves toward the magnetic axis the beam is swept toward perpendicular to reduce the Doppler width and avoid heating the plasma edge. At B/sub T/ = 10.0 Tesla the beam will be at normal incidence with strong absorption immediately on the high field side of the resonance (relativistic regime). We envisage using the ordinary mode (O-mode, /rvec E//sub RF/ /parallel/ /rvec B/) of polarization which is accessible from the outside (low-field side) of the torus provided the density is such that ω/sub pe/ ≤ ω ∼ ω/sub ce/ (max). 8 refs., 3 figs

  18. Compact tokamak reactors

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. 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 tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, 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 recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  19. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Mardani, M.; Amrollahi, R.; Koohestani, S.

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  20. Plasma engineering analyses of tokamak reactor operating space

    International Nuclear Information System (INIS)

    Houlberg, W.; Attenberger, S.E.

    1981-01-01

    A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models

  1. Approach to decision modeling for an ignition test reactor

    International Nuclear Information System (INIS)

    Howland, H.R.; Varljen, T.C.

    1977-01-01

    A comparison matrix decision model is applied to candidates for a D-T ignition tokamak (TNS), including assessment of semi-quantifiable or judgemental factors as well as quantitative ones. The results show that TNS is mission-sensitive with a choice implied between near-term achievability and reactor technology

  2. Design features of HTMR-hybrid toroidal magnet tokamak reactor

    International Nuclear Information System (INIS)

    Rosatelli, F.; Avanzini, P.G.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Brunelli, B.; Zampaglione, V.

    1984-01-01

    The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfil the scientific and technological objectives expected from next generation devices with size and costs as small as possible. A hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. The optimization procedure for the hybrid magnet, configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils are described. (author)

  3. Some problems of a tokamak with fast-neutral injection

    International Nuclear Information System (INIS)

    Pistunovich, V.I.

    1976-01-01

    An attempt has been made to formulate the fundamental physical problems which should be solved for designing a reactor on the basis of a tokamak with injection. The problems enumerated should be considered particularly for the suggested system, though qualitative solutions of some of them have been already known. They are the following. 1) Ionization of the atomic beam and distribution of the generated ion density. 2) Capture of fast ions by the tokamak magnetic field. 3) Anisotropy of the velocity distribution function. 4) The effect of the neutral gas density on the shape of the one-dimensional distribution function. 5)Stability of the ion distribution function on deceleration of the ion beam in plasma. The paper permits evaluating the advantages and shortcomings of the reactor on the basis of a tokamak with injection from viewpoints of the requirements on confinement and heating of plasma, and of the effect of impurities and neutral working gas on the output parameters of the reactor. The author believes that the regime of a two-component tokamak with small values of quality Q (<=) 3 shows greatest advantages over the ignition reactor, though in this case a reactor may be economically profitable (Q approximately 10) only with an increase of the total quality of the reactor by additional energy release in a blanket on using fission reactions

  4. Design considerations of modular pump limiters for large tokamaks

    International Nuclear Information System (INIS)

    Uckan, T.; Klepper, C.C.; Mioduszewski, P.K.

    1987-11-01

    Long-pulse (>10-s) and high-power (>10-MW) operation of large tokamaks requires multiple limiter modules for particle and heat removal, and the power load must be distributed among a number of modules. Because each added module changes the performance of all the others, a set of design criteria must be defined for the overall limiter system. The relationship between individual modules must also be considered from the standpoint of flux coverage and shadowing effects. This paper addresses these issues and provides design guidelines. Parameters of the individual modules are then determined from the system requirements for particle and power removal. Long-pulse operation of large tokamaks requires that the limiter modules be equipped with active cooling. At the leading edge of a module, the cooling channel determines the thickness of the limiter blade (or head). A model has been developed for estimating the system exhaust efficiency in terms of the parameters of the leading edge (i.e., its thickness and the design heat flux) in terms of given device parameters and the power load that must be removed. The impact on module design of state-of-the-art engineering technology for high heat removal is discussed. The choice of locations for the modules is also investigated, and the effects of shadowing between modules on particle and power removal are examined. The results are applied to the Tore Supra tokamak. Conceptual design parameters of the modular pump limiter system are given. 10 refs., 5 figs

  5. HT-7U superconducting tokamak: Physics design, engineering progress and schedule

    International Nuclear Information System (INIS)

    Wan Yuanxi

    2002-01-01

    The superconducting tokamak research program begun in China in ASIPP since 1994. The program is included in existent superconducting tokamak HT-7 and the next new superconducting tokamak HT-7U which is one of national key research projects in China. With the elongation cross-section, divertor and higher plasma parameter the main objectives of HT-7U are widely investigation both of the physics and technology for steady state advanced tokamak as well as the investigation of power and particle handle under steady-state operation condition. The physics and engineering design have been completed and significant progresses on R and D and fabrication have been achieved. HT-7U will begin assembly at 2003 and possible to get first plasma around 2004. (author)

  6. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

    Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

  7. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  8. Near-term tokamak-reactor designs with high-performance resistive magnets

    International Nuclear Information System (INIS)

    Cohn, D.R.; Bromberg, L.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Advanced Fusion Test Reactors (AFTR) designs have been developed using BITTER type magnets which are capable of steady state operation. The goals of compact AFTR designs (with major radii R approx. 2.5 - 4 m), include DT ignition with large physics margins; high duty cycle, long pulse operation; and DD-DT operation with low tritium concentration. Larger AFTR designs (R approx. 5 m), have the additional goal of early demonstration of self sufficiency in tritium production. The AFTR devices could also serve as prototypes for commercial reactors. Compact ignition test reactors have also been designed (R approx. 1 - 2 m). These designs use BITTER magnets that are inertially cooled starting at liquid nitrogen temperature. A detailed engineering design was developed for ZEPHYR

  9. A spheromak ignition experiment reusing Mirror Fusion Test Facility (MFTF) equipment

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1993-01-01

    Based on available experimental results and theory, a scenario is presented to achieve ohmic ignition in a spheromak by slow (∼ 10 sec.) helicity injection using power from the Mirror Fusion Test Facility (MFTF) substation. Some of the other parts needed (vacuum vessel, coils, power supplies, pumps, shielded building space) might also be obtained from MFTF or other salvage, as well as some components needed for intermediate experiments for additional verification of the concept (especially confinement scaling). The proposed ignition experiment would serve as proof-of-principle for the spheromak DT fusion reactor design published by Hagenson and Krakowski, with a nuclear island cost about ten times less than a tokamak of comparable power. Designs at even higher power density and lower cost might be possible using Christofilos' concept of a liquid lithium blanket. Since all structures would be protected from neutrons by the lithium blanket and the tritium inventory can be reduced by continuous removal from the liquid blanket, environmental and safety characteristics appear to be favorable

  10. National Ignition Facility design focuses on optics

    International Nuclear Information System (INIS)

    Hogan, W.J.; Atherton, L.J.; Paisner, J.A.

    1996-01-01

    Sometime in the year 2002, scientists at the National Ignition Facility (NIF) will focus 192 separate high-power ultraviolet laser beams onto a tiny capsule of deuterium and tritium, heating and compressing the material until it ignites and burns with a burst of fusion energy. The mission of NIF, which will contain the largest laser in the world, is to obtain fusion ignition and gain and to use inertial confinement fusion capabilities in nuclear weapons science experiments. The physics data provided by NIF experiments will help scientists ensure nuclear weapons reliability without the need for actual weapons tests; basic sciences such as astrophysics will also benefit. The facility faces stringent weapons-physics user requirements demanding peak pulse powers greater than 750 TW at 0.35 microm (only 500 TW is required for target ignition), pulse durations of 0.1 to 20 ns, beam steering on the order of several degrees, and target isolation from residual 1- and 0.5-microm radiation. Additional requirements include 50% fractional encircled beam energy in a 100-microm-diameter spot, with 95% encircled in a 200-microm spot. The weapons-effects community requires 1- and 0.5-microm light on target, beam steering to widely spaced targets, a target chamber accommodating oversized objects, well-shielded diagnostic areas, and elimination of stray light in the target chamber. The beamline design, amplifier configuration and requirements for optics are discussed here

  11. Design features of HTMR-Hybrid Toroidal Magnet Tokamak Reactor

    International Nuclear Information System (INIS)

    Rosatelli, F.; Avanzini, P.G.; Brunelli, B.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Zampaglione, V.

    1985-01-01

    The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfill the scientific and technological objectives expected from next generation devices (e.g. INTOR-NET) with size and costs as small as possible. An hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. In this paper the authors describe the optimization procedure for the hybrid magnet configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils

  12. A Fast Shutdown Technique for Large Tokamaks

    International Nuclear Information System (INIS)

    Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

    1999-01-01

    A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

  13. Evaluation of weldments in Type 21-6-9 stainless steel for Compact Ignition Tokamak structural applications: Phase 1

    International Nuclear Information System (INIS)

    Alexander, D.J.; Goodwin, G.M.; Bloom, E.E.

    1991-06-01

    Primary design considerations for the Compact Ignition Tokamak toroidal field-coil cases are yield strength and toughness in the temperature range from 77 to 300 K. Type 21-6-9 stainless steel, also still known by its original Armco Steel Company trade name Nitronic 40, is the proposed alloy for this application. It has high yield strength and usually adequate base metal toughness, but weldments in thick sections have not been adequately characterized in terms of mechanical properties or hot-cracking propensity. In this study, weldability of the alloy in heavy sections and the mechanical properties of the resultant welds were investigated including tensile yield strength and Charpy V-notch toughness at 77 K and room temperature. Weldments were made in four different base metals using seven different filler metals. None of the weldments showed any indication of hot-cracking problems. All base metals, including weldment heat-affected zones, were found to have adequate strength and impact toughness at both test temperatures. Weld metals, on the other hand, except ERNiCr-3 and ENiCrFe-3 had impact toughnesses of less than 67 J at 77 K. Inconel 82 had an average weld metal impact toughness of over 135 J at 77 K, and although its strength at 77 K is less than that of type 21-6-9 base metal, at this point it is considered to be the first-choice filler metal. Phase 2 of this program will concentrate on composition refinement and process/procedure optimization for the generic ERNiCr-3 composition and will generate a design data base for base and weld metal, including tensile, fracture toughness, and crack growth rate data

  14. Conceptual design of a Tokamak hybrid power reactor (THPR)

    International Nuclear Information System (INIS)

    Matsuoka, F.; Imamura, Y.; Inoue, M.; Asami, N.; Kasai, M.; Yanagisawa, I.; Ida, T.; Takuma, T.; Yamaji, K.; Akita, S.

    1987-01-01

    A conceptual design of a fusion-fission hybrid tokamak reactor has been carried out to investigate the engineering feasibility and promising scale of a commercial hybrid reactor power plant. A tokamak fusion driver based on the recent plasma scaling law is introduced in this design study. The major parameters and features of the reactor are R=6.06 m, a=1.66 m, Ip=11.8 MA, Pf=668 MW, double null divertor plasma and steady state burning with RF current drive. The fusion power has been determined with medium energy multiplication in the blanket so as to relieve thermal design problems and produce electric power around 1000 MW. Uranium silicide is used for the fast fission blanket material to promise good nuclear performance. The coolant of the blanket is FLIBE and the tritium breeding blanket material is Li 2 O ceramics providing breeding ratio above unity

  15. Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

    International Nuclear Information System (INIS)

    Sink, D.A.; Gibson, G.

    1979-03-01

    The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based on two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma

  16. The Tokamak Fusion Core Experiment studies

    International Nuclear Information System (INIS)

    Schmidt, J.A.; Sheffield, G.V.; Bushnell, C.

    1985-01-01

    The basic objective of the next major step in the US fusion programme has been defined as the achievement of ignition and long pulse equilibrium burn of a fusion plasma in the Tokamak Fusion Core Experiment (TFCX) device. Preconceptual design studies have seen completion of four candidate versions to provide the comparative information needed to narrow down the range of TFCX options before proceeding to the conceptual design phase. All four designs share the same objective and conform to common physics, engineering and costing criteria. The four base options considered differed mainly in the toroidal field coil design, two employing superconducting coils and the other two copper coils. In each case (copper and superconducting), one relatively conventional version was carried as well as a version employing more exotic toroidal field coil design assumptions. Sizes range from R=2.6 m for the smaller of the two copper versions to R=4.08 m for the larger superconducting option. In all cases, the plasma current was about 10 MA and the toroidal field about 4 T. (author)

  17. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

  18. Population projections, 0 to 50 mile radius from the CIT [Compact Ignition Tokamak] facility: Supplementary documentation for an environmental assessment for the CIT at PPPL

    International Nuclear Information System (INIS)

    Bentz, L.K.; Bender, D.S.

    1987-07-01

    This report contains estimates for current and projected population distributions for a 50-mile radius area around the Princeton Plasma Physics Laboratory (PPPL) in central New Jersey. It was prepared as supplemental information for an environmental assessment for the proposed Compact Ignition Tokamak facility at PPPL. The report contains estimates for the 1985 population, as well as projections for 1995, 2000, 2005 and 2010. Estimates are provided for municipalities and counties in the area, as well as for a sectorized annular grid centered at PPPL. Sectorized maps are included. An appendix contains technical details on the methodology used. 17 refs., 10 figs., 13 tabs

  19. Review of tokamak power reactor and blanket designs in the United States

    International Nuclear Information System (INIS)

    Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

    1986-01-01

    The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

  20. The insulation irradiation test program for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    McManamy, T.J.; Kanemoto, G.; Snook, P.

    1990-01-01

    The electrical insulation for the toroidal field coils of the Compact Ignition Tokamak (CIT) is expected to be exposed to radiation doses on the order of 10 10 rad with ∼90% of the dose from neutrons. The coils are cooled to liquid nitrogen temperature and then heated during the pulse to a peak temperature >300 K. In a program to evaluate the effects of radiation exposure on the insulators, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1, and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to approximately 5 x 10 9 rad and 3 x 10 10 rad with 35 to 40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was done by cycling the shear load for up to 30,000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths on the order of 120 MPa were measured. The behavior of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed nearly identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. 9 refs., 5 figs

  1. What is past is prologue: future directions in tokamak power reactor design research

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    Conceptual tokamak power reactor designs over the last five years have provided us with many fundamental insights regarding tokamaks as fusion reactors. This first generation of studies has helped lay the groundwork upon which to build improvements in reactor design and begin a process of optimization. After reviewing the first generation of studies and the primary conclusions they produced, we discuss four current designs that are representative of present trends in this area of research. In particular, we discuss the trends towards reduced reactor size and higher neutron wall loadings. Moving in this direction requires new approaches to many subsystem designs. We describe new approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets. We close with a discussion of the future role of conceptual reactor design research and the need for close interaction with ongoing experiments in fusion technology

  2. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

    Aymar, R.; Bareyt, B.; Bon Mardion, G.

    1980-10-01

    This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

  3. Design and operation of Alvand II C tokamak

    International Nuclear Information System (INIS)

    Avakian, M.; Azodi, H.; Naraghi, M.; Tabatabaie, H.; Rezvani, B.; Shahnazi, Sh.; Behrozinia, J.; Solaimani, M.

    1984-01-01

    ALVAND IIC is a research tokamak having a circular cross section with major radius of 45.5 cm and minor radius of 12.6 cm, which has been designed and constructed in the plasma physics division of NRC of AEOI. It is used for developing various diagnostics for low beta plasmas and study of the physical characteristics of these plasmas. This paper discusses the design of ALVAND IIC. (Author)

  4. Beam heating requirements for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Bertoncini, P.J.; Brooks, J.N.; Fasolo, J.A.; Stacey, W.M. Jr.

    1976-01-01

    Typical beam heating requirements for effective tokamak experimental power reactor (TEPR) operation have been studied in connection with the Argonne preliminary conceptual TEPR design. For an ignition level plasma (approximately 100 MWt fusion power) for the nominal case envisioned, the neutral beam is only used to heat the plasma to ignition. This typically requires a beam power output of 40 MW at 180 keV for about 3 sec with a total energy of 114 MJ supplied to the plasma. The beam requirements for an ignition device are not very sensitive to changes in wall-sputtered impurity levels or plasma resistivity. For a plasma that must be driven due to poor confinement, the beam must remain on for most of the burn cycle. For representative cases, beam powers of approximately 23 MW are required for a total on-time of 20 to 50 sec. Reqirements on power level, beam energy, on-time, and beam-generation efficiency all represent considerable advances over present technology. For the Argonne TEPR design, a total of 16 to 32 beam injectors is envisioned. For a 40-MW, 180-keV, one-component beam, each injector supplies about 7 to 14 A of neutrals to the plasma. For positive ion sources, about 50 to 100 A of ions are required per injector and some form of particle and/or energy recycling appears to be essential in order to meet the power and efficiency requirements

  5. DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Hseih, S.Y.; Danby, G.; Powell, J.R.

    1979-01-01

    The feasibility of demountable superconducting magnet systems has been examined in a design study of a DEALS [Demountable Externally Anchored Low Stress] TF magnet for an HFITR [High Field Ignition Test Reactor] Tokamak device. All parts of the system appear feasible, including the demountable superconducting joints. Measurements on small scale prototype joints indicate that movable pressure contact joints exhibit acceptable electrical, mechanical, and cryogenic performance. Such joints permit a relatively simple support structure and are readily demountable. Assembly and disassembly sequences are described whereby any failed portion of the magnet, or any part of the reactor inside the TF coils can be removed and replaced if necessary

  6. The Ignition Physics Study Group

    International Nuclear Information System (INIS)

    Sheffield, J.

    1987-01-01

    In the US magnetic fusion program there have been relatively few standing committees of experts, with the mandate to review a particular sub-area on a continuing basis. Generally, ad hoc committees of experts have been assembled to advise on a particular issue. There has been a lack of broad, systematic and continuing review and analysis, combining the wisdom of experts in the field, in support of decision making. The Ignition Physics Study Group (IPSG) provides one forum for the systematic discussion of fusion science, complementing the other exchanges of information, and providing a most important continuity in this critical area. In a similar manner to the European program, this continuity of discussion and the focus provided by a national effort, Compact Ignition Tokamak (CIT), and international effort, Engineering Test Reactor (ETR), are helping to lower those barriers which previously were an impediment to rational debate

  7. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  8. Review Committee report on the conceptual design of the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    1993-04-01

    This report discusses the following topics on the conceptual design of the Tokamak Physics Experiment: Role and mission of TPX; overview of design; physics design assessment; engineering design assessment; evaluation of cost, schedule, and management plans; and, environment safety and health

  9. Conceptual design of a fast-ignition laser fusion reactor FALCON-D

    International Nuclear Information System (INIS)

    Goto, T.; Ogawa, Y.; Okano, K.; Hiwatari, R.; Asaoka, Y.; Someya, Y.; Sunahara, A.; Johzaki, T.

    2008-10-01

    A new conceptual design of the laser fusion power plant FALCON-D (Fast ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast ignition method can achieve the sufficient fusion gain for a commercial operation (∼100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5 - 6 m radius). 1-D/2-D hydrodynamic simulations showed the possibility of the sufficient gain achievement with a 40 MJ target yield. The design feasibility of the compact dry wall chamber and solid breeder blanket system was shown through the thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. A moderate electric output (∼400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall concept not only reduces some difficulties accompanied with a liquid wall but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance time. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R and D issues required for this design are also discussed. (author)

  10. Conceptual design study of the moderate size superconducting spherical tokamak power plant

    Science.gov (United States)

    Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

    2015-06-01

    A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

  11. Dependence of CIT [Compact Ignition Tokamak] PF [poloidal field] coil currents on profile and shape parameters using the Control Matrix

    International Nuclear Information System (INIS)

    Strickler, D.J.; Peng, Y-K.M.; Jardin, S.C.; Pomphrey, N.

    1990-01-01

    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 κ = 2 divertor at beginning of flattop (BOFT) with a minimum stored energy solution for the coil current distribution. The pressure profile function is fixed

  12. Low Convergence path to Fusion I: Ignition physics and high margin design

    Science.gov (United States)

    Molvig, Kim; Schmitt, M. J.; McCall, G. H.; Betti, R.; Foula, D. H.; Campbell, E. M.

    2016-10-01

    A new class of inertial fusion capsules is presented that combines multi-shell targets with laser direct drive at low intensity (280 TW/cm2) to achieve robust ignition. These Revolver targets consist of three concentric metal shells, enclosing a volume of 10s of µg of liquid deuterium-tritium fuel. The inner shell pusher, nominally of gold, is compressed to over 2000 g/cc, effectively trapping the radiation and enabling ignition at low temperature (2.5 keV) and relatively low implosion velocity (20 cm/micro-sec) at a fuel convergence of 9. Ignition is designed to occur well ``upstream'' from stagnation, with implosion velocity at 90% of maximum, so that any deceleration phase mix will occur only after ignition. Mix, in all its non-predictable manifestations, will effect net yield in a Revolver target - but not the achievement of ignition and robust burn. Simplicity of the physics is the dominant principle. There is no high gain requirement. These basic physics elements can be combined into a simple analytic model that generates a complete target design specification given the fuel mass and the kinetic energy needed in the middle (drive) shell (of order 80 kJ). This research supported by the US DOE/NNSA, performed in part at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

  13. Start-up of spherical tokamak without a center solenoid

    International Nuclear Information System (INIS)

    Maekawa, Takashi; Nagata, Masayoshi

    2012-01-01

    For low-aspect tokamak reactors, spherical tokamak reactors, ST-type FESF/CTFs, it is essential to remove or minimize a central solenoid (CS). Even with the minimized CS, non-inductive start up of the plasma current is required. Rapid increase in the spontaneous plasma current at the final stage of current start-up drives ignition. At the initial stage, formation of plasma and magnetic surfaces are required. As non-inductive plasma start-up scenarios, ECH/ECCD, LHCD, HHFW, DC HELICITY injection, plasma merging and NBI have been studied. In the present article, the present status and future prospect of experimental and theoretical works on these subjects. (author)

  14. The Discharge Design of HL-2M with the Tokamak Simulation Code (TSC)

    International Nuclear Information System (INIS)

    Yudong Pan; Jardin, S.C.; Kes, C.

    2007-01-01

    We present results on the discharge design of the HL-2M tokamak, which is to be an upgrade to the existing HL-2A tokamak. We present simulation results for complete 5-sec. discharges, both double null and lower single null, for both ohmic and auxiliary heated discharges. We also discuss the vertical stability properties of the device

  15. National Ignition Facility system design requirements conventional facilities SDR001

    International Nuclear Information System (INIS)

    Hands, J.

    1996-01-01

    This System Design Requirements (SDR) document specifies the functions to be performed and the minimum design requirements for the National Ignition Facility (NIF) site infrastructure and conventional facilities. These consist of the physical site and buildings necessary to house the laser, target chamber, target preparation areas, optics support and ancillary functions

  16. Review of the general atomic experimental fusion power reactor initial conceptual design

    International Nuclear Information System (INIS)

    Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

    1976-01-01

    The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

  17. UWMAK-II: a conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    1975-10-01

    This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high

  18. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    International Nuclear Information System (INIS)

    Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

    2015-01-01

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

  19. Ignition and burn control in tokamak plasmas

    International Nuclear Information System (INIS)

    Borrass, K.; Gruber, O.; Lackner, K.; Minardi, E.; Neuhauser, J.; Wilhelm, R.; Wunderlich, R.; Bromberg, L.; Cohn, D.R.

    1981-01-01

    Different schemes for the control of the thermal instability in an ignited fusion reactor are analysed by zero- and one-dimensional models. Passive stabilization methods considered are ripple-enhanced ion heat conduction, the effect of the major-radius variation of the plasma column in a time-independent vertical field, and the combination of both effects, including the spatial variation of the toroidal-ripple amplitude. Active control methods analysed are high-Q-driven operation and feedback-controlled major-radius variation following different scenarios. One-dimensional analyses taking into account only conductive losses show the existence of a single unstable mode in the energy balance, justifying, under these assumptions, the study of only global control. (author)

  20. Status of Preliminary Design on the Assembly Tools for ITER Tokamak Machine

    International Nuclear Information System (INIS)

    Nam, Kyoung O; Park, Hyun Ki; Kim, Dong Jin; Moon, Jae Hwan; Kim, Byung Seok; Lee, Jae Hyuk; Shaw, Robert

    2012-01-01

    The ITER Tokamak device is principally composed of nine 40 .deg. sectors. Each 40 .deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils (TFC) and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard vacuum vessel thermal shields. Based on the design description document and final report prepared by the ITER organization (IO) and conceptual design, Korea has carried out the preliminary design of these assembly tools. The assembly strategy and relevant tools for the 40 .deg. sector sub-assembly and sector assembly at in-pit should be developed to satisfy the basic assembly requirements of the ITER Tokamak machine. Assembly strategy, preliminary design of the sector sub-assembly and assembly tools are described in this paper

  1. Electrical design of TNS

    International Nuclear Information System (INIS)

    Heck, F.M.; Schultz, J.H.; Smeltzer, G.S.

    1977-01-01

    The electrical design of the ORNL-Westinghouse next step (TNS) fusion reactor was begun in 1976, using a set of ground rules which were based on the overall program objectives. These objectives were to identify the design of reasonably-priced reactors, which would achieve ignition and be technology forcing. The term ''technology forcing'' was understood to mean the desirability of a large number of ignited D-T pulses and the incorporation of superconducting toroidal field (TF) coils, if at all possible. A trade study methodology was developed to compare different machine sizes and TF coil technologies and to aid in the selection of system and subsystem design approaches. The logic which led from the program objectives to the design ground rules and from the ground rules to the circuit selection is described below. The circuit design approaches were generalized and these models were incorporated into a computer program (COAST) which was used to examine the cost of overall tokamak systems as key design parameters were varied

  2. Simulation of the Compact Ignition Tokamak (CIT) conceptual design

    International Nuclear Information System (INIS)

    Carlson, K.E.; Wareing, T.L.

    1988-01-01

    Calculations have been made using the Advanced Thermal Hydraulic Energy Network Analysis (ATHENA) code that simulate the cool down of the cryostat and the performance of the condensing heat exchanger. The purpose of this simulation was to confirm the estimated 30 minute cool down time and to size a condensing heat exchanger for the CIT liquid nitrogen cooling system. This report includes a brief description of the ATHENA code, descriptions of proposed CIT cryostat and condenser designs, and the associated ATHENA models representing these design. This is followed by the ATHENA calculated results and conclusions concerning the results. 6 refs., 17 figs

  3. Influence of interface conditions on laser diode ignition of pyrotechnic mixtures: application to the design of an ignition device

    Energy Technology Data Exchange (ETDEWEB)

    Opdebeck, Frederic; Gillard, Philippe [Laboratoire Energetique Explosions et Structures de l' Universite d' Orleans, 63 boulevard de Lattre de Tassigny, 18020 cedex, Bourges (France); Radenac, D' Erwann [Laboratoire de combustion et de detonique, ENSMA, BP 109, 86960 cedex, Futuroscope (France)

    2003-01-01

    This paper treats of numerical modelling which simulates the laser ignition of pyrotechnic mixtures. The computation zone is divided into two fields. The first is used to take account of the heat loss with the outside. It can represent an optical fibre or a sapphire protective porthole. The second field represents the reactive tablet which absorbs the laser diode's beam. A specific feature of the model is that it incorporates a thermal contact resistance R{sub c} between the two computation fields. Through knowledge of the thermal, optical and kinetic properties, this code makes it possible to compute the ignition conditions. The latter are defined by the energy E{sub 50} and the time t{sub i} of ignition of any pyrotechnic mixture and for various ignition systems.This work was validated in the case of an ignition system consisting of a laser diode with an optical lens re-focussing system. The reactive tablet contains 62% by mass of iron and 38% by mass of KClO{sub 4}. Its porosity is 25.8%. After an evaluation of the laser's coefficient of absorption, the variations of the ignition parameters E{sub 50} and t{sub i} are studied as a function of the thermal contact resistance R{sub c}. Temperature profiles are obtained as a function of time and for various values of the thermal contact resistance R{sub c}. More fundamental observations are made concerning the position of the hot spot corresponding to priming. From this study, which concerns the heat exchange between the two media, several practical conclusions are given concerning the design of an ignition device. By evaluation of the thermal contact resistance R{sub c}, comparison with test results becomes possible and the results of the computations are in reasonable agreement with the test measurements. (authors)

  4. Continuous tokamak operation with an internal transformer

    International Nuclear Information System (INIS)

    Singer, C.E.; Mikkelsen, D.R.

    1982-10-01

    A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia timescales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors

  5. Transport simulations of ohmic ignition experiment: IGNITEX

    International Nuclear Information System (INIS)

    Uckan, N.A.; Howe, H.C.

    1987-01-01

    The IGNITEX device, proposed by Rosenbluth et al., is a compact, super-high-field, high-current, copper-coil tokamak envisioned to reach ignition with ohmic (OH) heating alone. Several simulations of IGNITEX were made with a 0-D global model and with the 1-D PROCTR transport code. It is shown that OH ignition is a sensitive function of the assumptions about density profile, wall reflectivity of synchrotron radiation, impurity radiation, plasma edge conditions, and additional anomalous losses. In IGNITEX, OH ignition is accessible with nearly all scalings based on favorable OH confinement (such as neo-Alcator). Also, OH ignition appears to be accessible for most (not all) L-mode scalings (such as Kaye-Goldston), provided that the density profile is not too broad (parabolic or more peaked profiles are needed), Z/sub eff/ is not too large (≤2), and anomalous radiation and alpha losses and/or other enhanced transport losses (/eta//sub i/ modes, edge convective energy losses, etc.) are not present. In IGNITEX, because the figure-of-merit parameters (aB 0 2 /q* /approximately/ IB 0 , etc.) are large, ignition can be accessed (either with OH heating alone or with the aid of a small amount of auxiliary power) at relatively low beta, far from stability limits. Once the plasma is ignited, thermal runaway is prevented naturally by a combination of increased synchrotron radiation, burnout of the fuel in the plasma core and replacement by thermal alphas, and the reduction in the thermal plasma confinement assumed in L-mode-like scalings. 12 refs., 5 figs., 1 tab

  6. Transport simulations of ohmic ignition experiment: IGNITEX

    International Nuclear Information System (INIS)

    Uckan, N.A.; Howe, H.C.

    1987-12-01

    The IGNITEX device, proposed by Rosenbluth et al., is a compact, super-high-field, high-current, copper-coil tokamak envisioned to reach ignition with ohmic (OH) heating alone. Several simulations of IGNITEX were made with a 0-D global model and with the 1-D PROCTR transport code. It is shown that OH ignition is a sensitive function of the assumptions about density profile, wall reflectivity of synchrotron radiation, impurity radiation, plasma edge conditions, and additional anomalous losses. In IGNITEX, OH ignition is accessible with nearly all scalings based on favorable OH confinement (such as neo-Alcator). Also, OH ignition appears to be accessible for most (not all) L-mode scalings (such as Kaye-Goldston), provided that the density profile is not too broad (parabolic or more peaked profiles are needed), Z/sub eff/ is not too large, and anomalous radiation and alpha losses and/or other enhanced transport losses (eta/sub i/ modes, edge convective energy losses, etc.) are not present. In IGNITEX, because the figure-of-merit parameters are large, ignition can be accessed (either with OH heating alone or with the aid of a small amount of auxiliary power) at relatively low beta, far from stability limits. Once the plasma is ignited, thermal runaway is prevented naturally by a combination of increased synchrotron radiation, burnout of the fuel in the plasma core and replacement by thermal alphas, and the reduction in the thermal plasma confinement assumed in L-mode-like scalings. 12 refs., 5 figs., 1 tab

  7. Conceptual design of the National Ignition Facility

    International Nuclear Information System (INIS)

    Paisner, J.A.; Kumpan, S.A.; Lowdermilk, W.H.; Boyes, J.D.; Sorem, M.

    1995-01-01

    DOE commissioned a Conceptual Design Report (CDR) for the National Ignition Facility (NIF) in January 1993 as part of a Key Decision Zero (KDO), justification of Mission Need. Motivated by the progress to date by the Inertial Confinement Fusion (ICF) program in meeting the Nova Technical Contract goals established by the National Academy of Sciences in 1989, the Secretary requested a design using a solid-state laser driver operating at the third harmonic (0.35 μm) of neodymium (Nd) glass. The participating ICF laboratories signed a Memorandum of Agreement in August 1993, and established a Project organization, including a technical team from the Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Sandia National Laboratories (SNL), and the Laboratory for Laser Energetics at the University of Rochester. Since then, we completed the NIF conceptual design, based on standard construction at a generic DOE Defense Program's site, and issued a 7,000-page, 27-volume CDR in May 1994.2 Over the course of the conceptual design study, several other key documents were generated, including a Facilities Requirements Document, a Conceptual Design Scope and Plan, a Target Physics Design Document, a Laser Design Cost Basis Document, a Functional Requirements Document, an Experimental Plan for Indirect Drive Ignition, and a Preliminary Hazards Analysis (PHA) Document. DOE used the PHA to categorize the NIF as a low-hazard, non-nuclear facility. On October 21, 1994 the Secretary of Energy issued a Key Decision One (KD1) for the NIF, which approved the Project and authorized DOE to request Office of Management and Budget-approval for congressional line-item FY 1996 NIF funding for preliminary engineering design and for National Environmental Policy Act activities. In addition, the Secretary declared Livermore as the preferred site for constructing the NIF. The Project will cost approximately $1.1 billion and will be completed at the end of FY 2002

  8. Conceptual design of economic compact reversed shear Tokamak (CRST)

    International Nuclear Information System (INIS)

    Okano, Kunihiko

    1998-01-01

    Two indices of performance for economic analysis of Tokamak are defined as toroidal β value: β t (%)=(plasma pressure)/(pressure of magnetic field) and Troyon coefficient β N . The pressure of magnetic field is defined as β t 2 /2μ 0 (Bt: strength of toroidal magnetic field and μ 0 : permeability). β N is determined in order to make possible compare β t between other devices. To increase β N is very important on the economic viewpoint. ITER is designed as 2.2 β N , 1 MW/m 2 average neutron wall load, 8.14 m large radius and 2.8 m small radius, but the above values of CRST are 5.5, 4.5 MW/m 2 , 5.4 m and 1.59 m, respectively. Development of industrial and physical technologies makes possible to minimize economic Tokamak. After ITER, we expect that economic fusion reactor is obtained by minimization. CRST satisfies the conditions of economic fusion reactor conduced by the economic analysis. CRST is designed as 5.4 m main radius and 116x10 4 kW electric output. Fundamental physics and technologies, conceptual and industrial design of CRST are explained. (S.Y.)

  9. Tokamak reactor designs as a function of aspect ratio

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Stambaugh, R.D.

    2000-01-01

    This paper assesses the technical and economic potential of tokamak power plants which utilize superconducting coil (SC) or normal conducting coil (NC) designs as a function of aspect ratio (A). Based on the results from plasma equilibrium calculations, the key physics design parameters of β N , β p , β T , and κ were fitted to parametric equations covering A in the range of 1.2-6. By using ARIES-RS and ARIES-ST as reference design points, a fusion reactor system code was used to project the performance and cost of electricity (COE) of SC and NC reactor designs over the same range of A. The principle difference between the SC and the NC designs are the inboard standoff distance between the coil and the inboard first wall, and the maximum central column current density used for respective coil types. Results show that at an output power of 2 GWe both NC and SC designs can project COE in the respectable range of 62-65 mill/kW h at gross thermal efficiency of 46%, with neutron wall loading (Γ n ) ∼7 MW/m 2 . More importantly, we have learned that based on the present knowledge of equilibrium physics and fusion power core components and system design we can project the performance and COE of reactor designs at least for the purpose of comparative assessment. Tokamak design points can then be selected and optimized for testing or commercial devices as a function of output power, A and Γ n for both SC and NC design options

  10. Indirect-drive noncryogenic double-shell ignition targets for the National Ignition Facility: Design and analysis

    International Nuclear Information System (INIS)

    Amendt, Peter; Colvin, J.D.; Tipton, R.E.; Hinkel, D.E.; Edwards, M.J.; Landen, O.L.; Ramshaw, J.D.; Suter, L.J.; Varnum, W.S.; Watt, R.G.

    2002-01-01

    Analysis and design of indirect-drive National Ignition Facility double-shell targets with hohlraum temperatures of 200 eV and 250 eV are presented. The analysis of these targets includes the assessment of two-dimensional radiation asymmetry and nonlinear mix. Two-dimensional integrated hohlraum simulations indicate that the x-ray illumination can be adjusted to provide adequate symmetry control in hohlraums specially designed to have high laser-coupling efficiency [Suter et al., Phys. Plasmas 7, 2092 (2000)]. These simulations also reveal the need to diagnose and control localized 10-15 keV x-ray emission from the high-Z hohlraum wall because of strong absorption by the high-Z inner shell. Preliminary estimates of the degree of laser backscatter from an assortment of laser-plasma interactions suggest comparatively benign hohlraum conditions. The application of a variety of nonlinear mix models and phenomenological tools, including buoyancy-drag models, multimode simulations and fall-line optimization, indicates a possibility of achieving ignition, i.e., fusion yields greater than 1 MJ. Planned experiments on the Omega laser will test current understanding of high-energy radiation flux asymmetry and mix-induced yield degradation in double-shell targets

  11. Robustness studies of ignition targets for the National Ignition Facility in two dimensions

    International Nuclear Information System (INIS)

    Clark, Daniel S.; Haan, Steven W.; Salmonson, Jay D.

    2008-01-01

    Inertial confinement fusion capsules are critically dependent on the integrity of their hot spots to ignite. At the time of ignition, only a certain fractional perturbation of the nominally spherical hot spot boundary can be tolerated and the capsule still achieve ignition. The degree to which the expected hot spot perturbation in any given capsule design is less than this maximum tolerable perturbation is a measure of the ignition margin or robustness of that design. Moreover, since there will inevitably be uncertainties in the initial character and implosion dynamics of any given capsule, all of which can contribute to the eventual hot spot perturbation, quantifying the robustness of that capsule against a range of parameter variations is an important consideration in the capsule design. Here, the robustness of the 300 eV indirect drive target design for the National Ignition Facility [Lindl et al., Phys. Plasmas 11, 339 (2004)] is studied in the parameter space of inner ice roughness, implosion velocity, and capsule scale. A suite of 2000 two-dimensional simulations, run with the radiation hydrodynamics code LASNEX, is used as the data base for the study. For each scale, an ignition region in the two remaining variables is identified and the ignition cliff is mapped. In accordance with the theoretical arguments of Levedahl and Lindl [Nucl. Fusion 37, 165 (1997)] and Kishony and Shvarts [Phys. Plasmas 8, 4925 (2001)], the location of this cliff is fitted to a power law of the capsule implosion velocity and scale. It is found that the cliff can be quite well represented in this power law form, and, using this scaling law, an assessment of the overall (one- and two-dimensional) ignition margin of the design can be made. The effect on the ignition margin of an increase or decrease in the density of the target fill gas is also assessed

  12. Conceptual design for the ZEPHYR neutral-beam injection system

    International Nuclear Information System (INIS)

    Cooper, W.S.; Elischer, V.P.; Goldberg, D.A.; Hopkins, D.B.; Jacobson, V.L.; Lou, K.H.; Tanabe, J.T.

    1981-03-01

    In June 1980, the Lawrence Berkeley Laboratory began a conceptual design study for a neutral beam injection system for the ZEPHYR ignition tokamak proposed by the Max-Planck-Institut fur Plasmaphysik in Garching, Germany. The ZEPHYR project was cancelled, and the LBL design effort concluded prematurely in January 1981. This report describes the conceptual design as it existed at that time, and gives brief consideration to a schedule, but does not deal with costs

  13. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  14. Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

    International Nuclear Information System (INIS)

    Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

    1988-01-01

    The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor plasma (Tokamak Ignition/Burn Experimental Reactor) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-dimensional transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability

  15. Design of plasma facing components for the SST-1 tokamak

    International Nuclear Information System (INIS)

    Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

    2000-01-01

    Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

  16. The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Miller, R.L.

    1989-01-01

    The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

  17. Perfecting of shielding calculation technique against the gamma rays arising from a Tokamak with the TFR experience. Application to the conceptual design Tokamak TORE 2 SUPRA

    International Nuclear Information System (INIS)

    Diop, Cheikh M'Backe.

    1980-09-01

    The conception of the necessary shielding around a conceptual design Tokamak requires to execute an estimated calculation of the doses due to the different radiation sources arising from the machine: the thermonuclear neutron source and the gamma ray source emitted during the interaction of the runaway electrons with the diaphragm. In this study, we propose a theorical method to calculate this gamma source. We tackle also the shielding problem of the conceptual design Tokamak: TORE 2 SUPRA [fr

  18. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

  19. Consideration of neutral beam prompt loss in the design of a tokamak helicon antenna

    International Nuclear Information System (INIS)

    Pace, D.C.; Van Zeeland, M.A.; Fishler, B.; Murphy, C.

    2016-01-01

    Highlights: • Neutral beam prompt losses place appreciable power on an in-vessel tokamak antenna. • Simulations predict prompt loss power and inform protective tile design. • Experiments confirm the validity of the prompt loss simulations. - Abstract: Neutral beam prompt losses (injected neutrals that ionize such that their first poloidal transit intersects with the wall) can put appreciable power on the outer wall of tokamaks, and this power may damage the wall or other internal components. These prompt losses are simulated including a protruding helicon antenna installation in the DIII-D tokamak and it is determined that 160 kW of power will impact the antenna during the injection of a particular neutral beam. Protective graphite tiles are designed in response to this modeling and the wall shape of the installed antenna is precisely measured to improve the accuracy of these calculations. Initial experiments confirm that the antenna component temperature increases according to the amount of neutral beam energy injected into the plasma. In this case, only injection of beams that are aimed counter to the plasma current produce an appreciable power load on the outer wall, suggesting that the effect is of little concern for tokamaks featuring only co-current neutral beam injection. Incorporating neutral beam prompt loss considerations into the design of this in-vessel component serves to ensure that adequate protection or cooling is provided.

  20. Consideration of neutral beam prompt loss in the design of a tokamak helicon antenna

    Energy Technology Data Exchange (ETDEWEB)

    Pace, D.C., E-mail: pacedc@fusion.gat.com; Van Zeeland, M.A.; Fishler, B.; Murphy, C.

    2016-11-15

    Highlights: • Neutral beam prompt losses place appreciable power on an in-vessel tokamak antenna. • Simulations predict prompt loss power and inform protective tile design. • Experiments confirm the validity of the prompt loss simulations. - Abstract: Neutral beam prompt losses (injected neutrals that ionize such that their first poloidal transit intersects with the wall) can put appreciable power on the outer wall of tokamaks, and this power may damage the wall or other internal components. These prompt losses are simulated including a protruding helicon antenna installation in the DIII-D tokamak and it is determined that 160 kW of power will impact the antenna during the injection of a particular neutral beam. Protective graphite tiles are designed in response to this modeling and the wall shape of the installed antenna is precisely measured to improve the accuracy of these calculations. Initial experiments confirm that the antenna component temperature increases according to the amount of neutral beam energy injected into the plasma. In this case, only injection of beams that are aimed counter to the plasma current produce an appreciable power load on the outer wall, suggesting that the effect is of little concern for tokamaks featuring only co-current neutral beam injection. Incorporating neutral beam prompt loss considerations into the design of this in-vessel component serves to ensure that adequate protection or cooling is provided.

  1. Ion cyclotron system design for KSTAR tokamak

    International Nuclear Information System (INIS)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H.

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

  2. Ion cyclotron system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

    1998-05-01

    The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

  3. Wildfire ignition resistant home design(WIRHD) program: Full-scale testing and demonstration final report.

    Energy Technology Data Exchange (ETDEWEB)

    Quarles, Stephen, L.; Sindelar, Melissa

    2011-12-13

    The primary goal of the Wildfire ignition resistant home design(WIRHD) program was to develop a home evaluation tool that could assess the ignition potential of a structure subjected to wildfire exposures. This report describes the tests that were conducted, summarizes the results, and discusses the implications of these results with regard to the vulnerabilities to homes and buildings.

  4. Impact of major design parameters on the economics of Tokamak power plants

    International Nuclear Information System (INIS)

    Abdou, M.A.; Ehst, D.; Maroni, V.; Stacey, W.M. Jr.

    1977-11-01

    A parametric systems studies program is now in an active stage at Argonne National Laboratory. This paper presents a summary of results from this systems analysis effort. The impact of major design parameters on the economics of tokamak power plants is examined. The major parameters considered are: (1) the plant power rating; (2) toroidal-field strength; (3) plasma β/sub t/; (4) aspect ratio; (5) plasma elongation; (6) inner blanket/shield thickness; and (7) neutron wall load. The performance characteristics and economics of tokamak power plants are also compared for two structural materials

  5. Concept design on RH maintenance of CFETR Tokamak reactor

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Songtao; Wan, Yuanxi; Li, Jiangang; Ye, Minyou; Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua

    2014-01-01

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed

  6. The engineering design of the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1994-01-01

    A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design. If adequately funded the construction project should be completed in the year 2000

  7. Status of the ITER tokamak nuclear shielding and radiological protection design

    Energy Technology Data Exchange (ETDEWEB)

    Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

    2016-11-01

    Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

  8. Trade studies of plasma elongation for next-step tokamaks

    International Nuclear Information System (INIS)

    Galambos, J.D.; Strickler, D.J.; Peng, Y.K.M.; Reid, R.L.

    1988-09-01

    The effect of elongation on minimum-cost devices is investigated for elongations ranging from 2 to 3. The analysis, carried out with the TETRA tokamak systems code, includes the effects of elongation on both physics (plasma beta limit) and engineering (poloidal field coil currents) issues. When ignition is required, the minimum cost occurs for elongations from 2.3 to 2.9, depending on the plasma energy confinement scaling used. Scalings that include favorable plasma current dependence and/or degradation with fusion power tend to have minimum cost at higher elongation (2.5-2.9); scalings that depend primarily on size result in lower elongation (/approximately/2.3) for minimum cost. For design concepts that include steady-state current-driven operation, minimum cost occurs at an elongation of 2.3. 12 refs., 13 figs

  9. Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak

    NARCIS (Netherlands)

    Hennen, B.A.; Westerhof, E.; Nuij, Pwjm; M.R. de Baar,; Steinbuch, M.

    2012-01-01

    Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron

  10. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  11. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  12. Re-Imagining Specialized STEM Academies: Igniting and Nurturing "Decidedly Different Minds", by Design

    Science.gov (United States)

    Marshall, Stephanie Pace

    2010-01-01

    This article offers a personal vision and conceptual design for reimagining specialized science, technology, engineering, and mathematics (STEM) academies designed to nurture "decidedly different" STEM minds and ignite a new generation of global STEM talent, innovation, and entrepreneurial leadership. This design enables students to engage…

  13. System design of toroidal field power supply of CDD tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zheng Zhi

    1996-12-01

    This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs.

  14. System design of toroidal field power supply of CDD tokamak

    International Nuclear Information System (INIS)

    Liu, Zheng Zhi.

    1996-12-01

    This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs

  15. Workshop on High Power ICH Antenna Designs for High Density Tokamaks

    Science.gov (United States)

    Aamodt, R. E.

    1990-02-01

    A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

  16. On the density limit of Tokamaks

    International Nuclear Information System (INIS)

    Lehnert, B.

    1982-12-01

    Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

  17. Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Y.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Cheng, Z. F.; Hou, S. Y.; Cheng, C.; Li, Z.; Wang, J. R.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-11-15

    The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 − 7 transition) line seems to be the best choice for ion temperature and plasma rotation measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.

  18. Recent progress in ignition fusion research on the National Ignition Facility

    International Nuclear Information System (INIS)

    Leeper, Ramon J.

    2011-01-01

    This paper will review the ignition fusion research program that is currently being carried out on the National Ignition Facility (NIF) located at Lawrence Livermore National Laboratory. This work is being conducted under the auspices of the National Ignition Campaign (NIC) that is a broad collaboration of national laboratories and universities that together have developed a detailed research plan whose goal is ignition in the laboratory. The paper will begin with a description of the NIF facility and associated experimental facilities. The paper will then focus on the ignition target and hohlraum designs that will be tested in the first ignition attempts on NIF. The next topic to be introduced will be a description of the diagnostic suite that has been developed for the initial ignition experiments on NIF. The paper will then describe the experimental results that were obtained in experiments conducted during the fall of 2009 on NIF. Finally, the paper will end with a description of the detailed experimental plans that have been developed for the first ignition campaign that will begin later this year. (author)

  19. Tokamak Physics Experiment (TPX) power supply design and development

    International Nuclear Information System (INIS)

    Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

    1995-01-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes

  20. Robust control design for the plasma horizontal position control on J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Yu, W.Z.; Chen, Z.P.; Zhuang, G.; Wang, Z.J.

    2013-01-01

    It is extremely important for tokamak to control the plasma position during routine discharge. However, the model of plasma in tokamak usually contains much of the uncertainty, such as structured uncertainties and unmodeled dynamics. Compared with the traditional PID control approach, robust control theory is more suitable to handle this problem. In the paper, we propose a H ∞ robust control scheme to control the horizontal position of plasma during the flat-top phase of discharge on Joint Texas Experimental Tokamak (J-TEXT) tokamak. First, the model of our plant for plasma horizontal position control is obtained from the position equilibrium equations. Then the H ∞ robust control framework is used to synthesize the controller. Based on this, an H ∞ controller is designed to minimize the regulation/tracking error. Finally, a comparison study is conducted between the optimized H ∞ robust controller and the traditional PID controller in simulations. The simulation results of the H ∞ robust controller show a significant improvement of the performance with respect to those obtained with traditional PID controller, which is currently used on our machine

  1. High-field, high-density tokamak power reactor

    International Nuclear Information System (INIS)

    Cohn, D.R.; Cook, D.L.; Hay, R.D.; Kaplan, D.; Kreischer, K.; Lidskii, L.M.; Stephany, W.; Williams, J.E.C.; Jassby, D.L.; Okabayashi, M.

    1977-11-01

    A conceptual design of a compact (R 0 = 6.0 m) high power density (average P/sub f/ = 7.7 MW/m 3 ) tokamak demonstration power reactor has been developed. High magnetic field (B/sub t/ = 7.4 T) and moderate elongation (b/a = 1.6) permit operation at the high density (n(0) approximately 5 x 10 14 cm -3 ) needed for ignition in a relatively small plasma, with a spatially-averaged toroidal beta of only 4%. A unique design for the Nb 3 Sn toroidal-field magnet system reduces the stress in the high-field trunk region, and allows modularization for simpler disassembly. The modest value of toroidal beta permits a simple, modularized plasma-shaping coil system, located inside the TF coil trunk. Heating of the dense central plasma is attained by the use of ripple-assisted injection of 120-keV D 0 beams. The ripple-coil system also affords dynamic control of the plasma temperature during the burn period. A FLIBE-lithium blanket is designed especially for high-power-density operation in a high-field environment, and gives an overall tritium breeding ratio of 1.05 in the slowly pumped lithium

  2. Static or feedback stabilization of the burn in a Tokamak

    International Nuclear Information System (INIS)

    Minardi, E.

    1980-02-01

    The control of the burn in an ignited Tokamak using a space and time dependent external vertical magnetic field is discussed. It is shown that a static field, suitably shaped in space, is able to stabilize the burn for a certain range of the plasma parameters of physical interest. An oscillating magnetic field with constant frequency and amplitude fixed by the initial plasma parameters stabilizes the burn in all situations. (orig.)

  3. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  4. The Tokamak Fusion Test Reactor D-T modifications and operations

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  5. Workshop on high power ICH antenna designs for high density tokamaks

    International Nuclear Information System (INIS)

    Aamodt, R.E.

    1990-01-01

    A workshop in high power ICH antenna designs for high density tokamaks was held in Boulder, Colorado on January 31 through February 2, 1990. The purposes of the workshop were to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of rf auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made

  6. An interim report on the materials and selection criteria analysis for the Compact Ignition Tokamak Toroidal Field Coil Turn-to-Turn Insulation System

    International Nuclear Information System (INIS)

    Campbell, V.W.; Dooley, J.B.; Hubrig, J.G.; Janke, C.J.; McManamy, T.J.; Welch, D.E.

    1990-01-01

    Design criteria for the Compact Ignition Tokamak, Toroidal-Field (TF) Coil, Turn-to-Turn Insulation System require an insulation sheet and bonding system that will survive cryogenic cycling in a radiation environment and maintain structural integrity during exposure to the significant compressive and shear loads associated with each operating cycle. For thermosetting resin systems, a complex interactive dependency exists between optimum peak value, in-service property performance capabilities of candidate generic materials; key handling and processing parameters required to achieve their optimum in-service property performance as an insulation system; and suitability of their handling and processing parameters as a function of design configuration and assembly methodology. This dependency is assessed in a weighted study matrix in which two principal programmatic approaches for the development of the TF Coil Subassembly Insulation System have been identified. From this matrix study, two viable approaches to the fabrication of the insulation sheet were identified: use of a press-formed sheet bonded in place with epoxy for mechanical bonding and tolerance take-up and formation of the insulation sheet by placement of dry cloth and subsequent vacuum pressure impregnation. Laboratory testing was conducted to screen a number of combinations of resins and hardeners on a generic basis. These combinations were chosen for their performance in similar applications. Specimens were tested to screen viscosity, thermal-shock tolerance, and cryogenic tolerance. Cryogenic shock and cryogenic temperature proved to be extremely lethal to many combinations of resin, hardener, and cure. Two combinations survived: a heavily flexibilized bisphenol A resin with a flexibilized amine hardener and a bisphenol A resin with cycloaliphatic amine hardener. 7 refs., 12 figs., 6 tabs

  7. Large coil program support structure conceptual design

    International Nuclear Information System (INIS)

    Litherland, P.S.

    1977-01-01

    The purpose of the Large Coil Program (LCP) is to perform tests on both pool boiling and force cooled superconducting toroidal field coils. The tests will attempt to approximate conditions anticipated in an ignition tokamak. The test requirements resulted in a coil support design which accommodates up to six (6) test coils and is mounted to a structure capable of resisting coil interactions. The steps leading to the present LCP coil support structure design, details on selected structural components, and the basic assembly sequence are discussed

  8. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  9. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  10. Tokamak blanket design study: FY 78 summary report

    International Nuclear Information System (INIS)

    1979-06-01

    A tokamak blanket cylindrical module concept was designed, developed, and analyzed after review of several existing generic concepts. The design is based on use of state-of-the-art structural materials (20% cold worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders and features direct wall cooling by flowing helium between the outer (first wall) cylinder and the inner lithium containing cylinder. Each cylinder is capable of withstanding full coolant pressure for enhanced reliability. Results show that stainless steel is a viable material for a first wall subjected to 4 MW/m 2 neutron and 1 MW/m 2 particle heat flux. A lifetime analysis showed that the first wall design meets the goal of operating at 20 minute cycles with 95% duty for 10 5 cycles. The design is attractive for further development, and additional work and supporting experiments are identified to reduce analytical uncertainties and enhance the design reliability

  11. Design point selection for an ignited ITER

    International Nuclear Information System (INIS)

    Spears, W.R.; Mizoguchi, T.; Perkins, J.; Putvinskij, S.; Flanagan, C.

    1989-01-01

    The parameter choice for the ITER design is dominated by the need to achieve ignition. Different energy confinement time scalings extrapolate to different device parameters, and the final design chosen must have a reasonable chance of success under the presently favoured scalings. This study investigates the device performance over the plasma current - aspect ratio design space, using sets of simplified equations and more detailed parametric analysis codes, with a view to clarifying the acceptable parameter region. Alterations to the initial assumptions of plasma elongation, plasma safety factor and peak toroidal field, and the effect of limiting the wall load, are also investigated. The study concludes that a device with aspect ratio around 3 and plasma current about 20 MA would be an appropriate choice. Under the assumptions of the study such a device would have a major radius of about 5.8m and would require the attainment of enhancement factors around 1.5 over L-mode scaling. (author). 6 refs, 11 figs, 3 tabs

  12. Internal combustion engines a detailed introduction to the thermodynamics of spark and compression ignition engines, their design and development

    CERN Document Server

    Benson, Rowland S

    1979-01-01

    Internal Combustion of Engines: A Detailed Introduction to the Thermodynamics of Spark and Compression Ignition Engines, Their Design and Development focuses on the design, development, and operations of spark and compression ignition engines. The book first describes internal combustion engines, including rotary, compression, and indirect or spark ignition engines. The publication then discusses basic thermodynamics and gas dynamics. Topics include first and second laws of thermodynamics; internal energy and enthalpy diagrams; gas mixtures and homocentric flow; and state equation. The text ta

  13. Tritium pellet injector design for tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.

    1985-01-01

    A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper

  14. Physics design and experimental study of tokamak divertor

    International Nuclear Information System (INIS)

    Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

    2007-06-01

    The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

  15. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  16. A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks

    International Nuclear Information System (INIS)

    Woolley, R.D.

    2009-01-01

    A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress

  17. A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks

    International Nuclear Information System (INIS)

    Woolley, Robert D.

    2009-01-01

    A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress.

  18. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    International Nuclear Information System (INIS)

    1995-01-01

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided

  19. Tokamak Physics EXperiment (TPX): Toroidal field magnet design, development and manufacture. SDRL 15, System design description. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-22

    This System Design Description, prepared in accordance with the TPX Project Management Plan provides a summary or TF Magnet System design features at the conclusion of Phase I, Preliminary Design and Manufacturing Research. The document includes the analytical and experimental bases for the design, and plans for implementation in final design, manufacturing, test, and magnet integration into the tokamak. Requirements for operation and maintenance are outlined, and references to sources of additional information are provided.

  20. Recent developments in the role of atomic processes in future tokamaks

    International Nuclear Information System (INIS)

    Post, D.

    1996-01-01

    Since the beginning of magnetic fusion research, reducing the impurity level in experiments has been strongly correlated with successful achievement of high performance plasmas. One of the most important examples of this was the recognition that the use of tungsten as a plasma facing material and the associated high radiative losses were responsible for the poor performance of the ORMAK and PLT tokamaks. Tungsten was replaced with graphite and the central plasma temperature in PLT increased a factor of ten. The magnetic fusion program is now planning on constructing an ignited fusion experiment. One of the major design issues is the reduction of the peak heat loads on the plasma facing components. It appears that the carefully controlled introduction of impurities can lead to a solution of the problem. copyright 1996 American Institute of Physics

  1. Neutral beam injection system design for KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Choi, B.H.; Lee, K.W.; Chung, K.S.; Oh, B.H.; Cho, Y.S.; Bae, Y.D.; Han, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    The NBI system for KSTAR (Korean Superconducting Tokamak Advanced Research) has been designed based on conventional positive ion beam technology. One beam line consists of three ion sources, three neutralizers, one bending magnet, and one drift tube. This system will deliver 8 MW deuterium beam to KSTAR plasma in normal operation to support the advanced experiments on heating, current drive and profile control. The key technical issues in this design were high power ion source(120 kV, 65 A), long pulse operation (300 seconds; world record is 30 sec), and beam rotation from vertical to horizontal direction. The suggested important R and D points on ion source and beam line components are also included. (author). 7 refs., 27 figs., 1 tab.

  2. Spatial filter lens design for the main laser of the National Ignition Facility

    International Nuclear Information System (INIS)

    Korniski, R.J.

    1998-01-01

    The National Ignition Facility (NIF), being designed and constructed at Lawrence Livermore National Laboratory (LLNL), comprises 192 laser beams The lasing medium is neodymium in phosphate glass with a fundamental frequency (1ω) of 1 053microm Sum frequency generation in a pair of conversion crystals (KDP/KD*P) will produce 1 8 megajoules of the third harmonic light (3ω or λ=351microm) at the target The purpose of this paper is to provide the lens design community with the current lens design details of the large optics in the Main Laser This paper describes the lens design configuration and design considerations of the Main Laser The Main Laser is 123 meters long and includes two spatial filters one 13 5 meters and one 60 meters These spatial filters perform crucial beam filtering and relaying functions We shall describe the significant lens design aspects of these spatial filter lenses which allow them to successfully deliver the appropriate beam characteristic onto the target For an overview of NIF please see ''Optical system design of the National Ignition Facility,'' by R Edward English. et al also found in this volume

  3. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  4. Conceptual design of HL-2M tokamak control system

    International Nuclear Information System (INIS)

    Xia Fan; Chen Liaoyuan; Song Xianming; Zhang Jinhua; Lou Cuiwen; Pan Yudong

    2009-01-01

    The static architecture, dynamic behavior, control theory and simulation of HL-2M tokamak control system are described. The real-time network will be build for the communication of real-time control among its subsystems and universal timing system will be build to guarantee the synchronization among the subsystems. The duty to achieve preprogrammed parameters is carried out by plasma discharge control. In order to reduce the damage made by discharge exception, the error-handing mechanism of supervision system is considered. The controllers of magnetic control system are designed to control the current, shape and position of plasma and simulation system is designed for testing the controllers. (authors)

  5. Safety in the ARIES Tokamak Design Study

    International Nuclear Information System (INIS)

    Herring, J.S.; Wong, C.P.-C.; Cheng, E.T.; Grotz, S.

    1989-01-01

    Safety is one of the primary goals of the ARIES Tokamak Design Study. Public safety goals are the achievement passive safety which is demonstrable in tests that could precede operation and the assurance that releases from accidents be passively limited such that no evacuation plan in necessary. Strategies for safety of the plant investment are factory fabrication, short construction times and a design such that no off-normal operational transient results in damage which could not be repaired in routine maintenance. ARIES-I, the first of three 'visions' of potential tokamak reactors, will use He at 5 MPa as a blanket coolant and SiC/composite ceramic for the first wall and blanket materials. Both the coolant and the structural material were chosen for their low activation, both in the short term after accidents and for long term waste management. The breeder, Li 4 SiO 4 , was also chosen for low activation. Contemporary plasma physics and aggressive technology are used in ARIES-I, which results in very high toroidal fields (24 T maximum at the coil). The stored TF energy will be about 130 GJ. A central concern is the safe discharge of this stored energy under electrical fault conditions and prevention of a failure in the magnet set from propagating into systems containing radioactive inventories. The TF coil system consists of 16 coils, each containing two separate windings powered by two independent power supplies. Arcs and shorts between the two power supply systems and across individual windings have been modeled. In addition, delay or failure in circuit breaker opening has been modeled. The safety impacts of LOCA, LOFA and disruptive events have also been evaluated. 8 refs., 4 figs., 7 tabs

  6. Design of an ignition target for the laser megajoule, mitigating parametric instabilities

    Science.gov (United States)

    Laffite, S.; Loiseau, P.

    2010-10-01

    Laser plasma interaction (LPI) is a critical issue in ignition target design. Based on both scaling laws and two-dimensional calculations, this article describes how we can constrain a laser megajoule (LMJ) [J. Ebrardt and J. M. Chaput, J. Phys.: Conf. Ser. 112, 032005 (2008)] target design by mitigating LPI. An ignition indirect drive target has been designed for the 2/3 LMJ step. It requires 0.9 MJ and 260 TW of laser energy and power, to achieve a temperature of 300 eV in a rugby-shaped Hohlraum and give a yield of about 20 MJ. The study focuses on the analysis of linear gain for stimulated Raman and Brillouin scatterings. Enlarging the focal spot is an obvious way to reduce linear gains. We show that this reduction is nonlinear with the focal spot size. For relatively small focal spot area, linear gains are significantly reduced by enlarging the focal spot. However, there is no benefit in too large focal spots because of necessary larger laser entrance holes, which require more laser energy. Furthermore, this leads to the existence, for a given design, of a minimum value for linear gains for which we cannot go below.

  7. Design of an ignition target for the laser megajoule, mitigating parametric instabilities

    International Nuclear Information System (INIS)

    Laffite, S.; Loiseau, P.

    2010-01-01

    Laser plasma interaction (LPI) is a critical issue in ignition target design. Based on both scaling laws and two-dimensional calculations, this article describes how we can constrain a laser megajoule (LMJ) [J. Ebrardt and J. M. Chaput, J. Phys.: Conf. Ser. 112, 032005 (2008)] target design by mitigating LPI. An ignition indirect drive target has been designed for the 2/3 LMJ step. It requires 0.9 MJ and 260 TW of laser energy and power, to achieve a temperature of 300 eV in a rugby-shaped Hohlraum and give a yield of about 20 MJ. The study focuses on the analysis of linear gain for stimulated Raman and Brillouin scatterings. Enlarging the focal spot is an obvious way to reduce linear gains. We show that this reduction is nonlinear with the focal spot size. For relatively small focal spot area, linear gains are significantly reduced by enlarging the focal spot. However, there is no benefit in too large focal spots because of necessary larger laser entrance holes, which require more laser energy. Furthermore, this leads to the existence, for a given design, of a minimum value for linear gains for which we cannot go below.

  8. Heavy ion fusion targets; issues for fast ignition

    International Nuclear Information System (INIS)

    Bangerter, Roger O.

    2014-01-01

    During the last 36 years researchers have suggested and evaluated a large number of target designs for heavy ion inertial fusion. The different target designs can be classified according to their mode of ignition, their method of implosion, and their size. Ignition modes include hot-spot ignition and fast ignition. Methods of implosion include direct drive and indirect drive. Historically there has been significant work on indirectly driven targets with hot-spot ignition. Recently there has been increasing interest in directly driven targets with ion driven fast ignition. In principle, fast ignition might lead to improved target performance. On the other hand, fast ignition imposes stringent requirements on accelerators and beam physics. Furthermore, fast ignition magnifies the importance of a number of traditional target physics issues associated with ion beam energy deposition and fuel preheat. This paper will discuss the advantages and disadvantages of the various classes of targets. It will also discuss some issues that must be resolved to assess the feasibility of ion fast ignition

  9. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  10. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  11. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  12. Physics design of the HL-2A tokamak

    International Nuclear Information System (INIS)

    Gao Qingdi; Li Fangzhu; Zhang Jinhua; Pan Yudong; Jiao Yiming

    2005-10-01

    An overview report for the physics design of the HL-2A tokamak is presented. By numerically analyzing the plasma shaping and the vertical instability due to plasma elongation, the requirements for the currents of poloidal magnetic field coils and the control system are put forward. Controlling the plasma profile by using NBI (neutral beam injection) and LHCD (lower hybrid current drive) is investigated, and the high performance modes of operation in HL-2A and modeled and designed. The magnetohydrodynamic instabilities in improved confinement configuration (RS configuration) are studied so as to point out the way of plasma control to perform stationary high performance discharges in HL-2A. In order to offer data for updating the HL-2A divertor, performances of the divertor plasma are simulated. (authors)

  13. A flexible software design to determine the plasma boundary in Damavand tokamak

    International Nuclear Information System (INIS)

    Ghadiri, R.; Esteki, M.H.; Sadeghi, Y.

    2014-01-01

    A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code “The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)” was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C language. (author)

  14. Electron Shock Ignition of Inertial Fusion Targets

    International Nuclear Information System (INIS)

    Shang, W. L.; Betti, R.; Hu, S. X.; Woo, K.; Hao, L.

    2017-01-01

    Here, it is shown that inertial fusion targets designed with low implosion velocities can be shock ignited using laser–plasma interaction generated hot electrons (hot-e) to obtain high-energy gains. These designs are robust to multimode asymmetries and are predicted to ignite even for significantly distorted implosions. Electron shock ignition requires tens of kilojoules of hot-e, which can only be produced on a large laser facility like the National Ignition Facility, with the laser to hot-e conversion efficiency greater than 10% at laser intensities ~10 16 W/cm 2 .

  15. Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

    International Nuclear Information System (INIS)

    Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

    2017-01-01

    Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

  16. Preliminary design study of a steady state tokamak device

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

    1992-09-01

    Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

  17. Alpha particle effects in burning tokamak plasmas: overview and specific examples

    International Nuclear Information System (INIS)

    Sigmar, D.J.

    1986-07-01

    Using the total power balance of an ignited tokamak plasma as a guideline, a range of alpha driven effects is surveyed regarding their impact on achieving and maintaining fusion burn. Specific examples of MHD and kinetic modes and multi species transport dynamics are discussed, including the possible interaction of these categories of effects. This power balance approach rather than a straightforward enumeration of possible effects serves to reveal their non-linear dependence and the ensuing fragility of our understanding of the approach to and maintenance of ignition. Specific examples are given of the interaction between α-power driven sawtoothing and ideal MHD stability, and direct α-effects on MHD modes including kinetic corrections. Anomalous ion heat transport and central impurity peaking mechanisms and anomalous and collisional α-transport including the ambipolar electric field are discussed

  18. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

  19. 14 CFR 33.69 - Ignitions system.

    Science.gov (United States)

    2010-01-01

    ... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.69 Ignitions system. Each..., except that only one igniter is required for fuel burning augmentation systems. [Amdt. 33-6, 39 FR 35466... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Ignitions system. 33.69 Section 33.69...

  20. Recent results relevant to ignition physics and machine design issues

    International Nuclear Information System (INIS)

    Coppi, B.; Airoldi, A.; Bombarda, F.

    2001-01-01

    The plasma regimes under which ignition can be achieved involve a characteristic range of parameters and issues on which information has been provided by recent experiments. In particular, these results have motivated a new, in-depth analysis of the expected performance of the Ignitor machine as well as of the plasma processes that it can investigate. The main results and recent advances in the design of key systems of the machine are reported. (author)

  1. Recent results relevant to ignition physics and machine design issues

    International Nuclear Information System (INIS)

    Coppi, B.; Airoldi, A.; Bombarda, F.

    1999-01-01

    The plasma regimes under which ignition can be achieved involve a characteristic range of parameters and issues on which information has been provided by recent experiments. In particular, these results have motivated a new, in-depth analysis of the expected performance of the Ignitor machine as well as of the plasma processes that it can investigate. The main results and recent advances in the design of key systems of the machine are reported. (author)

  2. Influence of MHD effects and edge conditions on ITER helium ash accumulation and sustained ignition

    International Nuclear Information System (INIS)

    Redi, M.H.; Cohen, S.A.

    1990-06-01

    Dilution of reacting species by build-up of helium ash and its effect on ignition in the ITER tokamak have been studies in a series of simulations with the one-dimensional BALDUR transport code. Thermal diffusivities, obtained from ITER scaling laws and with radial variations observed in JET, gave τ E ∼ 2--4 sec. Refueling of deuterium and tritium maintained constant electron density, while carbon recycling was 100% and the helium ash recycling was varied from 1.0 to 0.5. Including MHD effects, specifically sawteeth and beta limits, we find that ignition can be sustained for 200 seconds with R helium = 0.95. These simulations, the only non-zero-dimensional, time-dependent simulations thus far made for ITER plasmas, emphasize that edge plasma conditions, MHD behavior, and helium particle transport are critical synergistic issues for sustained ignition. 27 refs., 2 figs., 1 tab

  3. DEMONSTRATION OF THE ITER IGNITION FIGURE OF MERIT AT q95>4 IN STATIONARY PLASMAS IN DIII-D

    International Nuclear Information System (INIS)

    WADE, M.R.; LUCE, T.C.; POLITZER, P.A.; FERRON, J.R.; HYATT, A.W.; SCOVILLE, J.T.; La HAYE, R.J.; KINSEY, J.E.; LASNIER, C.J.; MURAKAMI, M.; PETY, C.C.

    2002-01-01

    In order to maximize the probability of achieving ignition, the present International Thermonuclear Experimental Reactor (ITER) [1] design (as well as many of its predecessors) is based on operation at high plasma current. This constraint poses many significant engineering challenges, primarily related to the possibility of a sudden termination of the plasma current. Currents induced in the vessel and associated systems in such an event can lead to large forces, and runaway electrons may cause damage to the interior of the vacuum vessel. Present design methods (including those used for ITER) assume that the probability of experiencing such a major disruption increases with plasma current at fixed magnetic field and size. Because fusion performance is assumed to scale in a similar manner, reactor designs tend to seek a compromise between increased fusion performance and reduced susceptibility to disruptions, generally resulting in a design with q 95 ∼ 3.0. Discharges recently developed in the DIII-D tokamak offer a way to obtain equivalent fusion performance with more margin against disruption consequences, having obtained an ignition figure of merit comparable to the ITER baseline scenario with q 95 = 4.5. These discharges have been shown to be stationary on the thermal, resistive, and wall time scales and involve feedback control only of global quantities rather than profiles

  4. Shock ignition: a brief overview and progress in the design of robust targets

    International Nuclear Information System (INIS)

    Atzeni, S; Marocchino, A; Schiavi, A

    2015-01-01

    Shock ignition is a laser direct-drive inertial confinement fusion (ICF) scheme in which the stages of compression and hot spot formation are partly separated. The fuel is first imploded at a lower velocity than in conventional ICF, reducing the threats due to Rayleigh–Taylor instability. Close to stagnation, an intense laser spike drives a strong converging shock, which contributes to hot spot formation. This paper starts with a brief overview of the theoretical studies, target design and experimental results on shock ignition. The second part of the paper illustrates original work aiming at the design of robust targets and computation of the relevant gain curves. Following Chang et al (2010 Phys. Rev. Lett. 104 135002) a safety factor for high gain, ITF* (analogous to the ignition threshold factor ITF introduced by Clark et al (2008 Phys. Plasmas 15 056305)), is evaluated by means of parametric 1D simulations with artificially reduced reactivity. SI designs scaled as in Atzeni et al (2013 New J. Phys. 15 045004) are found to have nearly the same ITF*. For a given target, such ITF* increases with implosion velocity and laser spike power. A gain curve with a prescribed ITF* can then be simply generated by upscaling a reference target with that value of ITF*. An interesting option is scaling in size by reducing the implosion velocity to keep the ratio of implosion velocity to self-ignition velocity constant. At a given total laser energy, targets with higher ITF* are driven to higher implosion velocity and achieve a somewhat lower gain. However, a 1D gain higher than 100 is achieved at an (incident) energy below 1 MJ, an implosion velocity below 300 km s −1 and a peak incident power below 400 TW. 2D simulations of mispositioned targets show that targets with a higher ITF* indeed tolerate larger displacements. (paper)

  5. Design parameters of Tokamak-7 system

    International Nuclear Information System (INIS)

    Ivanov, D.P.; Keilin, V.E.; Klimenko, E.Yu.; Strelkov, V.S.

    Superconducting windings for the main magnetic field of Tokamak-7 are discussed. The parameters of this facility are based on the use of commercially available superconducting materials for fields up to 80 kOe. Experimental parameters are described. (U.S.)

  6. Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

    International Nuclear Information System (INIS)

    Rose, R.P.

    1979-05-01

    This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

  7. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  8. Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

    International Nuclear Information System (INIS)

    Kim, Seokho H.; Berry, Jan

    2011-01-01

    U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclear pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.

  9. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m 2 and a particle heat flux of 1 MW/m 2 . Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  10. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  11. Multi-megajoule heating of large tokamaks with high energy heavy ion beams

    International Nuclear Information System (INIS)

    Dei-Cas, R.

    1981-07-01

    The fast neutral injection heating and RF heating for tokamak like plasmas are now well established. We consider in this paper the use of high energy (approximately 1 GeV) heavy ions (Xe 132 ) to reach ignition in JET or INTOR like tokamaks. The main advantages of such a method will be outlined. The capture and the confinement of heavy ions have been analysed in a particular case and with the described RF linac it seems possible to inject in the order of 50 MJ in 1 sec with a modest increase of the effective charge Zsub(eff)<1.05 in a JET-like plasma for a particle life time of 1 sec and then the additional radiated power should be maintained at a relatively low level in comparison to the injected power

  12. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  13. Modeling of thermal effects on TIBER II [Tokamak Ignition/Burn Experimental Reactor] divertor during plasma disruption

    International Nuclear Information System (INIS)

    Bruhn, M.L.; Perkins, L.J.

    1987-01-01

    Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs. 14 refs

  14. Optimization of the National Ignition Facility primary shield design

    International Nuclear Information System (INIS)

    Annese, C.E.; Watkins, E.F.; Greenspan, E.; Miller, W.F.

    1993-10-01

    Minimum cost design concepts of the primary shield for the National Ignition laser fusion experimental Facility (NIF) are searched with the help of the optimization code SWAN. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables one to address the time-dependent problem using relatively simple, time-independent transport calculations, thus significantly simplifying the design process. A novel approach was used for the identification of the optimal combination of constituents that will minimize the shield cost; it involves the generation, with SWAN, of effectiveness functions for replacing materials on an equal cost basis. The minimum cost shield design concept was found to consist of a mixture of polyethylene and low cost, low activation materials such as SiC, with boron added near the shield boundaries

  15. Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

    International Nuclear Information System (INIS)

    Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

    1979-09-01

    The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

  16. A preliminary design and structural analysis on the central column for supporting the full 40 .deg. Sectors at tokamak in pit

    International Nuclear Information System (INIS)

    Nam, Kyoungo; Park, Hyun Ki; Kim, Dong Jin; Moon, Jaeh Wan; Kim, Byung Seok; Watson, Emma; Shaw, Robert

    2012-01-01

    The ITER Tokamak machine is composed of nine 40 .deg. sectors shaped of torus. Each 40.deg. sector is made up of one 40 .deg. vacuum vessel (VV), two 20 .deg. toroidal filed coils and associated vacuum vessel thermal shield (VVTS) segments which consist of one inboard and two outboard VVTS. The VV/TFC/VVTS 40 .deg. sectors are sub assembled at assembly building respectively at sector sub assembly tool and then nine sub assembled 40 .deg. sectors are finally assembled at in-pit of Tokamak building. ITER sector assembly tools are the purpose built assembly tools to assemble nine 40 .deg. sectors into the complete ITER Tokamak machine at Tokamak in pit. Based on the design description document, final report prepared by the ITER organization (IO) and tooling requirements, Korea has carried out the conceptual and preliminary design of these assembly tools. Especially, the central column is the main tool, which is composed of some hollow cylinders, to support full nine 40 .deg. sectors at Tokamak in pit. Configuration and structural analysis of the central column are presented

  17. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C.

    2001-01-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m 2 . In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper

  18. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C

    2001-09-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper.

  19. Suprathermal electron studies in the TCV tokamak: Design of a tomographic hard-x-ray spectrometer

    International Nuclear Information System (INIS)

    Gnesin, S.; Coda, S.; Decker, J.; Peysson, Y.

    2008-01-01

    Electron cyclotron resonance heating and electron cyclotron current drive, disruptive events, and sawtooth activity are all known to produce suprathermal electrons in fusion devices, motivating increasingly detailed studies of the generation and dynamics of this suprathermal population. Measurements have been performed in the past years in the tokamak a configuration variable (TCV) tokamak using a single pinhole hard-x-ray (HXR) camera and electron-cyclotron-emission radiometers, leading, in particular, to the identification of the crucial role of spatial transport in the physics of ECCD. The observation of a poloidal asymmetry in the emitted suprathermal bremsstrahlung radiation motivates the design of a proposed new tomographic HXR spectrometer reported in this paper. The design, which is based on a compact modified Soller collimator concept, is being aided by simulations of tomographic reconstruction. Quantitative criteria have been developed to optimize the design for the greatly variable shapes and positions of TCV plasmas.

  20. Analysis of EAST tokamak cryostat anti-seismic performance

    International Nuclear Information System (INIS)

    Chen Wei; Kong Xiaoling; Liu Sumei; Ni Xiaojun; Wang Zhongwei

    2014-01-01

    A 3-D finite element model for EAST tokamak cryostat is established by using ANSYS. On the basis of the modal analysis, the seismic response of the EAST tokamak cryostat structure is calculated according to an input of the design seismic response spectrum referring to code for seismic design of nuclear power plants. Calculation results show that EAST cryostat displacement and stress response is small under the action of earthquake. According to the standards, EAST tokamak cryostat structure under the action of design seismic can meet the requirements of anti-seismic design intensity, and ensure the anti-seismic safety of equipment. (authors)

  1. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  2. A sustained-arc ignition system for internal combustion engines

    Science.gov (United States)

    Birchenough, A. G.

    1977-01-01

    A sustained-arc ignition system was developed for internal combustion engines. It produces a very-long-duration ignition pulse with an energy in the order of 100 millijoules. The ignition pulse waveform can be controlled to predetermined actual ignition requirements. The design of the sustained-arc ignition system is presented in the report.

  3. Safety analyses of the ARIES tokamak reactor designs

    International Nuclear Information System (INIS)

    Herring, J.S.; McCarthy, K.A.; Dolan, T.J.

    1994-01-01

    The ARIES design has sought to maximize environmental and safety advantages of fusion through careful selection of materials and design. The ARIES-I tokamak reactor design consists of an SiC composite structure for the first wall and blanket, cooled by 10MPa helium. The breeder is Li 2 ZrO 3 . The divertor consists of SiC composite tubes coated with 2mm tungsten. Loss-of-cooling accident (LOCA) calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. The ARIES-II design includes liquid lithium and vanadium, both of which have low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe with a total 1km early dose of about 88rem (0.88Sv). ARIES-III was an extensive examination of the viability of a D- 3 He fueled tokamak power reactor. Because neutrons are produced only through side reactions (D+D→ 3 He+n, and D+D→T+p followed by D+T→ 4 He+n), the reactor has a reduced activation of the first wall and shield, low afterheat and class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. We modeled a LOCA in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, below 600 C, release fractions are small. We analyzed the disposition of the 20g per day of tritium that is produced by D-D reactions and removed by vacuum pumps. The ARIES-IV coolant is helium and the breeder is lithium oxide. The structure is silicon carbide. Since the neutron multiplier, beryllium metal, is combustible, releasing about 60MJkg -1 , beryllium is the chief source of chemical energy. Less than 10% of the 24 Na inventory is likely to diffuse out of the SiC during a fire in which the beryllium is consumed. Therefore, the offsite dose would be less than 200rem. ((orig.))

  4. A survey of pulse shape options for a revised plastic ablator ignition design

    Energy Technology Data Exchange (ETDEWEB)

    Clark, D. S.; Milovich, J. L.; Hinkel, D. E.; Salmonson, J. D.; Peterson, J. L.; Berzak Hopkins, L. F.; Eder, D. C.; Haan, S. W.; Jones, O. S.; Marinak, M. M.; Robey, H. F.; Smalyuk, V. A.; Weber, C. R. [Lawrence Livermore National Laboratory Livermore, California 94550 (United States)

    2014-11-15

    Recent experimental results using the “high foot” pulse shape for inertial confinement fusion ignition experiments on the National Ignition Facility (NIF) [Moses et al., Phys. Plasmas 16, 041006 (2009)] have shown encouraging progress compared to earlier “low foot” experiments. These results strongly suggest that controlling ablation front instability growth can significantly improve implosion performance even in the presence of persistent, large, low-mode distortions. Simultaneously, hydrodynamic growth radiography experiments have confirmed that ablation front instability growth is being modeled fairly well in NIF experiments. It is timely then to combine these two results and ask how current ignition pulse shapes could be modified to improve one-dimensional implosion performance while maintaining the stability properties demonstrated with the high foot. This paper presents such a survey of pulse shapes intermediate between the low and high foot extremes in search of an intermediate foot optimum. Of the design space surveyed, it is found that a higher picket version of the low foot pulse shape shows the most promise for improved compression without loss of stability.

  5. About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2002-01-01

    In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate

  6. Design and development of AXUV-based soft X-ray diagnostic camera for Aditya Tokamak

    International Nuclear Information System (INIS)

    Raval, Jayesh V.; Purohit, Shishir; Joisa, Y. Shankara

    2015-01-01

    The hot tokamak plasma emits Soft X-rays (SXR) in accordance with the temperature and density which are important to be studied. A silicon photo diode array (AXUV16ELG, Opto-diode, USA) based prototype SXR diagnostics is designed and developed for ADITYA tokamak for the study of SXR radial intensity profile, internal disruption (Saw-tooth crash), MHD instabilities. The diagnostic is having an array of 16 detector of millimeter dimension in a linear configuration. Absolute Extreme Ultra Violate (AXUV) detector offers compact size, improved time response with considerably good quantum efficiency in the soft X-ray range (200 eV to 10 keV). The diagnostic is designed in competence with the ADITYA tokamak protocol. The diagnostic design geometry allows detector view the plasma through a slot hole (0.5 cm X 0.05 cm), 10 μm Beryllium foil filter window, cutting off energies below 750 eV. The diagnostic was installed on Aditya vacuum vessel at radial port no 7 enabling the diagnostics to view the core plasma. The spatial resolution designed for diagnostic configuration is 1.3 cm at plasma centre. The signal generated from SXR detector is acquired with a dedicated single board computer based data acquisition system at 50 kHz. The diagnostic took observation for the ohmically heated plasma. The data was then processed to construct spatial and temporal profile of SXR intensity for Aditya plasma. This information was complimentary to the Silicon surface barrier detector (SBD) based array for the same plasma discharge. The cross calibration between the two was considerably satisfactory under the assumptions considered. (author)

  7. Overview of the TIBER II design

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.

    1987-01-01

    The TIBER II Tokamak Ignition/Burn Experimental Reactor design is the result of efforts by numerous people and institutions, including many fusion laboratories, universities, and industries. This overview attempts to place the work in perspective. Major features of the design are compact size, low cost, and steady-state operation. These are achieved through plasma shaping and innovative features such as radiation tolerant magnets and optimized shielding. While TIBER II can operate in a pulsed mode, steady-state is preferred for nuclear testing. Current drive is achieved by a combination of lower hybrid and neutral beams. In addition to 10 MW of ECR is added for disruption control and current drive profiling

  8. Development of High Efficiency Clean Combustion Engine Designs for Spark-Ignition and Compression-Ignition Internal Combustion Engines

    Energy Technology Data Exchange (ETDEWEB)

    Marriott, Craig; Gonzalez, Manual; Russell, Durrett

    2011-06-30

    This report summarizes activities related to the revised STATEMENT OF PROJECT OBJECTIVES (SOPO) dated June 2010 for the Development of High-Efficiency Clean Combustion engine Designs for Spark-Ignition and Compression-Ignition Internal Combustion Engines (COOPERATIVE AGREEMENT NUMBER DE-FC26-05NT42415) project. In both the spark- (SI) and compression-ignition (CI) development activities covered in this program, the goal was to develop potential production-viable internal combustion engine system technologies that both reduce fuel consumption and simultaneously met exhaust emission targets. To be production-viable, engine technologies were also evaluated to determine if they would meet customer expectations of refinement in terms of noise, vibration, performance, driveability, etc. in addition to having an attractive business case and value. Prior to this activity, only proprietary theoretical / laboratory knowledge existed on the combustion technologies explored The research reported here expands and develops this knowledge to determine series-production viability. Significant SI and CI engine development occurred during this program within General Motors, LLC over more than five years. In the SI program, several engines were designed and developed that used both a relatively simple multi-lift valve train system and a Fully Flexible Valve Actuation (FFVA) system to enable a Homogeneous Charge Compression Ignition (HCCI) combustion process. Many technical challenges, which were unknown at the start of this program, were identified and systematically resolved through analysis, test and development. This report documents the challenges and solutions for each SOPO deliverable. As a result of the project activities, the production viability of the developed clean combustion technologies has been determined. At this time, HCCI combustion for SI engines is not considered production-viable for several reasons. HCCI combustion is excessively sensitive to control variables

  9. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

  10. Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

    International Nuclear Information System (INIS)

    Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

    1996-10-01

    Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

  11. Results from D-T experiments on TFTR and implications for achieving an ignited plasma

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Blanchard, W.

    1998-01-01

    Progress in the performance of tokamak devices has enable not only the production of significant bursts of fusion energy from deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET) but, more importantly, the initial study of the physics of burning magnetically confined plasmas. As a result of the worldwide research on tokamaks, the scientific and technical issues for achieving an ignited plasma are better understood and the remaining questions more clearly defined. The principal research topics which have been studied on TFTR are transport, magnetohydrodynamic stability, and energetic particle confinement. The integration of separate solutions to problems in each of these research areas has also been of major interest. Although significant advances, such as the reduction of turbulent transport by means of internal transport barriers, identification of the theoretically predicted bootstrap current, and the study of the confinement of energetic fusion alpha-particles have been made, interesting and important scientific and technical issues remain. In this paper, the implications for the TFTR experiments for overcoming these remaining issues will be discussed

  12. Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

    International Nuclear Information System (INIS)

    Welander, A.S.; Deranian, R.D.; Humphreys, D.A.; Leuer, J.A.; Walker, M.L.

    2005-01-01

    Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII-D, the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution

  13. Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

    2009-01-01

    Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

  14. A self-description data framework for Tokamak control system design

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Ming; Zhang, Jing [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zheng, Wei, E-mail: zhengwei@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Hu, Feiran; Zhuang, Ge [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2015-10-15

    Highlights: • The SDD framework can be applied to different Tokamak devices. • We explain how configuration settings of control systems are described in SDD models, namely components and connections. • Evolving SDD models are stored in a dynamic schema database. • The SDD editor supports plug-and-play SDD models. - Abstract: A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices. The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

  15. A self-description data framework for Tokamak control system design

    International Nuclear Information System (INIS)

    Zhang, Ming; Zhang, Jing; Zheng, Wei; Hu, Feiran; Zhuang, Ge

    2015-01-01

    Highlights: • The SDD framework can be applied to different Tokamak devices. • We explain how configuration settings of control systems are described in SDD models, namely components and connections. • Evolving SDD models are stored in a dynamic schema database. • The SDD editor supports plug-and-play SDD models. - Abstract: A Tokamak device consists of numerous control systems, which need to be integrated. CODAC (Control, Data Access and Communication) system requires the configuration settings of these control systems to carry out the integration smoothly. SDD (Self-description data) is designed to describe the static configuration of control systems. ITER CODAC group has released an SDD software package for control system designers to manage the static configuration, but it is specific for ITER plant control systems. Following the idea of ITER SDD, we developed a flexible and scalable SDD framework to develop SDD software for J-TEXT and other sophisticated devices. The SDD framework describes the configuration settings of various control systems, including physical and logical elements and their relation information, in SDD models which are classified into Components and Connections. The framework is composed of three layers: the MongoDB database, an open-source, dynamic schema, NoSQL (Not Only SQL) database; the SDD service, which maps SDD models to MongoDB and handles the transaction and business logic; the SDD applications, which can be used to create and maintain SDD information, and generate various kinds of output using the stored SDD information.

  16. Design of an O-mode frequency modulated reflectometry system for the measurement of Alborz Tokamak plasma density profile

    Energy Technology Data Exchange (ETDEWEB)

    Koohestani, Saeideh [Department of Energy Engineering and physics, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of); Amrollahi, Reza, E-mail: amrollahi@aut.ac.ir [Department of Energy Engineering and physics, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of); Moradi, Gholamreza [Department of Electrical Engineering, Amirkabir University of Technology, Tehran, 15875-4413, Islamic Republic of Iran (Iran, Islamic Republic of)

    2016-12-15

    Reflectometry is a common method for plasma diagnostic, in which microwaves are launched into the plasma and reflected at the critical surfaces. Comparing the reflected microwave signals with the launched waves would give rise to the plasma density profiles. In the present study, an ordinary mode (O-mode) frequency modulation (FM) reflectometry system has been designed for the electron density profile measurement of the Alborz Tokamak plasma. This system has been considered to operate at K-band (18–26.5 GHz) frequency range and scan the frequency band between 18 to 26 GHz in 40 μS. The density profile from major radius r = 47.9–51.55 cm can be measured in Alborz Tokamak plasma. Based on the Alborz Tokamak operational conditions, the characteristic frequencies, and some dimensional limitations, all parts of reflectometer have been designed so that an appropriate efficiency with minimum attenuation, especially in transmitting/receiving system would be achieved. A dual antenna and an oversized waveguide of X-band (8–12 GHz) for transmitting and receiving purposes and a balanced detector for absolute phase determination have been utilized. The details of the Alborz Tokamak FM reflectometry components focusing on the antenna and waveguide design and mounting are described in this paper. Additionally, the procedure of plasma profile reconstruction using the system output signal is discussed. This system uses signal phase shift to determine the position of the cutoff layer.

  17. Tokamak heating by neutral beams and adiabatic compression

    International Nuclear Information System (INIS)

    Furth, H.P.

    1973-08-01

    ''Realistic'' models of tokamak energy confinement strongly favor reactor operation at the maximum MHD-stable β-value, in order to maximize plasma density. Ohmic heating is unsuitable for this purpose. Neutral-beam heating plus compression is well suited; however, very large requirements on device size and injection power seem likely for a DT ignition experiment using a Maxwellian plasma. Results of the ATC experiment are reviewed, including Ohmic heating, neutral-beam heating, and production of two-energy-component plasmas (energetic deuteron population in deuterium ''target plasma''). A modest extrapolation of present ATC parameters could give zero-power conditions in a DT experiment of the two-energy-component type. (U.S.)

  18. Low temperature plasma near a tokamak reactor limiter

    International Nuclear Information System (INIS)

    Braams, B.J.; Singer, C.E.

    1985-01-01

    Analytic and two-dimensional computational solutions for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a Tokamak Fusion Core Experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10 13 cm -3 or less and the thermal diffusivity in the edge region is 2 x 10 4 cm 2 /s or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so a low plasma temperature near the limiter should be considered a likely possibility

  19. Mechanical design of the coils encapsulated of toroidal field of Tokamak TPM1

    International Nuclear Information System (INIS)

    Caldino H, U.; Francois L, J. L.

    2014-10-01

    The TPM1 is a small Tokamak that belongs to the Centro de Investigacion en Ciencias Aplicadas y Tecnologia Avanzada of Instituto Politecnico Nacional (CICATA-IPN); the project is under construction. Currently it has the vacuum chamber, and is intended that the machine can operate with electric pulses of 10 ms to study the behavior of plasmas in order to provide knowledge in the field of nuclear fusion by magnetic confinement. To achieve this goal is necessary to design the toroidal field coils which operate the Tokamak. This paper presents an analysis which was performed to obtain the correct configuration of coils depending on design parameters for operation of the machine. Once determined this configuration, an analysis of electromagnetic forces present in normal machine operation on one coil was conducted, this to know the stresses in the encapsulation of the same. Considering the pulsed operation, a thickness of 5 mm is determined in the encapsulated, considering fatigue failure based on studies of fatigue failures in epoxy resins. (Author)

  20. Rf modeling and design of a folded waveguide launcher for the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Bigelow, T.S.; Fogelman, C.F.; Baity, F.W.; Carter, M.D.; Hoffman, D.J.; Ryan, P.M.; Yugo, J.J.; Golovato, S.N.; Bonoli, P.

    1993-01-01

    The folded waveguide (FWG) launcher is being investigated as an improved antenna configuration for plasma heating in the ion cyclotron range of frequencies (ICRF). A development FWG launcher was successfully tested at Oak Ridge National Laboratory (ORNL) with a low-density plasma load and found to have significantly greater power density capability than current strap-type antennas operating in similar plasmas. To further test the concept on a high density tokamak plasma, a collaboration has been set up between ORNL and Massachusetts Institute of Technology (MIT) to develop and test an 80-MHz, 2-MW FWG on the Alcator C-Mod tokamak at MIT. The radio frequency (rf) electromagnetic modeling techniques and laboratory measurements used in the design of this antenna are described in this paper. A companion paper describes the mechanical design of the FWG

  1. Resistivity effects in non-inductive RF current drive via helicity injection by Alfven waves: the case of conventional and small aspect ratio Tokamaks

    International Nuclear Information System (INIS)

    Bruma, C.; Cuperman, S.; Komoshvili, K.

    1996-01-01

    Supplementary non-inductive current drive and heating are necessary to bring Tokamak plasmas into the ignition regime. The resonant excitation of shear Alfven waves (SAW) - in the continuum range (CR) or/and in the discrete global Alfven eigenmode spectrum (GAE's) - represents one potential, suitable method for this purpose. Within the framework of ideal MHD, the current drive (CD) via helicity injection in Tokamak plasmas has been considered by Cuperman et al (1996) and Komoshvili et al. (1996). This work is concerned with the investigation of the non-ideal resistive MHD effects on both the excitation of SAW's (CR's and GAE's) and the generation of non-inductive current drive via helicity injection in Tokamak plasmas. The research covers Tokamak aspect ratios ranging between large value cases (R/a = 10) and the very tight value case (R/ a = 1.2). (authors)

  2. First wall and blanket design for the STARFIRE commercial tokamak power reactor

    International Nuclear Information System (INIS)

    Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

    1979-01-01

    The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

  3. Conceptual designs of power tokamak-type thermonuclear reactors

    International Nuclear Information System (INIS)

    Shejndlin, A.E.; Nedospasov, A.V.

    1978-01-01

    Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

  4. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  5. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  6. The ARIES-III D-3He tokamak reactor: Design-point determination and parametric studies

    International Nuclear Information System (INIS)

    Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

    1991-01-01

    The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-3 design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-1 is included. 11 refs., 5 figs

  7. Progress on LMJ targets for ignition

    International Nuclear Information System (INIS)

    Cherfils-Clerouin, C; Boniface, C; Bonnefille, M; Fremerye, P; Galmiche, D; Gauthier, P; Giorla, J; Lambert, F; Laffite, S; Liberatore, S; Loiseau, P; Malinie, G; Masse, L; Masson-Laborde, P E; Monteil, M C; Poggi, F; Seytor, P; Wagon, F; Willien, J L

    2010-01-01

    Targets designed to produce ignition on the Laser MegaJoule are presented. The LMJ experimental plans include the attempt of ignition and burn of an ICF capsule with 160 laser beams, delivering up to 1.4MJ and 380TW. New targets needing reduced laser energy with only a small decrease in robustness have then been designed for this purpose. Working specifically on the coupling efficiency parameter, i.e. the ratio of the energy absorbed by the capsule to the laser energy, has led to the design of a rugby-shaped cocktail hohlraum. 1D and 2D robustness evaluations of these different targets shed light on critical points for ignition, that can be traded off by tightening some specifications or by preliminary experimental and numerical tuning experiments.

  8. Progress on LMJ targets for ignition

    Energy Technology Data Exchange (ETDEWEB)

    Cherfils-Clerouin, C; Boniface, C; Bonnefille, M; Fremerye, P; Galmiche, D; Gauthier, P; Giorla, J; Lambert, F; Laffite, S; Liberatore, S; Loiseau, P; Malinie, G; Masse, L; Masson-Laborde, P E; Monteil, M C; Poggi, F; Seytor, P; Wagon, F; Willien, J L, E-mail: catherine.cherfils@cea.f [CEA, DAM, DIF, F-91297 Arpajon (France)

    2010-08-01

    Targets designed to produce ignition on the Laser MegaJoule are presented. The LMJ experimental plans include the attempt of ignition and burn of an ICF capsule with 160 laser beams, delivering up to 1.4MJ and 380TW. New targets needing reduced laser energy with only a small decrease in robustness have then been designed for this purpose. Working specifically on the coupling efficiency parameter, i.e. the ratio of the energy absorbed by the capsule to the laser energy, has led to the design of a rugby-shaped cocktail hohlraum. 1D and 2D robustness evaluations of these different targets shed light on critical points for ignition, that can be traded off by tightening some specifications or by preliminary experimental and numerical tuning experiments.

  9. Asymmetric-shell ignition capsule design to tune the low-mode asymmetry during the peak drive

    International Nuclear Information System (INIS)

    Gu, Jianfa; Dai, Zhensheng; Song, Peng; Zou, Shiyang; Ye, Wenhua; Zheng, Wudi; Gu, Peijun; Wang, Jianguo; Zhu, Shaoping

    2016-01-01

    The low-mode radiation flux asymmetry in the hohlraum is a main source of performance degradation in the National Ignition Facility (NIF) implosion experiments. To counteract the deleterious effects of the large positive P2 flux asymmetry during the peak drive, this paper develops a new tuning method called asymmetric-shell ignition capsule design which adopts the intentionally asymmetric CH ablator layer or deuterium-tritium (DT) ice layer. A series of two-dimensional implosion simulations have been performed, and the results show that the intentionally asymmetric DT ice layer can significantly improve the fuel ρR symmetry, hot spot shape, hot spot internal energy, and the final neutron yield compared to the spherical capsule. This indicates that the DT asymmetric-shell capsule design is an effective tuning method, while the CH ablator asymmetric-shell capsule could not correct the fuel ρR asymmetry, and it is not as effective as the DT asymmetric-shell capsule design.

  10. Asymmetric-shell ignition capsule design to tune the low-mode asymmetry during the peak drive

    Science.gov (United States)

    Gu, Jianfa; Dai, Zhensheng; Song, Peng; Zou, Shiyang; Ye, Wenhua; Zheng, Wudi; Gu, Peijun; Wang, Jianguo; Zhu, Shaoping

    2016-08-01

    The low-mode radiation flux asymmetry in the hohlraum is a main source of performance degradation in the National Ignition Facility (NIF) implosion experiments. To counteract the deleterious effects of the large positive P2 flux asymmetry during the peak drive, this paper develops a new tuning method called asymmetric-shell ignition capsule design which adopts the intentionally asymmetric CH ablator layer or deuterium-tritium (DT) ice layer. A series of two-dimensional implosion simulations have been performed, and the results show that the intentionally asymmetric DT ice layer can significantly improve the fuel ρR symmetry, hot spot shape, hot spot internal energy, and the final neutron yield compared to the spherical capsule. This indicates that the DT asymmetric-shell capsule design is an effective tuning method, while the CH ablator asymmetric-shell capsule could not correct the fuel ρR asymmetry, and it is not as effective as the DT asymmetric-shell capsule design.

  11. Asymmetric-shell ignition capsule design to tune the low-mode asymmetry during the peak drive

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Jianfa, E-mail: gu-jianfa@iapcm.ac.cn; Dai, Zhensheng, E-mail: dai-zhensheng@iapcm.ac.cn; Song, Peng; Zou, Shiyang; Ye, Wenhua; Zheng, Wudi; Gu, Peijun; Wang, Jianguo; Zhu, Shaoping [Institute of Applied Physics and Computational Mathematics, Beijing 100088 (China)

    2016-08-15

    The low-mode radiation flux asymmetry in the hohlraum is a main source of performance degradation in the National Ignition Facility (NIF) implosion experiments. To counteract the deleterious effects of the large positive P2 flux asymmetry during the peak drive, this paper develops a new tuning method called asymmetric-shell ignition capsule design which adopts the intentionally asymmetric CH ablator layer or deuterium-tritium (DT) ice layer. A series of two-dimensional implosion simulations have been performed, and the results show that the intentionally asymmetric DT ice layer can significantly improve the fuel ρR symmetry, hot spot shape, hot spot internal energy, and the final neutron yield compared to the spherical capsule. This indicates that the DT asymmetric-shell capsule design is an effective tuning method, while the CH ablator asymmetric-shell capsule could not correct the fuel ρR asymmetry, and it is not as effective as the DT asymmetric-shell capsule design.

  12. Safety and environmental process for the design and construction of the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Brereton, S.J., LLNL

    1998-05-27

    The National Ignition Facility (NIF) is a U.S. Department of Energy (DOE) laser fusion experimental facility currently under construction at the Lawrence Livermore National Laboratory (LLNL). This paper describes the safety and environmental processes followed by NIF during the design and construction activities.

  13. Neutronics design for a spherical tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Deng Meigen; Feng Kaiming; Yang Bangchao

    2002-01-01

    Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

  14. New designs of LMJ targets for early ignition experiments

    International Nuclear Information System (INIS)

    Clerouin, C; Bonnefille, M; Dattolo, E; Fremerye, P; Galmiche, D; Gauthier, P; Giorla, J; Laffite, S; Liberatore, S; Loiseau, P; Malinie, G; Masse, L; Poggi, F; Seytor, P

    2008-01-01

    The LMJ experimental plans include the attempt of ignition and burn of an ICF capsule with 40 laser quads, delivering up to 1.4MJ and 380TW. New targets needing reduced laser energy with only a small decrease in robustness are then designed for this purpose. A first strategy is to use scaled-down cylindrical hohlraums and capsules, taking advantage of our better understanding of the problem, set on theoretical modelling, simulations and experiments. Another strategy is to work specifically on the coupling efficiency parameter, i.e. the ratio of the energy absorbed by the capsule to the laser energy, which is with parametric instabilities a crucial drawback of indirect drive. An alternative design is proposed, made up of the nominal 60 quads capsule, named A1040, in a rugby-shaped hohlraum. Robustness evaluations of these different targets are in progress

  15. New designs of LMJ targets for early ignition experiments

    Energy Technology Data Exchange (ETDEWEB)

    Clerouin, C; Bonnefille, M; Dattolo, E; Fremerye, P; Galmiche, D; Gauthier, P; Giorla, J; Laffite, S; Liberatore, S; Loiseau, P; Malinie, G; Masse, L; Poggi, F; Seytor, P [Commissariat a l' Energie Atomique, DAM-Ile de France, BP 12 91680 Bruyeres-le-Chatel (France)], E-mail: catherine.cherfils@cea.fr

    2008-05-15

    The LMJ experimental plans include the attempt of ignition and burn of an ICF capsule with 40 laser quads, delivering up to 1.4MJ and 380TW. New targets needing reduced laser energy with only a small decrease in robustness are then designed for this purpose. A first strategy is to use scaled-down cylindrical hohlraums and capsules, taking advantage of our better understanding of the problem, set on theoretical modelling, simulations and experiments. Another strategy is to work specifically on the coupling efficiency parameter, i.e. the ratio of the energy absorbed by the capsule to the laser energy, which is with parametric instabilities a crucial drawback of indirect drive. An alternative design is proposed, made up of the nominal 60 quads capsule, named A1040, in a rugby-shaped hohlraum. Robustness evaluations of these different targets are in progress.

  16. Fusion ignition via a magnetically-assisted fast ignition approach

    OpenAIRE

    Wang, W. -M.; Gibbon, P.; Sheng, Z. -M.; Li, Y. T.; Zhang, J.

    2016-01-01

    Significant progress has been made towards laser-driven fusion ignition via different schemes, including direct and indirect central ignition, fast ignition, shock ignition, and impact ignition schemes. However, to reach ignition conditions, there are still various technical and physical challenges to be solved for all these schemes. Here, our multi-dimensional integrated simulation shows that the fast-ignition conditions could be achieved when two 2.8 petawatt heating laser pulses counter-pr...

  17. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    HUMPHREYS, D.A.; FERRON, J.R.; JOHNSON, R.D; LEUER, J.A.; PENAFLOR, B.G.; WALKER, M.L.; WELANDER, A.S.; KHAYRUTDINOV, R.R; DOKOUKA, V.; EDGELL, D.H.; FRANSSON, C.M.

    2004-03-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

  18. Confinement of ignition and yield on the National Ignition Facility

    International Nuclear Information System (INIS)

    Tobin, M.; Karpenko, V.; Foley, D.; Anderson, A.; Burnham, A.; Reitz, T.; Latkowski, J.; Bernat, T.

    1996-01-01

    The National Ignition Facility Target Areas and Experimental Systems has reached mid-Title I design. Performance requirements for the Target Area are reviewed and design changes since the Conceptual Design Report are discussed. Development activities confirm a 5-m radius chamber and the viability of a boron carbide first wall. A scheme for cryogenic target integration with the NIF Target Area is presented

  19. Design study of electrical power supply system for tokamak fusion power reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

  20. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  1. Fast ignition: Physics progress in the US fusion energy program and prospects for achieving ignition

    International Nuclear Information System (INIS)

    Key, M.; Andersen, C.; Cowan, T.

    2003-01-01

    Fast ignition (FI) has significant potential advantages for inertial fusion energy and it is therefore being studied as an exploratory concept in the US fusion energy program. FI is based on short pulse isochoric heating of pre-compressed DT by intense beams of laser accelerated MeV electrons or protons. Recent experimental progress in the study of these two heating processes is discussed. The goal is to benchmark new models in order to predict accurately the requirements for full-scale fast ignition. An overview is presented of the design and experimental testing of a cone target implosion concept for fast ignition. Future prospects and conceptual designs for larger scale FI experiments using planned high energy petawatt upgrades of major lasers in the US are outlined. A long-term road map for FI is defined. (author)

  2. Conceptual radiation shielding design of superconducting tokamak fusion device by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Kawasaki, Hiromitsu; Okuno, Koichi

    2010-01-01

    A complete 3D neutron and photon transport analysis by Monte Carlo transport code system PHITS (Particle and Heavy Ion Transport code System) have been performed for superconducting tokamak fusion device such as JT-60 Super Advanced (JT-60SA). It is possible to make use of PHITS in the port streaming analysis around the devices for the tokamak fusion device, the duct streaming analysis in the building where the device is installed, and the sky shine analysis for the site boundary. The neutron transport analysis by PHITS makes it clear that the shielding performance of the superconducting tokamak fusion device with the cryostat is improved by the graphical results. From the standpoint of the port streaming and the duct streaming, it is necessary to calculate by 3D Monte Carlo code such as PHITS for the neutronics analysis of superconducting tokamak fusion device. (author)

  3. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  4. Tokamak nonmaxwellian plasma dynamics in thermonuclear regime

    International Nuclear Information System (INIS)

    Cotsaftis, M.

    1987-01-01

    To reach ignition in a Tokamak plasma, large additional power P aux has to be injected in the device on top of the Joule heating P OH =VI r , V the plasma loop voltage, I r the resistive port of plasma current. Typi-cally JH ∼ 1 KeV, whereas ignition would requi- re IG ∼ 7-10 KeV. To gain this factor 7, one at least should inject additional power P aux ∼ 7P OH , supposing that nothing, especially the heat transport, is modified. This is by far not the case, with the so-called energy lifetime degradation, largely observed in oil experiments (but less dramatic with divertors), where energy lifetime tau E behaves like P tot -b with b∼1/2. In large machines where ignition temperature is the target to be imperiously reached, this implies to inject a very large power, typically P aux ∼ 50 to 100 MW, depending on size and parameters and on actual transport. So it is of importance with such figures, or even larger ones owing to uncertain ties, to optimize at best injected power by increasing its efficiency, both with respect to possible transport laws, and to physical phenomena governing heat flow in the system from the sources. This leads to the concept of scenarios, as time sequences of power input, where physical properties of the plasma system are used to build up ion temperature so that ignition is reached with minimum P tot = P OH + P aux and with fixed Q = Q o > 1. Elements for this study are given. The method is outlined. The resulting system of equations describing the evolution of a thermonuclear plasma is given

  5. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  6. Features of a point design for Fast Ignition

    International Nuclear Information System (INIS)

    Tabak, M; Clark, D; Town, R P J; Key, M H; Amendt, P; Ho, D; Meeker, D J; Shay, H D; Lasinski, B F; Kemp, A; Divol, L; Mackinnon, A J; Patel, P; Strozzi, D; Grote, D P

    2010-01-01

    Fast Ignition is an inertial fusion scheme in which fuel is first assembled and then heated to the ignition temperature with an external heating source. In this note we consider cone and shell implosions where the energy supplied by short pulse lasers is transported to the fuel by electrons. We describe possible failure modes for this scheme and how to overcome them. In particular, we describe two sources of cone tip failure, an axis jet driven from the compressed fuel mass and hard photon preheat leaking through the implosion shell, and laser prepulse that can change the position of laser absorption and the angular distribution of the emitted electrons.

  7. Features of a point design for fast ignition

    International Nuclear Information System (INIS)

    Tabak, M.; Clark, D.; Town, R.J.; Key, M.H.; Amendt, P.; Ho, D.; Meeker, D.J.; Shay, H.D.; Lasinski, B.F.; Kemp, A.; Divol, L.; Mackinnon, A.J.; Patel, P.; Strozzi, D.; Grote, D.P.

    2009-01-01

    Fast Ignition is an inertial fusion scheme in which fuel is first assembled and then heated to the ignition temperature with an external heating source. In this note we consider cone and shell implosions where the energy supplied by short pulse lasers is transported to the fuel by electrons. We describe possible failure modes for this scheme and how to overcome them. In particular, we describe two sources of cone tip failure, an axis jet driven from the compressed fuel mass and hard photon preheat leaking through the implosion shell, and laser prepulse that can change the position of laser absorption and the angular distribution of the emitted electrons.

  8. Design study of toroidal magnets for tokamak experimental power reactors

    International Nuclear Information System (INIS)

    Stekly, Z.J.J.; Lucas, E.J.

    1976-12-01

    This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed

  9. Ignition and fusion burn in fast ignition scheme

    International Nuclear Information System (INIS)

    Takabe, Hideaki

    1998-01-01

    The target physics of fast ignition is briefly reviewed by focusing on the ignition and fusion burn in the off-center ignition scheme. By the use of a two dimensional hydrodynamic code with an alpha heating process, the ignition condition is studied. It is shown that the ignition condition of the off-center ignition scheme coincides with that of the the central isochoric model. After the ignition, a nuclear burning wave is seen to burn the cold main fuel with a velocity of 2 - 3 x 10 8 cm/s. The spark energy required for the off-center ignition is 2 - 3 kJ or 10 - 15 kJ for the core density of 400 g/cm 3 or 200 g/cm 3 , respectively. It is demonstrated that a core gain of more than 2,000 is possible for a core energy of 100 kJ with a hot spark energy of 13 kJ. The requirement for the ignition region's heating time is also discussed by modeling a heating source in the 2-D code. (author)

  10. Energetic particle destabilization of shear Alfven waves in stellarators and tokamaks

    International Nuclear Information System (INIS)

    Spong, D.A.; Carreras, B.A.; Hedrick, C.L.; Leboeuf, J.N.; Weller, A.

    1994-01-01

    An important issue for ignited devices is the resonant destabilization of shear Alfven waves by energetic populations. These instabilities have been observed in a variety of toroidal plasma experiments in recent years, including: beam-destabilized toroidal Alfven instabilities (TAE) in low magnetic field tokamaks, ICRF destabilized TAE's in higher field tokamaks, and global Alfven instabilities (GAE) in low shear stellarators. In addition, excitation and study of these modes is a significant goal of the TFIR-DT program and a component of the ITER physics tasks. The authors have developed a gyrofluid model which includes the wave-particle resonances necessary to excite such instabilities. The TAE linear mode structure is calculated nonperturbatively, including many of the relevant damping mechanisms, such as: continuum damping, non-ideal effects (ion FLR and electron collisionality), and ion/electron Landau damping. This model has been applied to both linear and nonlinear regimes for a range of experimental cases using measured profiles

  11. Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

    International Nuclear Information System (INIS)

    Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

    2006-01-01

    The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

  12. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

    Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

    2017-01-01

    Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

  13. Trajectory planning of tokamak flexible in-vessel inspection robot

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hesheng [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Lai, Yinping; He, Tao [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China)

    2015-10-15

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

  14. Trajectory planning of tokamak flexible in-vessel inspection robot

    International Nuclear Information System (INIS)

    Wang, Hesheng; Chen, Weidong; Lai, Yinping; He, Tao

    2015-01-01

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

  15. 1D radiation analysis for the fusion ignition experiment ZEPHYR

    International Nuclear Information System (INIS)

    Brockmann, H.; Ohlig, U.; Krause, H.

    1979-11-01

    For the proposed tokamak ignition experiment ZEPHYR a nuclear radiation analysis is performed for both the prompt radiation during a discharge and the delayed radiation after discharges. This analysis is restricted to a one-dimensional geometrical approximation of the apparatus. Nuclear heating data are derived for the prompt radiation and the shielding concept is investigated. Results for the delayed radiation are given for three activation scenarios, that are anticipated for machine operation. The cooling phase after one of these scenarios is investigated in some detail in order to trace the dominant contributions to biological dose rates back to machine components and to specific activation reactions. (orig.)

  16. Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Yang, T.F.; Lee, A.Y.; Ruck, G.W.; Prevenslik, T.V.; Smeltzer, G.

    1979-01-01

    The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm 2 ) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm 2 ) and for ISX-B 2 (11 kA/cm 2 ). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure

  17. Ignition inhibitors for cellulosic materials

    International Nuclear Information System (INIS)

    Alvares, N.J.

    1976-01-01

    By exposing samples to various irradiance levels from a calibrated thermal radiation source, the ignition responses of blackened alpha-cellulose and cotton cloth with and without fire-retardant additives were compared. Samples treated with retardant compounds which showed the most promise were then isothermally pyrolyzed in air for comparisons between the pyrolysis rates. Alpha-cellulose samples containing a mixture of boric acid, borax, and ammonium di-hydrogen phosphate could not be ignited by irradiances up to 4.0 cal cm -2 s-1 (16.7 W/cm 2 ). At higher irradiances the specimens ignited, but flaming lasted only until the flammable gases were depleted. Cotton cloth containing a polymeric retardant with the designation THPC + MM was found to be ignition-resistant to all irradiances below 7.0 cal cm -2 s -1 (29.3 W/cm 2 ). Comparison of the pyrolysis rates of the retardant-treated alpha-cellulose and the retardant-treated cotton showed that the retardant mechanism is qualitatively the same. Similar ignition-response measurements were also made with specimens exposed to ionizing radiation. It was observed that gamma radiation results in ignition retardance of cellulose, while irradiation by neutrons does not

  18. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    1978-08-01

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  19. Assessing the prospects for achieving double-shell ignition on the National Ignition Facility using vacuum hohlraums

    Science.gov (United States)

    Amendt, Peter

    2006-10-01

    The goal of demonstrating ignition on the National Ignition Facility (NIF) has motivated a revisit of double-shell (DS) [1] targets as a complementary path to the baseline cryogenic single-shell approach [2]. Benefits of DS targets include room-temperature deuterium-tritium (DT) fuel preparation, minimal hohlraum-plasma-mediated laser backscatter, low threshold-ignition temperatures (4 keV) for relaxed hohlraum x-ray flux asymmetry tolerances [3], and loose shock timing requirements. On the other hand, DS ignition presents several challenges, including room-temperature containment of high-pressure DT (790 atm) in the inner shell; strict concentricity requirements on the two shells; development of nanoporous, low-density, metallic foams for structural support of the inner shell and hydrodynamic instability mitigation; and effective control of perturbation growth on the high-Atwood number interface between the DT fuel and the high-Z inner shell. Recent progress in DS ignition target designs using vacuum hohlraums is described, offering the potential for low levels of laser backscatter from stimulated Raman and Brillouin processes. In addition, vacuum hohlraums have the operational advantages of room temperature fielding and fabrication simplicity, as well as benefiting from extensive benchmarking on the Nova and Omega laser facilities. As an alternative to standard cylindrical hohlraums, a rugby-shaped geometry is also introduced that may provide energetics and symmetry tuning benefits for more robust DS designs with yields exceeding 10 MJ for 2 MJ of 3w laser energy. The recent progress in hohlraum designs and required advanced materials development are scheduled to culminate in a prototype demonstration of a NIF-scale ignition-ready DS in 2007. [1] P. Amendt et al., PoP 9, 2221 (2002). [2] J.D. Lindl et al., PoP 11, 339 (2004). [3] M.N. Chizhkov et al., Laser Part. Beams 23, 261 (2005). In collaboration with C. Cerjan, A. Hamza, J. Milovich and H. Robey.

  20. Overview of the STARFIRE reference commercial tokamak fusion power reactor design

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Barry, K.

    1980-01-01

    The purpose of the STARFIRE study is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup, superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield

  1. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield

  2. Real-time horizontal position control for Aditya-upgrade tokamak

    International Nuclear Information System (INIS)

    Kumar, Rohit; Ghosh, Joydeep; Tanna, Rakesh L.

    2015-01-01

    Position of plasma column is required to be controlled in real time for improved operation of any tokamak. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed, which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. (author)

  3. Divertor design for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Hill, D.N.; Braams, B.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities

  4. Digital control of plasma position in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

    2002-03-01

    Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

  5. Physical studies of fast ignition in China

    International Nuclear Information System (INIS)

    He, X T; Cai, Hong-bo; Wu, Si-zhong; Cao, Li-hua; Zhang, Hua; He, Ming-qing; Chen, Mo; Wu, Jun-feng; Zhou, Cang-tao; Zhou, Wei-Min; Shan, Lian-qiang; Wang, Wei-wu; Zhang, Feng; Bi, Bi; Zhao, Zong-qing; Gu, Yu-qiu; Zhang, Bao-han; Wang, Wei; Fang, Zhi-heng; Lei, An-le

    2015-01-01

    Fast ignition approach to inertial confinement fusion is one of the important goals today, in addition to central hot spot ignition in China. The SG-IIU and PW laser facilities are coupled to investigate the hot spot formation for fast ignition. The SG-III laser facility is almost completed and will be coupled with tens kJ PW lasers for the demonstration of fast ignition. In recent years, for physical studies of fast ignition, we have been focusing on the experimental study of implosion symmetry, M-band radiation preheating and mixing, advanced fast ignition target design, and so on. In addition, the modeling capabilities and code developments enhanced our ability to perform the hydro-simulation of the compression implosion, and the particle-in-cell (PIC) and hybrid-PIC simulation of the generation, transport and deposition of relativistic electron beams. Considerable progress has been achieved in understanding the critical issues of fast ignition. (paper)

  6. Magnet systems for ''Bean-Shaped'' tokamak

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bromberg, L.; Cohn, D.R.; Mendelsohn, S.; Okabayashi, M.; Raynes, D.; Stiner, D.J.; Todd, A.M.; Williams, J.E.C.

    1983-01-01

    Bean-shaping of tokamak plasmas offers a method of reaching stable operation at (beta) > 10%. In order to establish the indentation of the ''bean'', a set of high- current ''pushing coils'' (> 5 MA in a reactor) must be located at the midplane as close as possible to the inboard edge of the plasma. If located in the bore of the TF coils, then maintenance of the pushing coils may be impossible, and the interlocking coils may prevent reactor modularity. If located outside, the required pushing-coil current may be unacceptably large. This dilemma is overcome with a unique TF coil design in which the inboard leg is bent outward in the form of an arc. The pushing coils are housed in the midplane indentation of this arc, just outside the TF coils but adequately close to the plasma. The arched coil transfers forces to the top and bottom legs, where it can be reacted by a clamp structure if necessary. This technique would allow demountable joints to be placed near the inoard leg (for copper TF coils). Another design approach to the pushing coils is to use liquid Li or Na as the conductor and coolant. The liquid metal ''coils'' can be placed immediately adjacent to the plasma, giving optimal control of the plasma shape with minimal coil current, although modularity of the reactor may have to be surrendered. Conceptual designs are presented of PF and TF coil systems for an ignition test reactor with about 14% and for a full-scale demonstration reactor with about 20%, both using copper TF coils

  7. Recent developments in the design of conceptual fusion reactors

    International Nuclear Information System (INIS)

    Ribe, F.L.

    1977-01-01

    Since the first round of conceptual fusion reactor designs in 1973 - 1974, there has been considerable progress in design improvement. Two recent tokamak designs of the Wisconsin and Culham groups, with increased plasma beta and wall loading (power density), lead to more compact reactors with easier maintenance. The Reference Theta-Pinch Reactor has undergone considerable upgrading in the design of the first wall insulator and blanket. In addition, a conceptual homopolar energy storage and transfer system has been designed. In the case of the mirror reactor, there are design changes toward improved modular construction and ease of handling, as well as improved direct converters. Conceptual designs of toroidal-multiple-mirror, liner-compression, and reverse-field pinch reactors are also discussed. A design is presented of a toroidal multiple-mirror reactor that combines the advantages of steady-state operation and high-aspect ratio. The liner-compression reactor eliminates a major problem of radiation damage by using a liquid-metal first wall that also serves as a neutron-thermalizing blanket. The reverse-field pinch reactor operates at higher beta, larger current density and larger aspect ratio than a tokamak reactor. These properties allow the possibility of ignition by ohmic heating alone and greater ease of maintenance

  8. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

  9. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  10. Laser Ignition Microthruster Experiments on KKS-1

    Science.gov (United States)

    Nakano, Masakatsu; Koizumi, Hiroyuki; Watanabe, Masashi; Arakawa, Yoshihiro

    A laser ignition microthruster has been developed for microsatellites. Thruster performances such as impulse and ignition probability were measured, using boron potassium nitrate (B/KNO3) solid propellant ignited by a 1 W CW laser diode. The measured impulses were 60 mNs ± 15 mNs with almost 100 % ignition probability. The effect of the mixture ratios of B/KNO3 on thruster performance was also investigated, and it was shown that mixture ratios between B/KNO3/binder = 28/70/2 and 38/60/2 exhibited both high ignition probability and high impulse. Laser ignition thrusters designed and fabricated based on these data became the first non-conventional microthrusters on the Kouku Kousen Satellite No. 1 (KKS-1) microsatellite that was launched by a H2A rocket as one of six piggyback satellites in January 2009.

  11. The National Ignition Facility Project

    International Nuclear Information System (INIS)

    Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

    1994-01-01

    The mission of the National Ignition Facility is to achieve ignition and gain in ICF targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effect testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule and costs associated with the construction project

  12. MAST: a Mega Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

    1995-01-01

    The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)

  13. Optimized beryllium target design for indirectly driven inertial confinement fusion experiments on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Simakov, Andrei N., E-mail: simakov@lanl.gov; Wilson, Douglas C.; Yi, Sunghwan A.; Kline, John L.; Batha, Steven H. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Clark, Daniel S.; Milovich, Jose L.; Salmonson, Jay D. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States)

    2014-02-15

    For indirect drive inertial confinement fusion, Beryllium (Be) ablators offer a number of important advantages as compared with other ablator materials, e.g., plastic and high density carbon. In particular, the low opacity and relatively high density of Be lead to higher rocket efficiencies giving a higher fuel implosion velocity for a given X-ray drive; and to higher ablation velocities providing more ablative stabilization and reducing the effect of hydrodynamic instabilities on the implosion performance. Be ablator advantages provide a larger target design optimization space and can significantly improve the National Ignition Facility (NIF) [J. D. Lindl et al., Phys. Plasmas 11, 339 (2004)] ignition margin. Herein, we summarize the Be advantages, briefly review NIF Be target history, and present a modern, optimized, low adiabat, Revision 6 NIF Be target design. This design takes advantage of knowledge gained from recent NIF experiments, including more realistic levels of laser-plasma energy backscatter, degraded hohlraum-capsule coupling, and the presence of cross-beam energy transfer.

  14. System studies for quasi-steady-state advanced physics tokamak

    International Nuclear Information System (INIS)

    Reid, R.L.; Peng, Y.K.M.

    1983-11-01

    Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated

  15. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  16. Development of a pre-ignition submodel for hydrogen engines

    Energy Technology Data Exchange (ETDEWEB)

    Al-Baghdadi, Sadiq [University of Babylon (Iraq). Dept. of Mechanical Engineering

    2005-10-15

    In hydrogen-fuelled spark ignition engine applications, the onset of pre-ignition remains one of the prime limitations that needs to be addressed to avoid its incidence and achieve superior performance. This paper describes a new pre-ignition submodel for engine modelling codes. The effects of changes in key operating variables, such as compression ratio, spark timing, intake pressure, and temperature on pre-ignition limiting equivalence ratios are established both analytically and experimentally. With the established pre-ignition model, it is possible not only to investigate whether pre-ignition is observed with changing operating and design parameters, but also to evaluate those parameters' effects on the maximum possible pre-ignition intensity. (author)

  17. Fast-wave current drive modelling for large non-circular tokamaks

    International Nuclear Information System (INIS)

    Batchelor, D.B.; Goldfinger, R.C.; Jaeger, E.F.; Carter, M.D.; Swain, D.W.; Ehst, D.; Karney, C.F.F.

    1990-01-01

    It is widely recognized that a key element in the development of an attractive tokamak reactor, and in the successful achievement of the mission of ITER, is the development of an efficient steady-state current drive technique. Fast waves in the ion cyclotron range of frequencies hold the promise to drive steady-state currents with the required efficiency and to effectively heat the plasma to ignition. Advantages over other heating and current drive techniques include low cost per watt and the ability to penetrate to the center of high-density plasmas. The primary issues that must be resolved are: can an antenna array be designed to radiate the required spectrum of waves and have adequate coupling properties? Will the rf power be efficiently absorbed by electrons in the desired velocity range without unacceptable parasitic damping by fuel ions or α particles? What will the efficiency of current drive be when toroidal effects such as trapped particles are included? Can a practical rf system be designed and integrated into the device? We have addressed these issues by performing extensive calculations with ORION, a 2-D code, and the ray tracing code RAYS, which calculate wave propagation, absorption and current drive in tokamak geometry, and with RIP, a 2-D code that self-consistently calculates current drive with MHD equilibrium. An important figure of merit in this context is the integrated, normalized current drive efficiency. The calculations that we present here emphasize the ITER device. We consider a low-frequency scenario such that no ion resonances appear in the machine, and a high-frequency scenario such that the deuterium second harmonic resonance is just outside the plasma and the tritium second harmonic is in the plasma, midway between the magnetic axis and the inside edge. In both cases electron currents are driven by combined TTMP and Landau damping of the fast waves

  18. Engineering and thermal-hydraulic design of water cooled PFC for SST-1 tokamak

    International Nuclear Information System (INIS)

    Paritosh Chaudhuri; Santra, P.; Rabi Prakash, N.; Khirwadkar, S.; Arun Prakash, A.; Ramash, G.; Dubey, S.; Chenna Reddy, D.; Saxena, Y.C.

    2005-01-01

    Full text of publication follows: Steady state Superconducting Tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. It is a large aspect ratio tokamak with a major radius of 1.1 m and minor radius of 0.20 m. SST-1 is designed for plasma discharge duration of ∼1000 seconds to obtain fully steady state plasma with total input power up to 1.0 MW. First Wall or Plasma Facing Components (PFC) is one or the major sub-systems of SST-1 tokamak consisting of divertors, passive stabilizers, baffles, and poloidal limiters are designed to be compatible for steady state operation. All the PFC has the same basic configuration: graphite tiles are mechanically attached to a back plate made of high strength copper alloy, and SS tubes are embedded in the groove made in the back plate. Same tube will be used for cooling during plasma operation and baking during wall conditioning. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m 2 . In addition to remove high heat fluxes, the PFC are also designed to be compatible for high temperature baking at 350 deg. C. Water was chosen as the coolant because of its appropriate thermal properties, and while baking, hot nitrogen gas would flow through these tubes to bake the PFC at high temperature. Extensive studies, involving different flow parameters and various cooling layouts, has been done to select the final cooling parameters and layout, compatible for cooling and baking. During steady state operation, divertor and passive stabilizer heat loads are expected to be 0.6 and 0.25 MW/m 2 . The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Since the tile must be mechanically attached to the back plate (heat sink), the fitting technique must provide the highest mechanical stress so that thermal transfer efficiency is maximized. Proper brazing of cooling tube on the copper back

  19. Fast-ignition heavy-ion fusion target by jet impact

    International Nuclear Information System (INIS)

    Velarde, P.; Ogando, F.; Eliezer, S.; Martinez-Val, J.M.

    2005-01-01

    A new target design for HIF, based on the fast-ignition principles, is proposed. Unlike the previous designs proposed so far, in this case just one energy source is needed to drive the whole process to ignition. The ultra-fast deposition of energy onto the compressed core is produced in this case by hypervelocity jets generated during the process. The collision of jets converts their kinetic energy into thermal energy of the nuclear fuel, which is expected to produce ignition under proper design. The process is studied in this paper, describing its most relevant features like jet production and later collision

  20. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.

    1979-01-01

    The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

  1. Tokamak-FED plasma-engineering assessments

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lyon, J.F.; Rutherford, P.H.

    1981-01-01

    A wide range of plasma assumptions and scenarios has been examined for the current US tokamak FED concept, which aims to provide a controlled, long pulse (approx. 100 s) burning plasma with an energy amplification of greater than or equal to 5, a fusion power of 180 MW, and a neutron wall load of greater than or equal to 0.4 MW/m 2 . The results of the assessment suggest that the current FED baseline parameters of R = 5.0 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 5.4 MA are appropriate in reaching the above plasma performance, despite uncertainties in several plasma physics areas, such as confinement scaling, achievable beta, impurity control, etc. To enhance the probability of achieving fusion ignition and to provide some margin against a short fall in our physics projections in FED, a limited operating capability at B/sub t/ = 4.6 T and I/sub p/ = 6.5 MA is incorporated. Various other options and remedies have also been assessed aiming to alleviate the impact of the uncertainties on the FED design concept. These approaches appear promising because they can be studied within the current fusion physics program and may lead to drastically more cost-effective FED concepts

  2. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  3. Shock Timing Plan for the National Ignition Campaign

    Science.gov (United States)

    Munro, D. H.; Robey, H. F.; Spears, B. K.; Boehly, T. R.

    2006-10-01

    We report progress on the design of the shock timing tuning procedure for the 2010 ignition campaign at the National Ignition Facility. Our keyhole target design provides adequate drive surrogacy for us to time the first three shocks empirically. The major risk to our plan is hard x-ray preheat, which can cause the diagnostic window to become opaque.

  4. Progress on LMJ targets for ignition

    Energy Technology Data Exchange (ETDEWEB)

    Cherfils-Clerouin, C; Boniface, C; Bonnefille, M; Dattolo, E; Galmiche, D; Gauthier, P; Giorla, J; Laffite, S; Liberatore, S; Loiseau, P; Malinie, G; Masse, L; Masson-Laborde, P E; Monteil, M C; Poggi, F; Seytor, P; Wagon, F; Willien, J L, E-mail: catherine.cherfils@cea.f [CEA, DAM, DIF, F-91297 Arpajon (France)

    2009-12-15

    Targets designed to produce ignition on the Laser Megajoule (LMJ) are being simulated in order to set specifications for target fabrication. The LMJ experimental plans include the attempt of ignition and burn of an ICF capsule with 160 laser beams, delivering up to 1.4 MJ and 380 TW. New targets needing reduced laser energy with only a small decrease in robustness have then been designed for this purpose. Working specifically on the coupling efficiency parameter, i.e. the ratio of the energy absorbed by the capsule to the laser energy, has led to the design of a rugby-ball shaped cocktail hohlraum; with these improvements, a target based on the 240-beam A1040 capsule can be included in the 160-beam laser energy-power space. Robustness evaluations of these different targets shed light on critical points for ignition, which can trade off by tightening some specifications or by preliminary experimental and numerical tuning experiments.

  5. Progress on LMJ targets for ignition

    International Nuclear Information System (INIS)

    Cherfils-Clerouin, C; Boniface, C; Bonnefille, M; Dattolo, E; Galmiche, D; Gauthier, P; Giorla, J; Laffite, S; Liberatore, S; Loiseau, P; Malinie, G; Masse, L; Masson-Laborde, P E; Monteil, M C; Poggi, F; Seytor, P; Wagon, F; Willien, J L

    2009-01-01

    Targets designed to produce ignition on the Laser Megajoule (LMJ) are being simulated in order to set specifications for target fabrication. The LMJ experimental plans include the attempt of ignition and burn of an ICF capsule with 160 laser beams, delivering up to 1.4 MJ and 380 TW. New targets needing reduced laser energy with only a small decrease in robustness have then been designed for this purpose. Working specifically on the coupling efficiency parameter, i.e. the ratio of the energy absorbed by the capsule to the laser energy, has led to the design of a rugby-ball shaped cocktail hohlraum; with these improvements, a target based on the 240-beam A1040 capsule can be included in the 160-beam laser energy-power space. Robustness evaluations of these different targets shed light on critical points for ignition, which can trade off by tightening some specifications or by preliminary experimental and numerical tuning experiments.

  6. The National Ignition Facility Project

    International Nuclear Information System (INIS)

    Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

    1994-01-01

    The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule, and costs associated with the construction project

  7. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-01-01

    The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

  8. A design study of superconducting energy storage system for a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Ueda, Kazuo

    1979-01-01

    A design study of a superconducting inductive energy storage system (SC-IES) has been carried out in commission with JAERI. The SC-IES is to be applied to the power supply system for a tokamak experimental fusion reactor. The study was initiated with the definition of the requirement for the SC-IES and selection of the coil shape. The design of the coil and the cryostat has been followed. The design parameters are: stored energy 10 GJ, B max 8 T, conductor Nb-Ti, overall size 18 m (diameter) x 10 m (height). Technical problems and usefullness of SC-IES are discussed also. (author)

  9. Conceptual design of a Langmuir probe system for the tokamak ASDEX-UPGRADE

    International Nuclear Information System (INIS)

    Anastassiadis, A.; Tsingas, A.C.; Tsois, N.N.; Zoumbos, G.A.

    1985-05-01

    The conceptual design of a Langmuir probe system for the tokamak ASDEX-UPG is presented. This system is intended to carry out electrostatic measurements, in space and time, on the boundary layer plasma over the largest possible volume of the divertor plasma during discharges. Conducted by preset design requirements a fast probe system is proposed. During discharges signal measurements will be performed by means of a data-acquisition system and the motion will be controlled by a real-time computer. The desired information concerning plasma parameters and the motion of the probe system will be available to the diagnostician via a video display unit. (author)

  10. A conceptual design and structural stabilities of in-pit assembly tools for the completion of final sector assembly at tokamak hall

    International Nuclear Information System (INIS)

    Nam, K.O.; Park, H.K.; Kim, D.J.; Ahn, H.J.; Kim, K.K.; Im, K.; Shaw, R.

    2010-01-01

    The final assembly of main components of the International Thermonuclear Experimental Reactor (ITER) tokamak, Vacuum Vessel (VV) and Toroidal Field Coils (TFCs), is achieved by the sequential assembly of the nine sub-assembled 40 o sectors in tokamak pit. Each sub-assembled 40 o sector is composed of one VV 40 o sector, two TFCs, and in-between Vacuum Vessel Thermal Shield (VVTS) segments. Sub-assembly is carried out in the assembly building and then the sub-assembled sectors are transferred into tokamak pit, in sequence, to complete sector assembly. The role of in-pit assembly tool is to support and align the sub-assembled sectors in tokamak pit. It also plays the role of reference datum during assembly until the completion of main components assembly. Korea Domestic Agency (KO DA) has developed the conceptual design of most ITER purpose-built assembly tools under the collaboration with the ITER Organization. Among the conceptual designs carried out, this paper describes the function, the structure, the selected material and the design results of the in-pit assembly tools comprising central column, radial beams and their supports, TF inner supports and in-pit working floor. The results of structural analysis using ANSYS for the various loading cases are given as well. The resultant stresses and deflections turned out to fall within the allowable ranges.

  11. Magnet design approach for pulsed tokamak reactors

    International Nuclear Information System (INIS)

    Kim, S.H.; Evans, K. Jr.; Ehst, D.A.

    1983-12-01

    A choice of various operating modes of a tokamak reactor will have considerable impact on the fatigue lives and cost of ohmic heating (OH), equilibrium field (EF), and toroidal field (TF) coils. OH AND EF coil requirements and their costs, as well as the effects of the fringing fields of the EF coils on the TF coils, have been studied under cyclic operation in the range of N = 10 2 to 10 6 cycles, spanning the range from a noninductively driven reactor (STARFIRE) to a conventional ohmically driven reactor. For a reference design of TF coils the design of the central OH solenoid has been studied as a function of its maximum field, B/sup OH/. Increasing requirements for structural support lead to only negligible increases in volt-seconds for B/sup OH/ greater than or equal to 10.0 T. Fatigue failure of the OH coil is not a concern for N less than or equal to 10 5 ; for N approx. 10 6 fatigue limits the strain to small values, resulting in small increases in structural requirements and modest decreases in volt-seconds. Should noninductive current drive be achievable we note that this not only eliminates the OH coil, but it also permits EF coil placement in the inboard region, which facilitates the creation of highly shaped plasma cross sections (large triangularity, or bean-shaped equilibria). We have computed the stored energy, coil configuration and fringing fields for a number of EF coil design options

  12. Hot-wire ignition of AN-based emulsions

    Energy Technology Data Exchange (ETDEWEB)

    Turcotte, Richard; Goldthorp, Sandra; Badeen, Christopher M. [Canadian Explosives Research Laboratory, Natural Resources Canada, Ottawa, Ontario, K1A 0G1 (Canada); Chan, Sek Kwan [Orica Canada Inc., Brownsburg-Chatham, Quebec (Canada)

    2008-12-15

    Emulsions based on ammonium nitrate (AN) and water locally ignited by a heat source do not undergo sustained combustion when the pressure is lower than some threshold value usually called the Minimum Burning Pressure (MBP). This concept is now being used by some manufacturers as a basis of safety. However, before a technique to reliably measure MBP values can be designed, one must have a better understanding of the ignition mechanism. Clearly, this is required to avoid under ignitions which could lead to the erroneous interpretation of failures to ignite as failures to propagate. In the present work, facilities to prepare and characterize emulsions were implemented at the Canadian Explosives Research Laboratory. A calibrated hot-wire ignition system operated in a high-pressure vessel was also built. The system was used to study the ignition characteristics of five emulsion formulations as a function of pressure and ignition source current. It was found that these mixtures exhibit complicated pre-ignition stages and that the appearance of endotherms when the pressure is lowered below some threshold value correlates with the MBP. Thermal conductivity measurements using this hot-wire system are also reported. (Abstract Copyright [2008], Wiley Periodicals, Inc.)

  13. High-Gain Shock Ignition on the National Ignition Facility

    Science.gov (United States)

    Perkins, L. J.; Lafortune, K.; Bailey, D.; Lambert, M.; MacKinnon, A.; Blackfield, D.; Comley, A.; Schurtz, G.; Ribeyre, X.; Lebel, E.; Casner, A.; Craxton, R. S.; Betti, R.; McKenty, P.; Anderson, K.; Theobald, W.; Schmitt, A.; Atzeni, S.; Schiavi, A.

    2010-11-01

    Shock ignition offers the possibility for a near-term test of high-gain ICF on the NIF at less than 1MJ drive energy and with day-1 laser hardware. We will summarize the status of target performance simulations, delineate the critical issues and describe the R&D program to be performed in order to test the potential of a shock-ignited target on NIF. In shock ignition, compressed fuel is separately ignited by a late-time laser-driven shock and, because capsule implosion velocities are significantly lower than those required for conventional hotpot ignition, simulations indicate that fusion energy gains of 60 may be achievable at laser energies around 0.5MJ. Like fast ignition, shock ignition offers high gain but requires only a single laser with less demanding timing and focusing requirements. Conventional symmetry and stability constraints apply, thus a key immediate step towards attempting shock ignition on NIF is to demonstrate adequacy of low-mode uniformity and shock symmetry under polar drive

  14. Design and realization of the J-TEXT tokamak central control system

    International Nuclear Information System (INIS)

    Yang Zhoujun; Zhuang Ge; Hu Xiwei; Zhang Ming; Qiu Shengshun; Wang Zhijiang; Ding Yonghua; Pan Yuan

    2009-01-01

    The Joint Texas Experimental Tokamak (J-TEXT), a medium-sized conventional tokamak, serves as a user experimental facility in the China-USA fusion research community. Development of a flexible and easy-to-use J-TEXT central control system (CCS) is of supreme importance for users to coordinate the experimental scenarios with full integration into the discharge operation. This paper describes in detail the structure and functions of the J-TEXT CCS system as well as the performance in practical implementation. Results obtained from both commissioning and routine operations show that the J-TEXT CCS system can offer a satisfactory and effective control that is reliable and stable. The J-TEXT tokamak achieved high-quality performance in its first-ever experimental campaign with this CCS system.

  15. Mechanical design and thermo-hydraulic simulation of the infrared thermography diagnostic of the WEST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Micolon, Frédéric, E-mail: frederic.micolon@cea.fr; Courtois, Xavier; Aumeunier, Marie-Hélène; Chenevois, Jean-Pierre; Larroque, Sébastien

    2015-10-15

    The WEST (Tungsten (W) Environment in Steady state Tokamak) project is a partial rebuild of the Tore Supra tokamak to make it an X-point metallic environment machine aimed at testing ITER technologies in relevant plasma environment. For the safe operation of the WEST tokamak, infra-red (IR) thermography is a crucial diagnostic as it is a sound and reliable way to detect hotspots or abnormal heating patterns on the plasma facing components (PFCs). Thus WEST will be fitted with middle/short-IR (1.5–2 μm or 3–5 μm) cameras in the upper port plugs to get a full view of the critical PFCs (in particular the new lower divertor) and radio-frequency (RF) heating antennas and one camera at the equatorial level to monitor the new upper divertor and the first wall. This paper describes the design of the up-to-date optical system along with the hydraulic analysis and the thermal and mechanical finite element analysis conducted to ensure adequate heat extraction capabilities. Boundary conditions and simulation results will be presented and discussed as well as technological solutions retained.

  16. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  17. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1998-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  18. Hydrodynamic modeling and simulations of shock ignition thresholds

    Directory of Open Access Journals (Sweden)

    Lafon M.

    2013-11-01

    Full Text Available The Shock Ignition (SI scheme [1] offers to reduce the laser requirements by relaxing the implosion phase to sub-ignition velocities and later adding an intense laser spike. Depending on laser energy, target characteristics and implosion velocity, high gains are expected [2,3]. Relevant intensities for scaled targets imploded in the velocity range from 150 to 400 km/s are defined at ignition thresholds. A range of moderate implosion velocities is specified to match safe implosions. These conditions for target design are then inferred for relevant NIF and LMJ shock-ignited targets.

  19. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.

    1980-04-01

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  20. Pulse heating and ignition for off-centre ignited targets

    International Nuclear Information System (INIS)

    Mahdy, A.I.; Takabe, H.; Mima, K.

    1999-01-01

    An off-centre ignition model has been used to study the ignition conditions for laser targets related to the fast ignition scheme. A 2-D hydrodynamic code has been used, including alpha particle heating. The main goal of the study is the possibility of obtaining a high gain ICF target with fast ignition. In order to determine the ignition conditions, samples with various compressed core densities having different spark density-radius product (i.e. areal density) values were selected. The study was carried out in the presence of an external heating source, with a constant heating rate. A dependence of the ignition conditions on the heating rate of the external pulse is demonstrated. For a given set of ignition conditions, our simulation showed that an 11 ps pulse with 17 kJ of injected energy into the spark area was required to achieve ignition for a compressed core with a density of 200 g/cm 3 and 0.5 g/cm 2 spark areal density. It is shown that the ignition conditions are highly dependent on the heating rate of the external pulse. (author)

  1. Capsule Performance Optimization in the National Ignition Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Landen, O L; MacGowan, B J; Haan, S W; Edwards, J

    2009-10-13

    A capsule performance optimization campaign will be conducted at the National Ignition Facility to substantially increase the probability of ignition. The campaign will experimentally correct for residual uncertainties in the implosion and hohlraum physics used in our radiation-hydrodynamic computational models before proceeding to cryogenic-layered implosions and ignition attempts. The required tuning techniques using a variety of ignition capsule surrogates have been demonstrated at the Omega facility under scaled hohlraum and capsule conditions relevant to the ignition design and shown to meet the required sensitivity and accuracy. In addition, a roll-up of all expected random and systematic uncertainties in setting the key ignition laser and target parameters due to residual measurement, calibration, cross-coupling, surrogacy, and scale-up errors has been derived that meets the required budget.

  2. Capsule performance optimization in the national ignition campaign

    International Nuclear Information System (INIS)

    Landen, O L; MacGowan, B J; Haan, S W; Edwards, J

    2010-01-01

    A capsule performance optimization campaign will be conducted at the National Ignition Facility [1] to substantially increase the probability of ignition. The campaign will experimentally correct for residual uncertainties in the implosion and hohlraum physics used in our radiation-hydrodynamic computational models before proceeding to cryogenic-layered implosions and ignition attempts. The required tuning techniques using a variety of ignition capsule surrogates have been demonstrated at the Omega facility under scaled hohlraum and capsule conditions relevant to the ignition design and shown to meet the required sensitivity and accuracy. In addition, a roll-up of all expected random and systematic uncertainties in setting the key ignition laser and target parameters due to residual measurement, calibration, cross-coupling, surrogacy, and scale-up errors has been derived that meets the required budget.

  3. Capsule performance optimization in the national ignition campaign

    Science.gov (United States)

    Landen, O. L.; MacGowan, B. J.; Haan, S. W.; Edwards, J.

    2010-08-01

    A capsule performance optimization campaign will be conducted at the National Ignition Facility [1] to substantially increase the probability of ignition. The campaign will experimentally correct for residual uncertainties in the implosion and hohlraum physics used in our radiation-hydrodynamic computational models before proceeding to cryogenic-layered implosions and ignition attempts. The required tuning techniques using a variety of ignition capsule surrogates have been demonstrated at the Omega facility under scaled hohlraum and capsule conditions relevant to the ignition design and shown to meet the required sensitivity and accuracy. In addition, a roll-up of all expected random and systematic uncertainties in setting the key ignition laser and target parameters due to residual measurement, calibration, cross-coupling, surrogacy, and scale-up errors has been derived that meets the required budget.

  4. Progress towards polar-drive ignition for the NIF

    Science.gov (United States)

    McCrory, R. L.; Betti, R.; Boehly, T. R.; Casey, D. T.; Collins, T. J. B.; Craxton, R. S.; Delettrez, J. A.; Edgell, D. H.; Epstein, R.; Frenje, J. A.; Froula, D. H.; Gatu-Johnson, M.; Glebov, V. Yu.; Goncharov, V. N.; Harding, D. R.; Hohenberger, M.; Hu, S. X.; Igumenshchev, I. V.; Kessler, T. J.; Knauer, J. P.; Li, C. K.; Marozas, J. A.; Marshall, F. J.; McKenty, P. W.; Meyerhofer, D. D.; Michel, D. T.; Myatt, J. F.; Nilson, P. M.; Padalino, S. J.; Petrasso, R. D.; Radha, P. B.; Regan, S. P.; Sangster, T. C.; Séguin, F. H.; Seka, W.; Short, R. W.; Shvydky, A.; Skupsky, S.; Soures, J. M.; Stoeckl, C.; Theobald, W.; Yaakobi, B.; Zuegel, J. D.

    2013-11-01

    The University of Rochester's Laboratory for Laser Energetics (LLE) performs direct-drive inertial confinement fusion (ICF) research. LLE's Omega Laser Facility is used to study direct-drive ICF ignition concepts, developing an understanding of the underlying physics that feeds into the design of ignition targets for the National Ignition Facility (NIF). The baseline symmetric-illumination, direct-drive-ignition target design consists of a 1.5 MJ multiple-picket laser pulse that generates four shock waves (similar to the NIF baseline indirect-drive design) and is predicted to produce a one-dimensional (1D) gain of 48. LLE has developed the polar-drive (PD) illumination concept (for NIF beams in the x-ray-drive configuration) to allow the pursuit of direct-drive ignition without significant reconfiguration of the beam paths on the NIF. Some less-invasive changes in the NIF infrastructure will be required, including new phase plates, polarization rotators, and a PD-specific beam-smoothing front end. A suite of PD ignition designs with implosion velocities from 3.5 to 4.3 × 107 cm s-1 are predicted to have significant 2D gains (Collins et al 2012 Bull. Am. Phys. Soc. 57 155). Verification of the physics basis of these simulations is a major thrust of direct-drive implosion experiments on both OMEGA and the NIF. Many physics issues are being examined with symmetric beam irradiation on OMEGA, varying the implosion parameters over a wide region of design space. Cryogenic deuterium-tritium target experiments with symmetric irradiation have produced areal densities of ˜0.3 g cm-2, ion temperatures over 3 keV, and neutron yields in excess of 20% of the ‘clean’ 1D predicted value. The inferred Lawson criterion figure of merit (Betti R. et al 2010 Phys. Plasmas 17 058102) has increased from 1.7 atm s (IAEA 2010) to 2.6 atm s.

  5. Preparing for polar-drive ignition on the National Ignition Facility

    Directory of Open Access Journals (Sweden)

    McKenty P.W.

    2013-11-01

    Full Text Available The implementation of polar drive (PD at the National Ignition Facility (NIF will enable the execution of direct-drive implosions while the facility is configured for x-ray drive. The Laboratory for Laser Energetics (LLE, in collaboration with LLNL, LANL and GA, is implementing PD on the NIF. LLE has designed and participates in the use of PD implosions for diagnostic commissioning on the NIF. LLE has an active experimental campaign to develop PD in both warm and cryogenic target experiments on OMEGA. LLE and its partners are developing a Polar Drive Project Execution Plan, which will provide a detailed outline of the requirements, resources, and timetable leading to PD-ignition experiments on the NIF.

  6. Control of plasma poloidal shape and position in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

    1997-11-01

    Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

  7. Ignition on the National Ignition Facility: a path towards inertial fusion energy

    International Nuclear Information System (INIS)

    Moses, Edward I.

    2009-01-01

    the world's attention on the possibility of an ICF energy option. NIF experiments to demonstrate ignition and gain will use central-hot-spot (CHS) ignition, where a spherical fuel capsule is simultaneously compressed and ignited. The scientific basis for CHS has been intensively developed (Lindl 1998 Inertial Confinement Fusion: the Quest for Ignition and Energy Gain Using Indirect Drive (New York: American Institute of Physics)) and has a high probability of success. Achieving ignition with CHS will open the door for other advanced concepts, such as the use of high-yield pulses of visible wavelength rather than ultraviolet and fast ignition concepts (Tabak et al 1994 Phys. Plasmas 1 1626-34, Tabak et al 2005 Phys. Plasmas 12 057305). Moreover, NIF will have important scientific applications in such diverse fields as astrophysics, nuclear physics and materials science. This paper summarizes the design, performance and status of NIF, experimental plans for NIC, and will present laser inertial confinement fusion-fission energy (LIFE) as a path to achieve carbon-free sustainable energy.

  8. Ignition on the National Ignition Facility: a path towards inertial fusion energy

    Science.gov (United States)

    Moses, Edward I.

    2009-10-01

    the world's attention on the possibility of an ICF energy option. NIF experiments to demonstrate ignition and gain will use central-hot-spot (CHS) ignition, where a spherical fuel capsule is simultaneously compressed and ignited. The scientific basis for CHS has been intensively developed (Lindl 1998 Inertial Confinement Fusion: the Quest for Ignition and Energy Gain Using Indirect Drive (New York: American Institute of Physics)) and has a high probability of success. Achieving ignition with CHS will open the door for other advanced concepts, such as the use of high-yield pulses of visible wavelength rather than ultraviolet and fast ignition concepts (Tabak et al 1994 Phys. Plasmas 1 1626-34, Tabak et al 2005 Phys. Plasmas 12 057305). Moreover, NIF will have important scientific applications in such diverse fields as astrophysics, nuclear physics and materials science. This paper summarizes the design, performance and status of NIF, experimental plans for NIC, and will present laser inertial confinement fusion-fission energy (LIFE) as a path to achieve carbon-free sustainable energy.

  9. Density limit investigations near and significantly above the Greenwald limit on the tokamaks TEXTOR-94 and RTP

    International Nuclear Information System (INIS)

    Rapp, J.; Koslowski, H.R.; Pospieszczyk, A.; Salzedas, F.; Vries, P.C. de; Schueller, F.C.; Hokin, S.; Messiaen, A.M.

    2001-01-01

    Ignition scenarios like those developed for ITER require plasma densities which will be close or above the Greenwald limit. Generally it is observed that exceeding this limit may lead to a degradation of plasma confinement or to a violent end of the discharge. The achievable density limit and the related processes, such as radiative instabilities and MHD phenomena, which eventually lead to disruption, have been investigated in the limiter tokamaks TEXTOR-94 and RTP. (author)

  10. A consistency analysis on the tokamak reactor plasmas

    International Nuclear Information System (INIS)

    Fukuyama, A.; Itoh, S.-I.; Itoh, K.

    1990-12-01

    The parameter regime which simultaneously fulfills the various physics constraints are looked for in the case of ITER grade tokamaks. The consistency analysis code is applied. It is found that, if the energy confinement time reaches 1.6 times of the prediction of the L-mode scaling law, the Q-value of about 4 is possible for the full current drive operation at the input power P in of 100MW (Q is the ratio of fusion output and P in ). In the ignition mode, where half of the current is inductively sustained, Q approaches to 15 for this circulating power. If only the L-mode is realized, Q is about 1.5 for P in ≅100 MW. (author)

  11. Development of Operation Scenario for Spherical Tokamak at SNU

    International Nuclear Information System (INIS)

    Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

    2009-01-01

    Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

  12. Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Hennen, B.A.; Westerhof, E.; De Baar, M.R.; Nuij, P.W.J.M.; Steinbuch, M.

    2012-01-01

    Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron cyclotron resonance heating and current drive (ECRH/ECCD) with a tearing mode and the stabilization of a mode at a specific width. In order to simulate these control tasks, the time evolution of a tearing mode subject to suppression by ECRH/ECCD and destabilization by a magnetic perturbation field is modelled using the generalized Rutherford equation. The model includes an equilibrium model and an ECRH/ECCD launcher model. The dynamics and static equilibria of this model are analysed. The model is linearized and based on the linearized model, linear feedback controllers are designed and simulated, demonstrating both alignment and width control of tearing modes in TEXTOR. (paper)

  13. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  14. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  15. Toroidal field coil design concept and structural support system for CTHR

    Energy Technology Data Exchange (ETDEWEB)

    Chianese, R. B.; Kelly, J. L.; Ruck, G. W.

    1980-09-01

    The CTHR conceptual design consists of a magnetically confined (tokamak) fusion reactor fitted with a fertile uranium blanket. The fusion driver concept was based on an ignited plasma. All concepts and parameters were selected on the basis that technical feasibility would be achieved by 1995 to assure a viable commercial operation in the early to mid-21st century. The reactor was designed to achieve good fissile fuel production, with electricity production being a second order priority. However, the resulting concepts that evolved were all excellent power producers which significantly improved the economic performance. The subsystems discussed in the following paragraphs provide a background of the application for the TF coil design described in this report.

  16. Toroidal field coil design concept and structural support system for CTHR

    International Nuclear Information System (INIS)

    Chianese, R.B.; Kelly, J.L.; Ruck, G.W.

    1980-09-01

    The CTHR conceptual design consists of a magnetically confined (tokamak) fusion reactor fitted with a fertile uranium blanket. The fusion driver concept was based on an ignited plasma. All concepts and parameters were selected on the basis that technical feasibility would be achieved by 1995 to assure a viable commercial operation in the early to mid-21st century. The reactor was designed to achieve good fissile fuel production, with electricity production being a second order priority. However, the resulting concepts that evolved were all excellent power producers which significantly improved the economic performance. The subsystems discussed in the following paragraphs provide a background of the application for the TF coil design described in this report

  17. Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1977-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

  18. Reactor costs and maintenance, with reference to the Culham Mark II conceptual Tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1976-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, due mainly to the low power density of the fusion reactor which affects both the reactor and building costs. In order to reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (orig.) [de

  19. Transmutation blanket design for a Tokamak system

    International Nuclear Information System (INIS)

    Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria A. Fortini; Costa, Antonella L.

    2011-01-01

    Sub-critical advanced reactor with a D-T fusion neutron source based on Tokamak technology is an innovative type of nuclear system. Due to the high quantity of neutrons produced by fusion reactions, it could be well spent in the transmutation process of the transuranic elements. Nevertheless, to achieve a successful transmutation, it is necessary to know the neutron fluence along the radial axis and its characteristics. In this work, it evaluated the neutron flux and interaction frequency along the radial axis changing the material of the first wall. W-alloy, beryllium and the combination of both were studied and regions more suitable to transmutation were determined. The results demonstrated that the better zone to place a transmutation blanket is limited by the heat sink and the shield block. Material arrangements W-alloy/W-alloy and W-alloy/Beryllium would be able to hold the requirements of high fluence and hardening spectrum needed to transuranic transmutation. The system was simulated using the MCNP5 code, the ITER Final Design Report, 2001, and the FENDL/MC-2.1 nuclear data library. (author)

  20. What is past is prologue: future directions in Tokamak Power Reactor Design Research

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    After reviewing the first generation of studies and the primary conclusions they produced, four current designs are discussed that are representative of present trends in this area of research. In particular, the trends towards reduced reactor size and higher neutron wall loadings are discussed. Moving in this direction requires new approaches to many subsystem designs. New approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets are described. A discussion is given of the future role of conceptual reactor design research and the need for close interactions with ongoing experiments in fusion technology

  1. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

    htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma

  2. On some interesting properties of the working temperature in a Tokamak reactor

    International Nuclear Information System (INIS)

    Brunelli, B.

    1987-01-01

    A D,T burning plasma has two equilibrium temperatures T/sub 1/ and T/sub 2/ wherein power-in equals power-out. At marginal ignition: T/sub 2/ = T/sub 1/ = T/sub 0/. It is shown that, under hypothesis usually satisfied in a Tokamak reactor, the temperature T/sub 0/ has a peculiar behaviour with respect to the reactor parameters. Simple expressions are given for T/sub 0/ T/sub 1/ and T/sub 2/ which have been found quite straightforward for a well-grounded discussion of the thermal reactor dynamics. Typical cases of interest are discussed

  3. Optical design of the National Ignition Facility main laser and switchyard/target area beam transport systems

    Science.gov (United States)

    Miller, John L.; English, R. Edward, Jr.; Korniski, Ronald J.; Rodgers, J. Michael

    1999-07-01

    The optical design of the main laser and transport mirror sections of the National Ignition Facility are described. For the main laser the configuration, layout constraints, multiple beam arrangement, pinhole layout and beam paths, clear aperture budget, ray trace models, alignment constraints, lens designs, wavefront performance, and pupil aberrations are discussed. For the transport mirror system the layout, alignment controls and clear aperture budget are described.

  4. Spherical tokamak without external toroidal fields

    International Nuclear Information System (INIS)

    Kaw, P.K.; Avinash, K.; Srinivasan, R.

    2001-01-01

    A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)

  5. Application of a two fluid theoretical plasma transport model on current tokamak reactor designs

    International Nuclear Information System (INIS)

    Ibrahim, E.; Fowler, T.K.

    1987-06-01

    In this work, the new theoretical transport models to TIBER II design calculations are described and the results are compared with recent experimental data in large tokamaks (TFTR, JET). Tang's method is extended to a two-fluid model treating ions and electrons separately. This allows for different ion and electron temperatures, as in recent low-density experiments in TFTR, and in the TIBER II design itself. The discussion is divided into two parts: (1) Development of the theoretical transport model and (2) calibration against experiments and application to TIBER II

  6. Review of the National Ignition Campaign 2009-2012

    International Nuclear Information System (INIS)

    Lindl, John; Landen, Otto; Edwards, John; Moses, Ed

    2014-01-01

    The National Ignition Campaign (NIC) was a multi-institution effort established under the National Nuclear Security Administration of DOE in 2005, prior to the completion of the National Ignition Facility (NIF) in 2009. The scope of the NIC was the planning and preparation for and the execution of the first 3 yr of ignition experiments (through the end of September 2012) as well as the development, fielding, qualification, and integration of the wide range of capabilities required for ignition. Besides the operation and optimization of the use of NIF, these capabilities included over 50 optical, x-ray, and nuclear diagnostic systems, target fabrication facilities, experimental platforms, and a wide range of NIF facility infrastructure. The goal of ignition experiments on the NIF is to achieve, for the first time, ignition and thermonuclear burn in the laboratory via inertial confinement fusion and to develop a platform for ignition and high energy density applications on the NIF. The goal of the NIC was to develop and integrate all of the capabilities required for a precision ignition campaign and, if possible, to demonstrate ignition and gain by the end of FY12. The goal of achieving ignition can be divided into three main challenges. The first challenge is defining specifications for the target, laser, and diagnostics with the understanding that not all ignition physics is fully understood and not all material properties are known. The second challenge is designing experiments to systematically remove these uncertainties. The third challenge is translating these experimental results into metrics designed to determine how well the experimental implosions have performed relative to expectations and requirements and to advance those metrics toward the conditions required for ignition. This paper summarizes the approach taken to address these challenges, along with the progress achieved to date and the challenges that remain. At project completion in 2009, NIF lacked

  7. A novel three-axis cylindrical hohlraum designed for inertial confinement fusion ignition

    Science.gov (United States)

    Kuang, Longyu; Li, Hang; Jing, Longfei; Lin, Zhiwei; Zhang, Lu; Li, Liling; Ding, Yongkun; Jiang, Shaoen; Liu, Jie; Zheng, Jian

    2016-10-01

    A novel ignition hohlraum for indirect-drive inertial confinement fusion is proposed, which is named three-axis cylindrical hohlraum (TACH). TACH is a kind of 6 laser entrance holes (LEHs) hohlraum, which is orthogonally jointed of three cylindrical hohlraums. Laser beams are injected through every entrance hole with the same incident angle of 55°. A view-factor simulation result shows that the time-varying drive asymmetry of TACH is less than 1.0% in the whole drive pulse period without any supplementary technology. Coupling efficiency of TACH is close to that of 6 LEHs spherical hohlraum with corresponding size. Its plasma-filling time is close to that of typical cylindrical ignition hohlraum. Its laser plasma interaction has as low backscattering as the outer cone of the cylindrical ignition hohlraum. Therefore, TACH combines most advantages of various hohlraums and has little predictable risk, providing an important competitive candidate for ignition hohlraum.

  8. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  9. High-Performance Cryogenic Designs for OMEGA and the National Ignition Facility

    Science.gov (United States)

    Goncharov, V. N.; Collins, T. J. B.; Marozas, J. A.; Regan, S. P.; Betti, R.; Boehly, T. R.; Campbell, E. M.; Froula, D. H.; Igumenshchev, I. V.; McCrory, R. L.; Myatt, J. F.; Radha, P. B.; Sangster, T. C.; Shvydky, A.

    2016-10-01

    The main advantage of laser symmetric direct drive (SDD) is a significantly higher coupled drive laser energy to the hot-spot internal energy at stagnation compared to that of laser indirect drive. Because of coupling losses resulting from cross-beam energy transfer (CBET), however, reaching ignition conditions on the NIF with SDD requires designs with excessively large in-flight aspect ratios ( 30). Results of cryogenic implosions performed on OMEGA show that such designs are unstable to short-scale nonuniformity growth during shell implosion. Several CBET reduction strategies have been proposed in the past. This talk will discuss high-performing designs using several CBET-mitigation techniques, including using drive laser beams smaller than the target size and wavelength detuning. Designs that are predicted to reach alpha burning regimes as well as a gain of 10 to 40 at the NIF-scale will be presented. Hydrodynamically scaled OMEGA designs with similar CBET-reduction techniques will also be discussed. This material is based upon work supported by the Department Of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  10. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  11. Direct electrical arc ignition of hybrid rocket motors

    Science.gov (United States)

    Judson, Michael I., Jr.

    Hybrid rockets motors provide distinct safety advantages when compared to traditional liquid or solid propellant systems, due to the inherent stability and relative inertness of the propellants prior to established combustion. As a result of this inherent propellant stability, hybrid motors have historically proven difficult to ignite. State of the art hybrid igniter designs continue to require solid or liquid reactants distinct from the main propellants. These ignition methods however, reintroduce to the hybrid propulsion system the safety and complexity disadvantages associated with traditional liquid or solid propellants. The results of this study demonstrate the feasibility of a novel direct electrostatic arc ignition method for hybrid motors. A series of small prototype stand-alone thrusters demonstrating this technology were successfully designed and tested using Acrylonitrile Butadiene Styrene (ABS) plastic and Gaseous Oxygen (GOX) as propellants. Measurements of input voltage and current demonstrated that arc-ignition will occur using as little as 10 watts peak power and less than 5 joules total energy. The motor developed for the stand-alone small thruster was adapted as a gas generator to ignite a medium-scale hybrid rocket motor using nitrous oxide /and HTPB as propellants. Multiple consecutive ignitions were performed. A large data set as well as a collection of development `lessons learned' were compiled to guide future development and research. Since the completion of this original groundwork research, the concept has been developed into a reliable, operational igniter system for a 75mm hybrid motor using both gaseous oxygen and liquid nitrous oxide as oxidizers. A development map of the direct spark ignition concept is presented showing the flow of key lessons learned between this original work and later follow on development.

  12. Design of new superconducting central solenoid of SST-1 tokamak

    International Nuclear Information System (INIS)

    Prasad, Upendra; Pradhan, Subrata; Ghate, Mahesh

    2015-01-01

    The key role of the central solenoid (CS) magnet of a Tokamak is for gas breakdown, ramp up and maintaining of plasma current for longer duration. The magnetic flux change in CS along with other PF coils generates magnetic null and induces electric field in toroidal direction. The induced toroidal electric field accelerates the residual electrons which collide with the neutrals and an avalanche takes place which led to the net plasma in the vacuum vessel of a Tokamak. In order to maximize the CS volt-sec capability, the higher magnetic field with a greater magnetic flux linkage is necessary. In order to facilitate all these requirements of SST-1 a new superconducting CS has been designed for SST-1. The design of new central solenoid has two bases; first one is physics and second is smart engineering in limited bore diameter of ∼655 mm. The physics basis of the design includes volt-sec storage capacity of ∼0.8 volt-sec, magnetic field null around 0.2 m over major radius of 1.1 m and toroidal electric field of ∼0.3 volt/m.The engineering design of new CS consists of Nb 3 Sn cable in conduit conductor (CICC) of operating current of 14 kA @ 4.5 K at 6 T, consolidated winding pack, smart quench detection system, protection system, housing cryostat and conductor terminations and joint design. The winding pack consists of 576 numbers of turns distributed in four layers with 0.75 mm FRP tape soaked with cyanide Easter based epoxy resin turn insulation and 3 mm of ground insulation. The inter-layer low resistance (∼1 nΩ) at 14 kA @ 4.5 K terminal praying hand joints has been designed for making winding pack continuous. The total height of winding pack is 2500 mm. The stored energy of this winding pack is ∼3 MJ at 14 kA of operating current. The expected heat load at cryogenic temperature is ∼10 W per layer, which requires helium mass flow rate of 1.4 g/s at 1.4 bars @ 4.5 K. The typical diameter and height of housing cryostat are 650 mm and 2563 mm with 80 K

  13. The Ignition Target for the National Ignition Facility

    International Nuclear Information System (INIS)

    Atherton, L J; Moses, E I; Carlisle, K; Kilkenny, J

    2007-01-01

    The National Ignition Facility (NIF) is a 192 beam Nd-glass laser facility presently under construction at Lawrence Livermore National Laboratory (LLNL) for performing inertial confinement fusion (ICF) and experiments studying high energy density (HED) science. When completed in 2009, NIF will be able to produce 1.8 MJ, 500 TW of ultraviolet light for target experiments that will create conditions of extreme temperatures (>10 8 K), pressures (10-GBar) and matter densities (> 100 g/cm 3 ). A detailed program called the National Ignition Campaign (NIC) has been developed to enable ignition experiments in 2010, with the goal of producing fusion ignition and burn of a deuterium-tritium (DT) fuel mixture in millimeter-scale target capsules. The first of the target experiments leading up to these ignition shots will begin in 2008. Targets for the National Ignition Campaign are both complex and precise, and are extraordinarily demanding in materials fabrication, machining, assembly, cryogenics and characterization. An overview of the campaign for ignition will be presented, along with technologies for target fabrication, assembly and metrology and advances in growth and x-ray imaging of DT ice layers. The sum of these efforts represents a quantum leap in target precision, characterization, manufacturing rate and flexibility over current state-of-the-art

  14. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

  15. Mechanical design of the coils encapsulated of toroidal field of Tokamak TPM1; Diseno mecanico del encapsulado de las bobinas de campo toroidal del Tokamak TPM1

    Energy Technology Data Exchange (ETDEWEB)

    Caldino H, U.; Francois L, J. L., E-mail: ucaldino@outlook.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    The TPM1 is a small Tokamak that belongs to the Centro de Investigacion en Ciencias Aplicadas y Tecnologia Avanzada of Instituto Politecnico Nacional (CICATA-IPN); the project is under construction. Currently it has the vacuum chamber, and is intended that the machine can operate with electric pulses of 10 ms to study the behavior of plasmas in order to provide knowledge in the field of nuclear fusion by magnetic confinement. To achieve this goal is necessary to design the toroidal field coils which operate the Tokamak. This paper presents an analysis which was performed to obtain the correct configuration of coils depending on design parameters for operation of the machine. Once determined this configuration, an analysis of electromagnetic forces present in normal machine operation on one coil was conducted, this to know the stresses in the encapsulation of the same. Considering the pulsed operation, a thickness of 5 mm is determined in the encapsulated, considering fatigue failure based on studies of fatigue failures in epoxy resins. (Author)

  16. Studies into laser ignition of confined pyrotechnics

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, S.R.; Russell, D.A. [Centre for Applied Laser Spectroscopy, DASSR, Defence Academy, Cranfield University, Shrivenham, Swindon (United Kingdom)

    2008-10-15

    Ignition tests were carried out on three different pyrotechnics using laser energy from the multimode output from an Ar-Ion laser (av) at 500 nm and a near-IR diode laser pigtailed to a fibre optic cable and operating at 808 nm. The pyrotechnics investigated were: G20 black powder, SR44 and SR371C. The confined ignition tests were conducted in a specially designed ignition chamber. Pyrotechnics were ignited by a free space beam entering the chamber through an industrial sapphire window in the case of the Ar-ion laser. For the NIR diode laser, fibre was ducted through a block into direct contact with the pyrotechnic. The Ar-Ion laser was chosen as this was found to ignite all three pyrotechnics in the unconfined condition. It also allowed for a direct comparison of confined/unconfined results to be made. The threshold laser flux densities to initiate reproducible ignitions at this wavelength were found to be between {proportional_to}12.7 and {proportional_to}0.16 kW cm{sup -2}. Plotted on the ignition maps are the laser flux densities versus the start of ignition times for the three confined pyrotechnics. It was found from these maps that the times for confined ignition were substantially lower than those obtained for unconfined ignition under similar experimental conditions. For the NIR diode laser flux densities varied between {proportional_to}6.8 and {proportional_to}0.2 kW cm{sup -2}. The minimum ignition times for the NIR diode laser for SR371C ({proportional_to}11.2 ms) and G20 ({proportional_to}17.1 ms) were faster than those achieved by the use of the Ar-ion laser. However, the minimum ignition time was shorter ({proportional_to}11.7 ms) with the Ar-ion laser for SR44. (Abstract Copyright [2008], Wiley Periodicals, Inc.)

  17. A new metric of the low-mode asymmetry for ignition target designs

    International Nuclear Information System (INIS)

    Gu, Jianfa; Dai, Zhensheng; Fan, Zhengfeng; Zou, Shiyang; Ye, Wenhua; Pei, Wenbing; Zhu, Shaoping

    2014-01-01

    In the deuterium-tritium inertial confinement fusion implosion experiments on the National Ignition Facility, the measured neutron yield and hot spot pressure are significantly lower than simulations. Understanding the underlying physics of the deficit is essential to achieving ignition. This paper investigates the low-mode areal density asymmetry in the main fuel of ignition capsule. It is shown that the areal density asymmetry breaks up the compressed shell and significantly reduces the conversion of implosion kinetic energy to hot spot internal energy, leading to the calculated hot spot pressure and neutron yield quite close to the experimental data. This indicates that the low-mode shell areal density asymmetry can explain part of the large discrepancy between simulations and experiments. Since only using the hot spot shape term could not adequately characterize the effects of the shell areal density asymmetry on implosion performance, a new metric of the low-mode asymmetry is developed to accurately measure the probability of ignition

  18. Two-Dimensional Simulations of Electron Shock Ignition at the Megajoule Scale

    Science.gov (United States)

    Shang, W.; Betti, R.

    2016-10-01

    Shock ignition uses a late strong shock to ignite the hot spot of an inertial confinement fusion capsule. In the standard shock-ignition scheme, an ignitor shock is launched by the ablation pressure from a spike in laser intensity. Recent experiments on OMEGA have shown that focused beams with intensity up to 6 ×1015 W /cm2 can produce copious amounts of hot electrons. The hot electrons are produced by laser-plasma instabilities (LPI's) and can carry up to 15 % of the instantaneous laser power. Megajoule-scale targets will likely produce even more hot electrons because of the large plasma scale length. We show that it is possible to design ignition targets with low implosion velocities that can be shock ignited using LPI-generated hot electrons to obtain high energy gains. These designs are robust to low-mode asymmetries and they ignite even for highly distorted implosions. Electron shock ignition requires tens of kilojoules of hot electrons, which can only be produced on a large laser facility like the National Ignition Facility. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  19. Shock Tube Ignition Delay Data Affected by Localized Ignition Phenomena

    KAUST Repository

    Javed, Tamour

    2016-12-29

    Shock tubes have conventionally been used for measuring high-temperature ignition delay times ~ O(1 ms). In the last decade or so, the operating regime of shock tubes has been extended to lower temperatures by accessing longer observation times. Such measurements may potentially be affected by some non-ideal phenomena. The purpose of this work is to measure long ignition delay times for fuels exhibiting negative temperature coefficient (NTC) and to assess the impact of shock tube non-idealities on ignition delay data. Ignition delay times of n-heptane and n-hexane were measured over the temperature range of 650 – 1250 K and pressures near 1.5 atm. Driver gas tailoring and long length of shock tube driver section were utilized to measure ignition delay times as long as 32 ms. Measured ignition delay times agree with chemical kinetic models at high (> 1100 K) and low (< 700 K) temperatures. In the intermediate temperature range (700 – 1100 K), however, significant discrepancies are observed between the measurements and homogeneous ignition delay simulations. It is postulated, based on experimental observations, that localized ignition kernels could affect the ignition delay times at the intermediate temperatures, which lead to compression (and heating) of the bulk gas and result in expediting the overall ignition event. The postulate is validated through simple representative computational fluid dynamic simulations of post-shock gas mixtures which exhibit ignition advancement via a hot spot. The results of the current work show that ignition delay times measured by shock tubes may be affected by non-ideal phenomena for certain conditions of temperature, pressure and fuel reactivity. Care must, therefore, be exercised in using such data for chemical kinetic model development and validation.

  20. Design study of superconducting inductive energy storages for tokamak fusion reactor

    International Nuclear Information System (INIS)

    1977-08-01

    Design of the superconducting inductive energy storages (SC-IES) has been studied. One SC-IES is for the power supply system in a experimental tokamak fusion reactor, and the other in a future practical reactor. Study started with definition of the requirements of SC-IES, followed by optimization of the coil shape and determination of major parameters. Then, the coil and the vessel were designed, including the following: for SC-IES of the experimental reactor, stored energy 10 GJ, B max 8 T, conductor NbTi and size 18 m diameter x 10 m height; for SC-IES of the practical reactor, stored energy 56 GJ, B max 10.5 T, conductor Nb 3 Sn and size 26 m diameter x 15 m height. Design of the coil protection system and an outline of the auxiliary systems (for refrigeration and evacuation) are also given, and further, problems and usefullness of SC-IES. (auth.)

  1. Progress towards polar-drive ignition for the NIF

    International Nuclear Information System (INIS)

    McCrory, R.L.; Betti, R.; Boehly, T.R.; Collins, T.J.B.; Craxton, R.S.; Delettrez, J.A.; Edgell, D.H.; Epstein, R.; Froula, D.H.; Glebov, V.Yu.; Goncharov, V.N.; Harding, D.R.; Hohenberger, M.; Hu, S.X.; Igumenshchev, I.V.; Kessler, T.J.; Knauer, J.P.; Casey, D.T.; Frenje, J.A.; Gatu-Johnson, M.

    2013-01-01

    The University of Rochester's Laboratory for Laser Energetics (LLE) performs direct-drive inertial confinement fusion (ICF) research. LLE's Omega Laser Facility is used to study direct-drive ICF ignition concepts, developing an understanding of the underlying physics that feeds into the design of ignition targets for the National Ignition Facility (NIF). The baseline symmetric-illumination, direct-drive–ignition target design consists of a 1.5 MJ multiple-picket laser pulse that generates four shock waves (similar to the NIF baseline indirect-drive design) and is predicted to produce a one-dimensional (1D) gain of 48. LLE has developed the polar-drive (PD) illumination concept (for NIF beams in the x-ray–drive configuration) to allow the pursuit of direct-drive ignition without significant reconfiguration of the beam paths on the NIF. Some less-invasive changes in the NIF infrastructure will be required, including new phase plates, polarization rotators, and a PD-specific beam-smoothing front end. A suite of PD ignition designs with implosion velocities from 3.5 to 4.3 × 10 7 cm s −1 are predicted to have significant 2D gains (Collins et al 2012 Bull. Am. Phys. Soc. 57 155). Verification of the physics basis of these simulations is a major thrust of direct-drive implosion experiments on both OMEGA and the NIF. Many physics issues are being examined with symmetric beam irradiation on OMEGA, varying the implosion parameters over a wide region of design space. Cryogenic deuterium–tritium target experiments with symmetric irradiation have produced areal densities of ∼0.3 g cm −2 , ion temperatures over 3 keV, and neutron yields in excess of 20% of the ‘clean’ 1D predicted value. The inferred Lawson criterion figure of merit (Betti R. et al 2010 Phys. Plasmas 17 058102) has increased from 1.7 atm s (IAEA 2010) to 2.6 atm s. (paper)

  2. High-frequency gyrotrons and their application to tokamak plasma heating

    International Nuclear Information System (INIS)

    Kreischer, K.E.

    1981-01-01

    A comprehensive analysis of high frequency (100 to 200 GHz) and high power (> 100 kW) gyrotrons has been conducted. It is shown that high frequencies will be required in order for electron cyclotron radiation to propagate to the center of a compact tokamak power reactor. High power levels will be needed in order to ignite the plasma with a reasonable number of gyrotron units. In the first part of this research, a set of analytic expressions, valid for all TE cavity modes and all harmonics, is derived for the starting current and frequency detuning using the Vlasov-Maxwell equations in the weakly relativistic limit. The use of an optical cavity is also investigated

  3. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  4. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

    New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

  5. National Ignition Facility (NIF) Control Network Design and Analysis

    International Nuclear Information System (INIS)

    Bryant, R M; Carey, R W; Claybourn, R V; Pavel, G; Schaefer, W J

    2001-01-01

    The control network for the National Ignition Facility (NIF) is designed to meet the needs for common object request broker architecture (CORBA) inter-process communication, multicast video transport, device triggering, and general TCP/IP communication within the NIF facility. The network will interconnect approximately 650 systems, including the embedded controllers, front-end processors (FEPs), supervisory systems, and centralized servers involved in operation of the NIF. All systems are networked with Ethernet to serve the majority of communication needs, and asynchronous transfer mode (ATM) is used to transport multicast video and synchronization triggers. CORBA software infra-structure provides location-independent communication services over TCP/IP between the application processes in the 15 supervisory and 300 FEP systems. Video images sampled from 500 video cameras at a 10-Hz frame rate will be multicast using direct ATM Application Programming Interface (API) communication from video FEPs to any selected operator console. The Ethernet and ATM control networks are used to broadcast two types of device triggers for last-second functions in a large number of FEPs, thus eliminating the need for a separate infrastructure for these functions. Analysis, design, modeling, and testing of the NIF network has been performed to provide confidence that the network design will meet NIF control requirements

  6. Capsule performance optimization in the National Ignition Campaigna)

    Science.gov (United States)

    Landen, O. L.; Boehly, T. R.; Bradley, D. K.; Braun, D. G.; Callahan, D. A.; Celliers, P. M.; Collins, G. W.; Dewald, E. L.; Divol, L.; Glenzer, S. H.; Hamza, A.; Hicks, D. G.; Hoffman, N.; Izumi, N.; Jones, O. S.; Kirkwood, R. K.; Kyrala, G. A.; Michel, P.; Milovich, J.; Munro, D. H.; Nikroo, A.; Olson, R. E.; Robey, H. F.; Spears, B. K.; Thomas, C. A.; Weber, S. V.; Wilson, D. C.; Marinak, M. M.; Suter, L. J.; Hammel, B. A.; Meyerhofer, D. D.; Atherton, J.; Edwards, J.; Haan, S. W.; Lindl, J. D.; MacGowan, B. J.; Moses, E. I.

    2010-05-01

    A capsule performance optimization campaign will be conducted at the National Ignition Facility [G. H. Miller et al., Nucl. Fusion 44, 228 (2004)] to substantially increase the probability of ignition by laser-driven hohlraums [J. D. Lindl et al., Phys. Plasmas 11, 339 (2004)]. The campaign will experimentally correct for residual uncertainties in the implosion and hohlraum physics used in our radiation-hydrodynamic computational models before proceeding to cryogenic-layered implosions and ignition attempts. The required tuning techniques using a variety of ignition capsule surrogates have been demonstrated at the OMEGA facility under scaled hohlraum and capsule conditions relevant to the ignition design and shown to meet the required sensitivity and accuracy. In addition, a roll-up of all expected random and systematic uncertainties in setting the key ignition laser and target parameters due to residual measurement, calibration, cross-coupling, surrogacy, and scale-up errors has been derived that meets the required budget.

  7. Capsule performance optimization in the National Ignition Campaign

    International Nuclear Information System (INIS)

    Landen, O. L.; Bradley, D. K.; Braun, D. G.; Callahan, D. A.; Celliers, P. M.; Collins, G. W.; Dewald, E. L.; Divol, L.; Glenzer, S. H.; Hamza, A.; Hicks, D. G.; Izumi, N.; Jones, O. S.; Kirkwood, R. K.; Michel, P.; Milovich, J.; Munro, D. H.; Robey, H. F.; Spears, B. K.; Thomas, C. A.

    2010-01-01

    A capsule performance optimization campaign will be conducted at the National Ignition Facility [G. H. Miller et al., Nucl. Fusion 44, 228 (2004)] to substantially increase the probability of ignition by laser-driven hohlraums [J. D. Lindl et al., Phys. Plasmas 11, 339 (2004)]. The campaign will experimentally correct for residual uncertainties in the implosion and hohlraum physics used in our radiation-hydrodynamic computational models before proceeding to cryogenic-layered implosions and ignition attempts. The required tuning techniques using a variety of ignition capsule surrogates have been demonstrated at the OMEGA facility under scaled hohlraum and capsule conditions relevant to the ignition design and shown to meet the required sensitivity and accuracy. In addition, a roll-up of all expected random and systematic uncertainties in setting the key ignition laser and target parameters due to residual measurement, calibration, cross-coupling, surrogacy, and scale-up errors has been derived that meets the required budget.

  8. Hard X-ray studies on the Castor tokamak

    International Nuclear Information System (INIS)

    Mlynar, J.

    1990-04-01

    The electron runaway processes in tokamaks are discussed with regard to hard X radiation measurements. The origin and confinement of runaway electrons, their bremsstrahlung spectra and the influence of lower hybrid current drive on the distribution of high-energy electrons are analyzed for the case of the Castor tokamak. The hard X-ray spectrometer designed for the Castor tokamak is also described and preliminary qualitative results of hard X-ray measurements are presented. The first series of integral measurements made it possible to map the azimuthal dependence of the hard X radiation

  9. TIBER-II TF [toroidal-field] winding pack design

    International Nuclear Information System (INIS)

    Kerns, J.A.; Miller, J.R.; Slack, D.S.; Summers, L.T.

    1987-01-01

    The superconducting, toroidal-field (TF) coils in the Tokamak Ignition/Burn Engineering Reactor (TIBER II) are designed with cable-in-conduit conductor (CICC) using Nb 3 Sn composite strands. To design the CICC winding pack, we used an optimization technique that maximizes the conductor stability without violating the constraints imposed by the structure, electrical insulation, quench protection, and fabrication technique. Detailed helium-properties codes calculate the heat removal along a flow path, and detailed field calculations determine the temperature, current, and stability margins. The conductor sheath is designed as distributed structure to partially support the combined in-plane and out-of-plane loads generated within the winding pack. Pancakes of the coil are wound, reacted, and insulated before being potted in the case. This design is aggressive but fully consistent with good engineering practice. 5 refs., 4 figs., 2 tabs

  10. A conceptual design of superconducting spherical tokamak reactor

    International Nuclear Information System (INIS)

    Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi

    2012-01-01

    This paper presents a fusion reactor concept named 'JUST (Japanese Universities' Super Tokamak reactor)'. From the plasma confinement system to the power generation system is evaluated in this work. JUST design has features as follows: the superconducting magnet, the steady state operation with high bootstrap current fraction, the easy replacement of neutron damaged first wall, the high heat flux in the divertor, and the low cost (or high β). By winding the OH solenoid over the center stack of toroidal field coil, we have the low aspect ratio and the 80cm thick neutron shield to protect the superconducting center stack. JUST is designed by using the 0-D transport code under the assumption that the energy confinement time is 1.8 times of the IPB98(y,2) scaling. Main parameters are as follows: the major radius of 4.5m, the aspect ratio of 1.8, the elongation ratio of 2.5, the toroidal field of 2.36T, the plasma current of 18MA, the toroidal beta of 22%, the central electron and ion temperature of 15keV and the fusion thermal power of 2.4GW. By using the mercury heat exchanger and the steam turbine, the heat efficiency is 33% and the electric power is 0.74GW. (author)

  11. Designs for highly nonlinear ablative Rayleigh-Taylor experiments on the National Ignition Facility

    International Nuclear Information System (INIS)

    Casner, A.; Masse, L.; Liberatore, S.; Jacquet, L.; Loiseau, P.; Poujade, O.; Smalyuk, V. A.; Bradley, D. K.; Park, H. S.; Remington, B. A.; Igumenshchev, I.; Chicanne, C.

    2012-01-01

    We present two designs relevant to ablative Rayleigh-Taylor instability in transition from weakly nonlinear to highly nonlinear regimes at the National Ignition Facility [E. I. Moses, J. Phys.: Conf. Ser. 112, 012003 (2008)]. The sensitivity of nonlinear Rayleigh-Taylor instability physics to ablation velocity is addressed with targets driven by indirect drive, with stronger ablative stabilization, and by direct drive, with weaker ablative stabilization. The indirect drive design demonstrates the potential to reach a two-dimensional bubble-merger regime with a 20 ns duration drive at moderate radiation temperature. The direct drive design achieves a 3 to 5 times increased acceleration distance for the sample in comparison to previous experiments allowing at least 2 more bubble generations when starting from a three-dimensional broadband spectrum.

  12. Designs for highly nonlinear ablative Rayleigh-Taylor experiments on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Casner, A.; Masse, L.; Liberatore, S.; Jacquet, L.; Loiseau, P.; Poujade, O. [CEA, DAM, DIF, F-91297 Arpajon (France); Smalyuk, V. A.; Bradley, D. K.; Park, H. S.; Remington, B. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Igumenshchev, I. [Laboratory of Laser Energetics, University of Rochester, Rochester, New York 14623-1299 (United States); Chicanne, C. [CEA, DAM, VALDUC, F-21120 Is-sur-Tille (France)

    2012-08-15

    We present two designs relevant to ablative Rayleigh-Taylor instability in transition from weakly nonlinear to highly nonlinear regimes at the National Ignition Facility [E. I. Moses, J. Phys.: Conf. Ser. 112, 012003 (2008)]. The sensitivity of nonlinear Rayleigh-Taylor instability physics to ablation velocity is addressed with targets driven by indirect drive, with stronger ablative stabilization, and by direct drive, with weaker ablative stabilization. The indirect drive design demonstrates the potential to reach a two-dimensional bubble-merger regime with a 20 ns duration drive at moderate radiation temperature. The direct drive design achieves a 3 to 5 times increased acceleration distance for the sample in comparison to previous experiments allowing at least 2 more bubble generations when starting from a three-dimensional broadband spectrum.

  13. Study of ignition in a high compression ratio SI (spark ignition) methanol engine using LES (large eddy simulation) with detailed chemical kinetics

    International Nuclear Information System (INIS)

    Zhen, Xudong; Wang, Yang

    2013-01-01

    Methanol has been recently used as an alternative to conventional fuels for internal combustion engines in order to satisfy some environmental and economical concerns. In this paper, the ignition in a high compression ratio SI (spark ignition) methanol engine was studied by using LES (large eddy simulation) with detailed chemical kinetics. A 21-species, 84-reaction methanol mechanism was adopted to simulate the auto-ignition process of the methanol/air mixture. The MIT (minimum ignition temperature) and MIE (minimum ignition energy) are two important properties for designing safety standards and understanding the ignition process of combustible mixtures. The effects of the flame kernel size, flame kernel temperature and equivalence ratio were also examined on MIT, MIE and IDP (ignition delay period). The methanol mechanism was validated by experimental test. The simulated results showed that the flame kernel size, temperature and energy dramatically affected the values of the MIT, MIE and IDP for a methanol/air mixture, the value of the ignition delay period was not only related to the flame kernel energy, but also to the flame kernel temperature. - Highlights: • We used LES (large eddy simulation) coupled with detailed chemical kinetics to simulate methanol ignition. • The flame kernel size and temperature affected the minimum ignition temperature. • The flame kernel temperature and energy affected the ignition delay period. • The equivalence ratio of methanol–air mixture affected the ignition delay period

  14. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  15. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  16. The National Ignition Facility and Industry

    Science.gov (United States)

    Harri, J. G.; Paisner, J. A.; Lowdermilk, W. H.; Boyes, J. D.; Kumpan, S. A.; Sorem, M. S.

    1994-09-01

    The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. The National Ignition Facility construction project will require the best of our construction industries and its success will depend on the best products offered by hundreds of the nation's high technology companies. Three-fourths of the construction costs will be invested in industry. This article reviews the design, cost and schedule, and required industrial involvement associated with the construction project.

  17. The National Ignition Facility and industry

    International Nuclear Information System (INIS)

    Harri, J.G.; Lowdermilk, W.H.; Paisner, J.A.; Boyes, J.D.; Kumpan, S.A.; Sorem, M.S.

    1994-01-01

    The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. The National Ignition Facility construction project will require the best of national construction industries and its success will depend on the best products offered by hundreds of the nation's high technology companies. Three-fourths of the construction costs will be invested in industry. This article reviews the design, cost and schedule, and required industrial involvement associated with the construction project

  18. Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2000-01-01

    SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

  19. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  20. Tokamak Fusion Core Experiment maintenance study

    International Nuclear Information System (INIS)

    Snyder, A.M.; Watts, K.D.

    1985-01-01

    The recently completed Tokamak Fusion Core Experiment (TFCX) design project was carried out to investigate potential next generation tokamak concepts. An important aspect of this project was the early development and incorporation of remote maintainability throughout the design process. This early coordination and incorporation of maintenance aspects to the design of the device and facilities would assure that the machine could ultimately be maintained and repaired in an efficient and cost effective manner. To meet this end, a rigorously formatted engineering trade study was performed to determine the preferred configuration for the TFCX reactor based primarily on maintenance requirements. The study indicated that the preferred design was one with an external vacuum vessel and torrodial field coils that could be removed via a simple radial motion. The trade study is presented and the preferred TFCX configuration is described