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Sample records for compact tokamak reactors

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

    SciTech Connect

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

    1996-09-13

    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.

  2. The ignition physics study group supports the compact ignition tokamak and engineering test reactor programs

    SciTech Connect

    Sheffield, J.

    1987-01-01

    This report presents a collection of Vugraphs dealing with the Compact Ignition Tokamak (CIT) and the Engineering Test Reactor (ETR). The role of the Ignition Physics Study Group is defined. Several design goals are presented. (JDH)

  3. First wall and blanket design for a high wall loading compact tokamak power reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Abdel-Khalik, S.I.; Corradini, M.L.; El-Afify, M.; Huh, K.Y.; Kuleinski, G.L.; Wittenberg, L.J.

    1985-07-01

    Among the specific limitations which tend to complicate a compact high wall loading (HWL) tokamak reactor design are high surface and nuclear heating, compactness leading to crowded components, unlikely breeding on the inboard side and frequent first wall/blanket replacement. This paper describes the mechanical, thermal hydraulic and tritium aspects of an improved blanket design for a high ..beta.. (20%), high wall loading (R 10 MW/m/sup 2/) compact fusion power reactor of 1000 MW /sub th/ power output.

  4. The conceptual design of a robust, compact, modular tokamak reactor based on high-field superconductors

    NASA Astrophysics Data System (ADS)

    Whyte, D. G.; Bonoli, P.; Barnard, H.; Haakonsen, C.; Hartwig, Z.; Kasten, C.; Palmer, T.; Sung, C.; Sutherland, D.; Bromberg, L.; Mangiarotti, F.; Goh, J.; Sorbom, B.; Sierchio, J.; Ball, J.; Greenwald, M.; Olynyk, G.; Minervini, J.

    2012-10-01

    Two of the greatest challenges to tokamak reactors are 1) large single-unit cost of each reactor's construction and 2) their susceptibility to disruptions from operation at or above operational limits. We present an attractive tokamak reactor design that substantially lessens these issues by exploiting recent advancements in superconductor (SC) tapes allowing peak field on SC coil > 20 Tesla. A R˜3.3 m, B˜9.2 T, ˜ 500 MW fusion power tokamak provides high fusion gain while avoiding all disruptive operating boundaries (no-wall beta, kink, and density limits). Robust steady-state core scenarios are obtained by exploiting the synergy of high field, compact size and ideal efficiency current drive using high-field side launch of Lower Hybrid waves. The design features a completely modular replacement of internal solid components enabled by the demountability of the coils/tapes and the use of an immersion liquid blanket. This modularity opens up the possibility of using the device as a nuclear component test facility.

  5. Activation analysis of the compact ignition tokamak

    SciTech Connect

    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. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  7. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Hoffman, H. A.; Logan, B. G.; Campbell, R. B.

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boiling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection temperatures, and only a relatively small natural-draft heat exchanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

  8. On the survivability of diagnostic windows in the CIT (Compact Ignition Tokamak) reactor

    SciTech Connect

    Taylor, A.

    1988-11-01

    The problem of radiation induced stresses in CIT diagnostic windows is discussed. Existing data indicate windows of existing design will probably survive if placed on the periphery of the reactor. There is a lack of adequate engineering data upon which the design and survivability of windows can be based. 22 refs., 5 figs., 2 tabs.

  9. Physics of compact ignition tokamak designs

    SciTech Connect

    Singer, C.E.; Ku, L.P.; Bateman, G.; Seidl, F.; Sugihara, M.

    1986-03-01

    Models for predicting plasma performance in compact ignition experiments are constructed on the basis of theoretical and empirical constraints and data from tokamak experiments. Emphasis is placed on finding transport and confinement models which reproduce results of both ohmically and auxiliary heated tokamak data. Illustrations of the application of the models to compact ignition designs are given.

  10. System studies of compact ignition tokamaks

    SciTech Connect

    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.

  11. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    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{sup 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.

  12. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    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.

  13. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

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

    NASA Astrophysics Data System (ADS)

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

    2009-05-01

    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 CO2 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. CO2 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.

  15. System studies of Compact Ignition Tokamaks

    SciTech Connect

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

    1986-11-01

    The Fusion Engineering Design Center (FEDC) Tokamak Systems Code is used to perform trade studies in accordance with the Compact Ignition Tokamak (CIT) physics and engineering guidelines. The authors examine various toroidal field coil (TFC) configurations, preload levels, and coil materials. Use of Inconel-copper composite material results in the smallest sized devices for both bucked and wedged TFCs and wedged-only TFCs. Preload levels of 23 Mkg are needed for the minimum sized devices, and for the lower strength materials, the minimum size is sensitive to the preload level. Results from these trade studies help lead to the choice of the baseline CIT point at R = 1.25 m and B = 10.4 T.

  16. Plasma transport in a compact ignition tokamak

    SciTech Connect

    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 /sup 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.

  17. Physics evaluation of compact tokamak ignition experiments

    SciTech Connect

    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/S/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.

  18. Cooldown of the Compact Ignition Tokamak

    SciTech Connect

    Keeton, D.C.

    1987-08-01

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.

  19. Overview of the compact ignition tokamak

    SciTech Connect

    Flanagan, C.A.

    1986-01-01

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

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

    SciTech Connect

    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.

  1. Plasma diagnostics for the compact ignition tokamak

    SciTech Connect

    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/sup 21/m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs.

  2. Physics aspects of the Compact Ignition Tokamak

    SciTech Connect

    Post, D.; Bateman, G.; Houlberg, W.; Bromberg, L.; Cohn, D.; Colestock, P.; Hughes, M.; Ignat, D.; Izzo, R.; Jardin, S.

    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/ approx. 2 x 10/sup 20/ sec m/sup -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 approx. 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.

  3. Decommissioning the Tokamak Fusion Test Reactor

    SciTech Connect

    Spampinato, P.T.; Walton, G.R.

    1993-10-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives.

  4. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. 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. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  5. Maintenance concept development for the Compact Ignition Tokamak

    SciTech Connect

    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.

  6. Development of tokamak reactor system analysis code NEW-TORSAC

    NASA Astrophysics Data System (ADS)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  7. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  8. Compact Ignition Tokamak Program: R and D needs

    SciTech Connect

    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.

  9. Time-dependent simulations of a Compact Ignition Tokamak

    SciTech Connect

    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.

  10. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

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

    SciTech Connect

    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.

  12. Magnet design approach for pulsed tokamak reactors

    SciTech Connect

    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/sup 2/ to 10/sup 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/sup 5/; for N approx. 10/sup 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.

  13. Resistive demountable toroidal-field coils for tokamak reactors

    SciTech Connect

    Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

    1981-07-01

    Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments.

  14. Preliminary design of the CIT (Compact Ignition Tokamak) cryostat

    SciTech Connect

    Goins, M.L.

    1989-01-01

    For the Compact Ignition Tokamak (CIT) to achieve the performance goals set forth, the toroidal field (TF) and poloidal field (PF) coil systems must operate in a cryogenic temperature regime. The cryostat has been designed to provide and maintain this environment. The preliminary design activity is addressing the design issues and interfaces necessary to provide a cryogenic vessel that will maintain a maximum temperature differential of 8{degree}C between the outer vessel wall and the ambient test cell conditions; operate in a pressure range of +5 psig to {minus}2 psig; accommodate numerous penetrations, including cooling, diagnostic, and gravity support items; and maintain a maximum leak rate of gaseous nitrogen at 1 l/s at 1 atm. Conceptually, the cryostat consists of thermal insulation sandwiched between an inner primary stainless steel pressure vessel and a thin outer stainless steel wall. Design activities have concentrated on determining the size and shape of the primary vessel wall and selecting the best candidate thermal insulation materials for future irradiation testing. The following shapes of the upper and lower cryostat structure were analyzed: a standard ASME torispherical domed top and bottom; a nonstandard domed top and bottom; and a 2{degree} sloped conical top and bottom contour. Screening of candidate insulation materials was based on lowest thermal conductivity over the range of temperatures anticipated in the CIT environment; low material cost and apparent ease of assembly; and survivability of material in the CIT irradiation environment. This paper presents the configuration development of the cryostat used to maintain the cryogenic temperature environment for CIT. 3 refs., 3 figs., 3 tabs.

  15. Compact reactor/ORC power source

    SciTech Connect

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500/sup 0/C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370/sup 0/C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs.

  16. Beta optimization in the context of reactor relevant tokamaks

    SciTech Connect

    Manickam, J.

    1990-08-01

    In a reactor relevant tokamak the appropriate definition of {beta}, the ratio of the particle and magnetic field pressures, is {beta}* {equivalent to} (2 < p{sup 2} >{sup {1/2}} /B{sup 2}), which exceeds the conventional definition by a factor dependent on the pressure peaking factor, PPF. A simple scaling is obtained which relates the two definitions, {beta}*/{beta}{approx equal}0.9 {plus} 0.15 PPF. Stability properties are determined in terms of {beta}* in a circular and dee-shaped tokamak. 4 refs., 6 figs.

  17. The impact of improved physics on commercial tokamak reactors

    SciTech Connect

    Galambos, J.D.; Perkins, L.J.; Haney, S.; Mandrekas, J.

    1994-01-01

    Improvements in the confinement and beta capability of tokamak devices have long been a goal of the fusion program. We examine the impact of improvements in present day confinement and beta capabilities on commercial tokamak reactors. We characterize confinement with the achievable enhancement factor (H) over the ITER89 Power scaling confinement time, and beta by the Troyon coefficient g. A surprisingly narrow range of plasma confinement and beta are found to be useful in minimizing the cost of electricity for a tokamak reactor. Improvements in only one of these quantities is not useful beyond some point, without accompanying improvements in the other. For the plasma beta limited by a Troyon coefficient (g) near 4.3 (%mT/MA), confinement levels characterized by H factor enhancements of only 2 are useful for our nominal steady-state driven tokamak. These confinement levels are similar to those observed in present day experiments. If the permissible Troyon beta coefficient is near 6, the useful H factor confinement range increases to 2.5, still close to present day confinement levels. Inductively driven, pulsed reactors have somewhat increased useful ranges of confinement, relative to the steady-state cases. For a Troyon beta limit coefficient g near 4.3, H factors up to 2.5 are useful, and for g near 6, H factors up to 3 are useful.

  18. Materials needs for compact fusion reactors

    SciTech Connect

    Krakowski, R.A.

    1983-01-01

    The economic prospects for magnetic fusion energy can be dramatically improved if for the same total power output the fusion neutron first-wall (FW) loading and the system power density can be increased by factors of 3 to 5 and 10 to 30, respectively. A number of compact fusion reactor embodiments have been proposed, all of which would operate with increased FW loadings, would use thin (0.5 to 0.6 m) blankets, and would confine quasi-steady-state plasma with resistive, water-cooled copper or aluminum coils. Increased system power density (5 to 15 MWt/m/sup 3/ versus 0.3 to 0.5 MW/m/sup 3/), considerably reduced physical size of the fusion power core (FPC), and appreciably reduced economic leverage exerted by the FPC and associated physics result. The unique materials requirements anticipated for these compact reactors are outlined against the well documented backdrop provided by similar needs for the mainline approaches. Surprisingly, no single materials need that is unique to the compact systems is identified; crucial uncertainties for the compact approaches must also be addressed by the mainline approaches, particularly for in-vacuum components (FWs, limiters, divertors, etc.).

  19. Tritium experience in the Tokamak Fusion Test Reactor

    SciTech Connect

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N.; Hogan, J.

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  20. A Compact Quasi-axisymmetric Stellarator Reactor

    SciTech Connect

    L.P. Ku; the ARIES-CS Team

    2003-10-20

    We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils.

  1. Dust divertor for a tokamak fusion reactor

    SciTech Connect

    Tang, X Z; Delzanno, G L

    2009-01-01

    Micron-size tungsten particulates find equilibrium position in the magnetized plasma sheath in the normal direction of the divertor surface, but are convected poloidally and toroidally by the sonic-ion-flow drag parallel to the divertor surface. The natural circulation of dust particles in the magnetized plasma sheath can be used to set up a flowing dust shield that absorbs and exhausts most of the tokamak heat flux to the divertor. The size of the particulates and the choice of materials offer substantial room for optimization.

  2. WILDCAT: a catalyzed D-D tokamak reactor

    SciTech Connect

    Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

    1981-11-01

    WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.

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

    SciTech Connect

    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.

  4. Plasma Physics Regimes in Tokamaks with Li Walls

    SciTech Connect

    L.E. Zakharo; N.N. Gorelenkov; R.B. White; S.I. Krasheninnikov; G.V. Pereverzev

    2003-08-21

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors.

  5. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    SciTech Connect

    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.

  6. Preconceptual design and assessment of a Tokamak hybrid reactor

    NASA Astrophysics Data System (ADS)

    Teofilo, V. L.; Leonard, B. R., Jr.; Aase, D. T.; Bickford, W. E.; McCormick, N. J.; McGrath, R. T.; Morrison, J. E.; Perry, R. T.; Schulte, S. C.; Willingham, C. E.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant was 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 Nb3Sn 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 were made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis was 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.

  7. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    E. Perry; J. Chrzanowski; K. Rule; M. Viola; M. Williams; R. Strykowsky

    1999-11-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling.

  8. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    SciTech Connect

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

  9. DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR

    SciTech Connect

    Rule, Keith; Perry, Erik; Parsells, Robert

    2003-02-27

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

  10. Tangential and Vertical Compact Torus Injection Experiments on the STOR-M Tokamak

    NASA Astrophysics Data System (ADS)

    Xiao, Chijin; D, Liu; S, Livingstone; A, K. Singh; E, Zhang; A, Hirose

    2005-04-01

    This paper describes the setup and results of compact torus (CT) injection experiments on the STOR-M tokamak. Tangential CT injection into STOR-M induced H-mode-like phenomena including doubling the electron density, reduction in the Hα radiation level, suppression of the floating potential fluctuations, suppression of the m = 2 Mirnov oscillations, and increase in the global energy confinement time. Experimental setup, bench-test results, and some preliminary injection data for vertical CT injection experiments on STOR-M will be shown. In addition, numerical simulations of the CT trajectories in tokamak discharges for both tangential and vertical injection geometries will be discussed.

  11. Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas

    NASA Astrophysics Data System (ADS)

    Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira

    2010-11-01

    As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.

  12. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  13. Neutron Dosimetry Tokamak Fusion Test Reactor Lithium Blanket Module

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderl, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket Module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment.

  14. Operation of a tokamak reactor in the radiative improved mode

    NASA Astrophysics Data System (ADS)

    Morozov, D. Kh.; Mavrin, A. A.

    2016-03-01

    The operation of a nuclear fusion reactor has been simulated within a model based on experimental results obtained at the TEXTOR-94 tokamak and other facilities in which quasistationary regimes were achieved with long confinement times, high densities, and absence of the edge-localized mode. The radiative improved mode of confinement studied in detail at the TEXTOR-94 tokamak is the most interesting such regime. One of the most important problems of modern tokamaks is the problem of a very high thermal load on a divertor (or a limiter). This problem is quite easily solved in the radiative improved mode. Since a significant fraction of the thermal energy is reemitted by an impurity, the thermal loading is significantly reduced. As the energy confinement time τ E at high densities in the indicated mode is significantly larger than the time predicted by the scaling of ITERH-98P(y, 2), ignition can be achieved in a facility much smaller than the ITER facility at plasma temperatures below 20 keV. The revealed decrease in the degradation of the confinement time τ E with an increase in the introduced power has been analyzed.

  15. The spheromak as a compact fusion reactor

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.

    1987-03-01

    After summarizing the economic and utility-based rationale for compact, higher-power-density fusion reactors, the gun-sustained spheromak concept is explored as one of a number of poloidal-field-dominated confinement configurations that might improve the prospects for economically attractive and operationally simplified fusion power plants. Using a comprehensive physics/engineering/costing model for the spheromak, guided by realistic engineering constraints and physics extrapolation, a range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported. The results presented herein provide the basis for conceptual engineering designs of key fusion-power-core (FPC) subsystems and more detailed plasma modeling of this promising, high mass-power-density concept, which stresses single-piece FPC maintenance, steady-state current drive through electrostatic magnetic helicity injection, a simplified co-axial electrode-divertor, and efficient resistive-coal equilibrium-field coils. The optimal FPC size and the cost estimates project a system that competes aggressively with the best offered by alternative energy sources while simplifying considerably the complexity that has generally been associated with most approaches to magnetic fusion energy.

  16. Neutron wall loading of Tokamak reactors

    NASA Astrophysics Data System (ADS)

    Wong, C. P. C.

    2000-12-01

    Neutron wall loading (Γn) is a key parameter for the selection of fusion power core component materials. It also impacts the economic, performance, design, safety and environmental aspect of the fusion power plant. This paper reports the determination of the range of Γn for economically competitive fusion power plants based on the analysis that couples the MHD stability physics results to a system design code. Cost of electricity (COE) was selected as the parameter to be minimized. For both normal conducting and superconducting coil options, at thermal efficiency of 46% and at the power output range of 1-2 GW(e) the average neutron wall loading is 4-7 MW/m2. For a given power output, higher thermal efficiency will allow lower Γn. At the above range of Γn, in order to have economical fusion power reactors, for the solid first wall design option, high thermal efficiency of 46% to 57.5% requires the use of alloys like V and W-alloy, respectively. The corresponding COE can be projected to be in the economically competitive range of 62-54.6 mill/kWh.

  17. Demountable Toroidal Field Magnets for Use in a Compact Modular Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Mangiarotti, F. J.; Goh, J.; Takayasu, M.; Bromberg, L.; Minervini, J. V.; Whyte, D.

    2014-05-01

    A concept of demountable toroidal field magnets for a compact fusion reactor is discussed. The magnets generate a magnetic field of 9.2 T on axis, in a 3.3 m major radius tokamak. Subcooled YBCO conductors have a critical current density adequate to provide this large magnetic field, while operating at 20 K reduces thermodynamic cooling cost of the resistive electrical joints. Demountable magnets allow for vertical replacement and maintenance of internal components, potentially reducing cost and time of maintenance when compared to traditional sector maintenance. Preliminary measurements of contact resistance of a demountable YBCO electrical joint between are presented.

  18. Comparing Linear Microinstability of the National Compact Stellarator Expriment and a Shaped Tokamak

    SciTech Connect

    J.A. Baumgaertel, G.W. Hammett and D.R. Mikkelsen

    2012-11-20

    One metric for comparing con nement properties of di erent magnetic fusion energy con gurations is the linear critical gradient of drift wave modes. The critical gradient scale length determines the ratio of the core to pedestal temperature when a plasma is limited to marginal stability in the plasma core. The gyrokinetic turbulence code GS2 was used to calculate critical temperature gradients for the linear, collisionless ion tem- perature gradient (ITG) mode in the National Compact Stellarator Experiment (NCSX) and a prototypical shaped tokamak, based on the pro les of a JET H-mode shot and the stronger shaping of ARIES-AT. While a concern was that the narrow cross section of NCSX at some toroidal locations would result in steep gradients that drive instabilities more easily, it is found that other stabilizing e ects of the stellarator con guration o set this so that the normalized critical gradients for NCSX are competitive with or even better than for the tokamak. For the adiabatic ITG mode, NCSX and the tokamak had similar critical gradients, though beyond marginal stability, NCSX had larger growth rates. However, for the kinetic ITG mode, NCSX had a higher critical gradient and lower growth rates until a/LT ≈ 1:5 a/LT;crit, when it surpassed the tokamak's. A discussion of the results presented with respect to a/LT vs R/LT is included.

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

    SciTech Connect

    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/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 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.

  20. Winding insulation in electromagnetic systems for Tokamak reactor plants

    NASA Astrophysics Data System (ADS)

    Maslov, V. V.; Trubachev, S. G.

    1985-01-01

    Magnetic containment of the plasma in nuclear fusion reactors of the Tokamak type requires electromagnets with insulation which must withstand high temperatures and thermal shocks as well as ionizing radiation in various forms and electric fields, and mechanical loads. Insulation materials to ensure adequate thermophysical and mechanical properties are evaluated, followed by design of insulation systems with satisfactory performance characteristics. Data on neutron fluence energy characteristics and radiation absorption doses during neutron interactions are essential for such an evaluation. Materials considered for insulation in electromagnets with superconductor and cryoresistance windings are glass mica tape with epoxy compound impregnation, glass cloth with epoxy compound impregnation (STE), polyimide glass cloth with adhesive coating (LSNL), glass Textolite with epoxy phenolic binder (STEN), epoxy resin paste with mineral fillers (PE), and polyurethane compound modified by epoxy resin with mineral filler (KPU).

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

    SciTech Connect

    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.

  2. Tritium pellet injector design for tokamak fusion test reactor

    SciTech Connect

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.; Combs, S.K.; Easterly, C.E.; Lunsford, R.V.; Milora, S.L.; Schuresko, D.D.; White, J.A.; Williamson, D.H.

    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.

  3. Ignition in a tokamak reactor with INTOR-like parameters

    SciTech Connect

    Singer, C.E.; Seidl, F.G.P.; Post, D.E.; Rutherford, P.H.

    1980-08-01

    The requirements for ignition in a tokamak reactor with INTOR-like parameters were studied using a one-dimensional transport code. With empirical electron energy diffusivity Chi/sub e/, ignition was obtained with 60 to 75 MW of neutral beam injection at a volume average pressure ratio <..beta..> = 4 to 5% under a variety of conditions. Charging Chi/sub e/ gave ignition at the same <..beta..> if the plasma minor radius varied as a proportional to Chi/sub e/sup 1/2/. The maximum impurity concentration which allows ignition was found to be comparable to that for the much simpler case of a homogeneous plasma with radiative losses only. In long pulse simulations with efficient helium pumping, the maximum toroidal field ripple which allowed ignition was 2.0% (peak to peak) at the plasma edge. Ignition was maintained with over 99% recycling of helium ash using 5% less than maximum ripple.

  4. Conceptual design study of a superconducting spherical tokamak reactor with a self-consistent system analysis code

    NASA Astrophysics Data System (ADS)

    Hong, B. G.; Hwang, Y. S.; Kang, J. S.; Lee, D. W.; Joo, H. G.; Ono, M.

    2011-11-01

    In a spherical tokamak (ST) reactor, the radial build of toroidal field coil and the shield play a key role in determining the size of the reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with a one-dimensional radiation transport code. A conceptual design study of a compact superconducting ST reactor with an aspect ratio of up to 2.0 is conducted and the optimum radial build is identified. It is shown that the use of an improved shielding material and high-temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at a low aspect ratio, and that by using an inboard neutron reflector instead of a breeding blanket, tritium self-sufficiency is possible with an outboard blanket only and thus a compact-sized all superconducting coil ST reactor is viable.

  5. Compact formulas for bounce/transit averaging in axisymmetric tokamak geometry

    SciTech Connect

    Duthoit, F.-X.; Brizard, A. J.; Hahm, T. S.

    2014-12-15

    Compact formulas for bounce and transit orbit averaging of the fluctuation-amplitude eikonal factor in axisymmetric tokamak geometry, which is frequently encountered in bounce-gyrokinetic description of microturbulence, are given in terms of the Jacobi elliptic functions and elliptic integrals. These formulas are readily applicable to the calculation of the neoclassical susceptibility in the framework of modern bounce-gyrokinetic theory. In the long-wavelength limit for axisymmetric electrostatic perturbations, we recover the expression for the Rosenbluth-Hinton residual zonal flow [M. N. Rosenbluth and F. L. Hinton, Phys. Rev. Lett. 80, 724 (1998)] accurately.

  6. Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

    SciTech Connect

    Baker, C.C.

    1980-01-01

    STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described.

  7. Hollow current profile scenarios for advanced tokamak reactor operations

    SciTech Connect

    Gourdain, P.-A.; Leboeuf, J.-N.

    2009-11-15

    Advanced tokamak scenarios are a possible approach to boosting reactor performances. Such schemes usually trigger current holes, a particular magnetohydrodynamics equilibrium where no current or pressure gradients exist in the core of the plasma. While such equilibria have large bootstrap fractions, flat pressure profiles in the plasma core may not be optimal for a reactor. However, moderate modifications of the equilibrium current profile can lead to diamagnetism where most of the pressure gradient is now balanced by poloidal currents and the toroidal magnetic field. In this paper, we consider the properties of diamagnetic current holes, also called ''dual equilibria,'' and demonstrate that fusion throughput can be significantly increased in such scenarios. Their stability is investigated using the DCON code. Plasmas with a beta peak of 30% and an average beta of 6% are found stable to both fixed and free-boundary modes with toroidal mode numbers n=1-4, as well as Mercier and high-n ballooning modes. This is not surprising as these scenarios have a normal beta close to 3.

  8. Current generation by helicons and lower hybrid waves in modern tokamaks and reactors ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect

    Vdovin, V. L.

    2013-02-15

    The innovative concept and 3D full-wave code modeling the off-axis current drive by radio-frequency (RF) waves in large-scale tokamaks, ITER and DEMO, for steady-state operation with high efficiency is proposed. The scheme uses the helicon radiation (fast magnetosonic waves at high (20-40) ion cyclotron frequency harmonics) at frequencies of 500-700 MHz propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by helicons, in conjunction with the bootstrap current, ensure the maintenance of a given value of the total current in the stability margin q(0) {>=} 2 and q(a) {>=} 4, and will help to have regimes with a negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure {beta}{sub N} > 3 (the so-called advanced scenarios) of interest for the commercial reactor. Modeling with full-wave three-dimensional codes PSTELION and STELEC showed flexible control of the current profile in the reactor plasmas of ITER and DEMO, using multiple frequencies, the positions of the antennae and toroidal wave slow down. Also presented are the results of simulations of current generation by helicons in the DIII-D, T-15MD, and JT-60AS tokamaks. Commercially available continuous-wave klystrons of the MW/tube range are promising for commercial stationary fusion reactors. The compact antennae of the waveguide type are proposed, and an example of a possible RF system for today's tokamaks is given. The advantages of the scheme (partially tested at lower frequencies in tokamaks) are a significant decline in the role of parametric instabilities in the plasma periphery, the use of electrically strong resonator-waveguide type antennae, and substantially greater antenna-plasma coupling.

  9. Compact Reversed-Field Pinch Reactors (CRFPR): preliminary engineering considerations

    SciTech Connect

    Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Embrechts, M.J.; Schnurr, N.M.; Battat, M.E.; LaBauve, R.J.; Davidson, J.W.

    1984-08-01

    The unique confinement physics of the Reversed-Field Pinch (RFP) projects to a compact, high-power-density fusion reactor that promises a significant reduction in the cost of electricity. The compact reactor also promises a factor-of-two reduction in the fraction of total cost devoted to the reactor plant equipment (i.e., fusion power core (FPC) plus support systems). In addition to operational and developmental benefits, these physically smaller systems can operate economically over a range of total power output. After giving an extended background and rationale for the compact fusion approaches, key FPC subsystems for the Compact RFP Reactor (CRFPR) are developed, designed, and integrated for a minimum-cost, 1000-MWe(net) system. Both the problems and promise of the compact, high-power-density fusion reactor are quantitatively evaluated on the basis of this conceptual design. The material presented in this report both forms a framework for a broader, more expanded conceptual design as well as suggests directions and emphases for related research and development.

  10. Ion cyclotron transmission spectroscopy in the Tokamak Fusion Test Reactor

    SciTech Connect

    Greene, G.J.

    1993-09-01

    The propagation of waves in the ion cyclotron range of frequencies has been investigated experimentally in the Tokamak Fusion Test Reactor. A small, broadband, radiofrequency (rf) magnetic probe located outside the plasma limiter, at a major radius near that of the plasma center, was excited with a low power, frequency swept source (1--200 MHz). Waves propagating to a distant location were detected with a second, identical probe. The rf transmission spectrum revealed a region of attenuation over a band of frequencies for which the minority fundamental resonance was located between the outer plasma edge and the major radius of the probe location. Distinct, non-overlapping attenuation bands were observed from hydrogen and helium-3 minority species; a distinct tritium band should be observed in future DT experiments. Rapid spectrum acquisition during a helium-3 gas puff experiment showed that the wave attenuation involved the plasma core and was not a surface effect. A model in which the received power varied exponentially with the minority density, averaged over the resonance region, fit the time evolution of the probe signal relatively well. Estimation of a 1-d tunneling parameter from the experimental observations is discussed. Minority concentrations of less than 0.5 % can be resolved with this measurement.

  11. Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Program

    NASA Astrophysics Data System (ADS)

    McGuire, Thomas

    2015-11-01

    The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. An overview of the concept and its diamagnetic, high beta magnetically encapsulated linear ring cusp confinement scheme will be given. The analytical model of the major loss mechanisms and predicted performance will be discussed, along with the major physics challenges. Key features of an operational CFR reactor will be highlighted. The proposed developmental path following the current experimental efforts will be presented. ©2015 Lockheed Martin Corporation. All Rights Reserved.

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

    SciTech Connect

    Haines, J.R.

    1987-01-01

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

  13. Compact retarding field energy analyzer for the tokamak ISTTOK boundary plasma

    SciTech Connect

    Nedzelskiy, I. S.; Silva, C.; Figueiredo, H.; Fernandes, H.; Varandas, C. A. F.

    2006-10-15

    The retarding field energy analyzer (RFEA) remains the more reliable diagnostic to measure the ion temperature in the boundary plasmas of magnetic fusion devices. A compact, simple design and inexpensive RFEA has been developed for investigations on the tokamak ISTTOK. It consists of a stainless steel pinhole (with a diameter of 0.6 mm), three fine nickel grids with a separation of 1 mm, and a collector, all insulated by mica. All the components are placed inside a boron nitride housing with dimensions of 14x14x23 mm{sup 3}. The RFEA has been tested in both ion and electron modes. The conditions of the RFEA operation are discussed, and preliminary measurements of the ion and electron temperature profiles presented.

  14. Plans for the CIT (Compact Ignition Tokamak) instrumentation and control system

    SciTech Connect

    Preckshot, G.G.

    1987-10-07

    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.

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

    SciTech Connect

    Mazzucato, E.; Fidone, I.; Granata, G.

    1987-06-01

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

  16. Fuel systems for compact fast space reactors

    SciTech Connect

    Cox, C.M.; Dutt, D.S.; Karnesky, R.A.

    1983-12-01

    About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom density, operating temperatures, fuel swelling, fission gas release, fuel-cladding compatibility, and consequences of failure. It was concluded that UO/sub 2/ and UN fuels show approximately equal performance potential and that UC fuel has lesser potential. W/Re alloys have performed quite well as cladding materials, and Ta, Nb, and Mo/Re alloys, in conjunction with W diffusion barriers, show good promise. Significant issues to be addressed in the future include high burnup swelling of UN, effects of UO/sub 2/-Li coolant reaction in the event of fuel pin failure, and development of an irradiation performance data base with prototypically configured fuel pins irradiated in a fast neutron flux.

  17. High-energy tritium beams as current drivers in tokamak reactors

    SciTech Connect

    Mikkelsen, D.R.; Grisham, L.R.

    1983-04-01

    The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams.

  18. Alpha particle losses from Tokamak Fusion Test Reactor deuterium-tritium plasmas

    SciTech Connect

    Darrow, D.S.; Zweben, S.J.; Batha, S.

    1996-01-01

    Because alpha particle losses can have a significant influence on tokamak reactor viability, the loss of deuterium-tritium alpha particles from the Tokamak Fusion Test Reactor (TFTR) has been measured under a wide range of conditions. In TFTR, first orbit loss and stochastic toroidal field ripple diffusion are always present. Other losses can arise due to magnetohydrodynamic instabilities or due to waves in the ion cyclotron range of frequencies. No alpha particle losses have yet been seen due to collective instabilities driven by alphas. Ion Bernstein waves can drive large losses of fast ions from TFTR, and details of those losses support one element of the alpha energy channeling scenario.

  19. Analysis on damage to TF coils of a compact reversed shear tokamak CREST

    NASA Astrophysics Data System (ADS)

    Huang, Q.; Zheng, S.; Lu, L.; Zeng, Q.; Hiwatari, R.; Asaoka, Y.; Okano, K.; Ogawa, Y.

    2007-08-01

    CREST is a conceptual tokamak reactor design with high β plasma, high thermal efficiency, competitive cost and water-cooled ferritic steel components. Some of its parameters are similar to those of the ITER advanced mode plasma. In this manuscript, the specific issues and analysis on damage to TF coils of CREST were carried out based on the three-dimensional model of the CREST with the widely used code MCNP/4C and the IAEA latest released FENDL/2.1 data library. Damage to some specific regions of the TF coils near large openings and at the inboard mid-plane are calculated and analyzed. Parameters such as the distributions of nuclear heat density, fast neutron flux, dose rate to the epoxy insulator, and peak displacement dose to Cu conductor for the TF coil near these regions were calculated and analyzed. The shield thicknesses at these regions are optimized.

  20. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    NASA Astrophysics Data System (ADS)

    Sager, G. T.; Wong, C. P. C.; Kapich, D. D.; McDonald, C. F.; Schleicher, R. W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankine and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed.

  1. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  2. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  3. The development of a universal diagnostic probe system for Tokamak fusion test reactor

    NASA Technical Reports Server (NTRS)

    Mastronardi, R.; Cabral, R.; Manos, D.

    1982-01-01

    The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

  4. UCLA program in reactor studies: The ARIES tokamak reactor study. Progress report, December 1, 1990--November 30, 1991

    SciTech Connect

    Not Available

    1991-12-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. 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. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ``modest`` extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  5. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    SciTech Connect

    Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

    1993-11-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility.

  6. Spectroscopic diagnostics of superthermal electrons with high-number harmonic EC radiation in tokamak reactor plasmas

    NASA Astrophysics Data System (ADS)

    Minashin, P. V.; Kukushkin, A. B.

    2015-03-01

    A method of spectroscopic diagnostics of the average perpendicular-to-magnetic-field momentum of the superthermal component of the electron velocity distribution (EVD), based on the high-number-harmonic electron cyclotron (EC) radiation, is suggested for nuclear fusion-reactor plasmas under condition of a strong auxiliary heating (e.g. in tokamak DEMO, a next step after tokamak ITER). The method is based on solving an inverse problem for reconstruction of the EVD in parallel and perpendicular-to-magnetic-field components of electron momentum at high and moderate energies responsible for the emission of the high-number-harmonic EC radiation.

  7. Compact, battery powered, wireless digitizers for in situ data acquisitions in the sino-united spherical tokamak

    NASA Astrophysics Data System (ADS)

    Liu, Yangqing; Tan, Yi; Ke, Rui; Yang, Hao; Wang, Wenhao; Gao, Zhe

    2015-07-01

    Potential isolation and long cable drive are very important in acquiring certain signals from tokamak diagnostics. Compact, battery powered, wireless digitizers for in situ data acquisition have been developed and routinely used in the sino-united spherical tokamak to solve the problems of isolation and long cables. The wireless digitizers utilize the integrated analog to digital converters and the static random access memory of microcontrollers but transfer data wirelessly. They consist of simple and concise circuits but have considerable performances of 12-16 bit in resolution and 500-1000 kS/s in sample rate. Wireless triggering and energy saving are two major challenges of the wireless digitizers. Wireless transceivers in the data link layer are used as trigger and can reduce the trigger jitters to be smaller than 1 μs. In order to reduce the energy consumption, the wireless digitizers are waken only when the tokamak is about to discharge. After discharges, they turn to a periodic checking mode with current consumption smaller than 200 μA. Because of low duty cycle, the wireless digitizers have a battery life of up to four weeks. In general, the wireless digitizers have better performance than normal isolation amplifiers and can greatly simplify the cable connections. They are very suitable for the data acquisition of dangerous and/or susceptible analog signals in tokamaks.

  8. Compact, battery powered, wireless digitizers for in situ data acquisitions in the sino-united spherical tokamak.

    PubMed

    Liu, Yangqing; Tan, Yi; Ke, Rui; Yang, Hao; Wang, Wenhao; Gao, Zhe

    2015-07-01

    Potential isolation and long cable drive are very important in acquiring certain signals from tokamak diagnostics. Compact, battery powered, wireless digitizers for in situ data acquisition have been developed and routinely used in the sino-united spherical tokamak to solve the problems of isolation and long cables. The wireless digitizers utilize the integrated analog to digital converters and the static random access memory of microcontrollers but transfer data wirelessly. They consist of simple and concise circuits but have considerable performances of 12-16 bit in resolution and 500-1000 kS/s in sample rate. Wireless triggering and energy saving are two major challenges of the wireless digitizers. Wireless transceivers in the data link layer are used as trigger and can reduce the trigger jitters to be smaller than 1 μs. In order to reduce the energy consumption, the wireless digitizers are waken only when the tokamak is about to discharge. After discharges, they turn to a periodic checking mode with current consumption smaller than 200 μA. Because of low duty cycle, the wireless digitizers have a battery life of up to four weeks. In general, the wireless digitizers have better performance than normal isolation amplifiers and can greatly simplify the cable connections. They are very suitable for the data acquisition of dangerous and/or susceptible analog signals in tokamaks. PMID:26233380

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

    SciTech Connect

    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.

  10. TRAIL: a Tokamak Rail Gun Limiter for fusion reactors

    SciTech Connect

    Powell, J R; Yu, W S; Fillo, J A; Usher, J L

    1980-01-01

    An attractive new limiter concept is investigated. The Tokamak Rail Gun Limiter (TRAIL) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state-of-the-art (several km/sec). Accelerators injecting pellets at approx. 1 km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  11. Radioactive waste disposal characteristics of candidate tokamak demonstration reactors

    SciTech Connect

    Hoffman, E.A.; Stacey, W.M.; Hertel, N.E.

    1998-08-01

    Results from the current physics, materials and blanket R and D programs are combined with physics and engineering design constraints to characterize candidate tokamak demonstration plant (DEMO) designs. Blanket designs based on the principal structural materials, breeding materials and coolants being developed for the DEMO were adapted from the literature. Neutron flux and activation calculations were performed, and several radioactive waste disposal indices were evaluated, for each design. Of the primary low-activation structural materials under development in the US, it appears that vanadium and ferritic steel alloys, and possibly silicon carbide, could lead to DEMO designs which could satisfy realistic low-level waste (LLW) criteria, provided that impurities can be controlled within plausible limits. Allowable LLW concentrations are established for the limiting alloying and impurity elements. All breeding materials and neutron multipliers considered meet the LLW criterion.

  12. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    SciTech Connect

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  13. Comparative study of pulsed and steady-state tokamak reactor burn cycles

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1984-05-01

    Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles.

  14. Alpha-particle losses in compact torsatron reactors

    SciTech Connect

    Painter, S.L.; Lyon, J.F.

    1989-01-01

    Loss of alpha particles in compact torsatron reactors is studied. For 6, 9, and 12 field period reactors, the direct loss is a relatively weak function of radius and energy and varies from approx. =33% for M = 6 to approx. =18% for M = 12. Loss of alpha particles through scattering into the loss region is calculated using the Fokker-Plank equation for fast ions and found to contribute an additional alpha-particle energy loss of approx. =15%. The consequences of these relatively large losses for torsatron reactor design are discussed. The relationship between the direct particle losses and the magnetic field structure is also studied. Orbit losses from a variety of stellarator configurations are calculated and a figure-of-merit that characterizes the orbit confinement of a magnetic configuration is deduced from these calculations. This figure-of-merit is used to show how the direct losses might be reduced at low aspect-ratio. Effects of finite beta on the direct particle losses are also addressed, and are shown to significantly increase the direct losses in some configurations. 15 refs., 8 figs.

  15. Analysis of measurement errors for Thomson diagnostics of non-Maxwellian plasmas in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Sdvizhenskii, P. A.; Kukushkin, A. B.; Kurskiev, G. S.; Mukhin, E. E.; Bassan, M.

    2016-01-01

    The study is stimulated by the expected noticeable deviation of the electron velocity distribution function (eVDF) from a Maxwellian under condition of a strong auxiliary heating of electron plasmas in tokamak-reactors. The key principles of accuracy estimation of the Thomson scattering diagnostic of non-Maxwellian plasmas in tokamak-reactors are presented. The algorithm extends the conventional approach to the assessment of non-Maxwellian plasmas measurements errors for a broad class of deviations of the eVDF from a Maxwellian. The algorithm is based on solving the inverse problem many times to determine main parameters of the eVDF with allowance for all possible sources of error and statistical variation of the input parameters of the problem. The method is applied to a preliminary analysis of the advantages of the formerly suggested use of various wavelengths of probing laser radiation in the Thomson diagnostics of non-Maxwellian plasma on the example of the core plasma Thomson scattering diagnostic system which is under design for ITER tokamak. The results obtained confirm the relevance of the diversification of the probing laser radiation wavelength.

  16. Feedback stabilization of the resistive shell mode in a tokamak fusion reactor

    SciTech Connect

    Fitzpatrick, R.

    1997-07-01

    Stabilization of the {open_quotes}resistive shell mode{close_quotes} is vital to the success of the {open_quotes}advanced tokamak{close_quotes} concept. The most promising reactor relevant approach is to apply external feedback using, for instance, the previously proposed {open_quotes}fake rotating shell{close_quotes} scheme [R. Fitzpatrick and T. H. Jensen, Phys. Plasmas {bold 3}, 2641 (1996)]. This scheme, like other simple feedback schemes, only works if the feedback controlled conductors are located inside the {open_quotes}critical radius{close_quotes} at which a perfectly conducting shell is just able to stabilize the ideal external kink mode. In general, this is not possible in a reactor, since engineering constraints demand that any feedback controlled conductors be placed {ital outside} the neutron shielding blanket (i.e., relatively far from the edge of the plasma). It is demonstrated that the fake rotating shell feedback scheme can be modified so that it works even when the feedback controlled conductors are located well beyond the critical radius. The gain, bandwidth, current, and total power requirements of such a feedback system for a reactor sized plasma are estimated to be less than 100, a few Hz, a fews tens of kA, and a few MW, respectively. These requirements could easily be met using existing technology. It is concluded that feedback stabilization of the resistive shell mode is possible in a tokamak fusion reactor. {copyright} {ital 1997 American Institute of Physics.}

  17. Model for collisional fast ion diffusion into Tokamak Fusion Test Reactor loss cone

    SciTech Connect

    Chang, C.S. |; Zweben, S.J.; Schivell, J.; Budny, R.; Scott, S.

    1994-08-01

    An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium-tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR). A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the TRANSP code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing-down path due to reduction in banana thickness, which enables us to understand, for the first time. the dependence of the collisional loss rate on Z{sub eff}.

  18. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    SciTech Connect

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  19. Compact and Lightweight Sabatier Reactor for Carbon Dioxide Reduction

    NASA Technical Reports Server (NTRS)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

    2011-01-01

    The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA s cabin Atmosphere Revitalization System and In-Situ Resource Utilization architectures for both low-earth orbit and long-term manned space missions. In the current International Space Station (ISS) and other low orbit missions, metabolically-generated CO2 is removed from the cabin air and vented into space, resulting in a net loss of O2. This requires a continuous resupply of O2 via water electrolysis, and thus highlights the need for large water storage capacity. For long-duration space missions, the amount of life support consumables is limited and resupply options are practically nonexistent, thus atmosphere resource management and recycle becomes crucial to significantly reduce necessary O2 and H2O storage. Additionally, the potential use of the Martian CO2-rich atmosphere and Lunar regolith to generate life support consumables and propellant fuels is of interest to NASA. Precision Combustion, Inc. (PCI) has developed a compact, lightweight Microlith(Registered TradeMark)-based Sabatier (CO2 methanation) reactor which demonstrates the capability of achieving high CO2 conversion and near 100% CH4 selectivity at space velocities of 30,000-60,000 hr-1. The combination of the Microlith(Registered TradeMark) substrates and durable, novel catalyst coating permitted efficient Sabatier reactor operation that favors high reactant conversion, high selectivity, and long-term durability. This paper presents the reactor development and performance results at various operating conditions. Additionally, results from 100-hr durability tests and mechanical vibration tests are discussed.

  20. Conceptual design of a Commercial Tokamak Hybrid Reactor (CTHR)

    NASA Astrophysics Data System (ADS)

    1980-12-01

    This design was developed as a first generation commercial plant for the production of fissile fuel to support a significant number of client light water reactor (LWR) plants. The study was carried out in sufficient depth of indicate no insurmountable technical problems exist, assuming the physics of the fusion driver is verified, and has provided a basis for deriving cost estimates of the hybrid plant as well as estimates of the hybrid/LWR symbiotic system busbar electricity costs. This energy system has the potential to be optimized such that the net cost of electricity becomes competitive with conventional LWR plants as the price of U308 exceeds $100 per pound.

  1. Articulated limiter blade for a tokamak fusion reactor

    DOEpatents

    Doll, David W.

    1985-01-01

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  2. Articulated limiter blade for a tokamak fusion reactor

    DOEpatents

    Doll, D.W.

    1982-10-21

    A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

  3. Tokamak reactor for treating fertile material or waste nuclear by-products

    DOEpatents

    Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

    2012-10-02

    Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

  4. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

  5. Pressure drop considerations of a lithium cooled fusion breeder tokamak reactor blanket

    SciTech Connect

    Wong, C.P.C.

    1983-12-06

    Liquid lithium was selected as one of the coolants for the 1983 fusion breeder blanket used on the magnetically confined tokamak fusion reactor, and as a result, the thermal-hydraulic calculations were dominated by magnetohydrodynamic (MHD) considerations. The applicable sets of MHD equations for the engineering thermal-hydraulic design were reviewed and compared. Special attention was given to the MHD calculations for the fertile material zone, a packed bed of composite beryllium and thorium balls, since this region can dominate the thermal-hydraulic behavior of this blanket module. To keep the pressure drops acceptable, fertile fuel balls were omitted in the inboard blanket.

  6. Images of plasma disruption effects in the Tokamak Fusion Test Reactor

    SciTech Connect

    Maqueda, R.J.; Wurden, G.A.

    1999-02-01

    Fast-framing imaging of visible radiation from magnetically confined plasmas has lately become a useful tool for both machine operation and physics studies. Using an intensified, commercial Kodak Ektapro imaging system, the effects of a plasma disruption were observed in the Tokamak Fusion Test Reactor (TFTR). The high-energy runaway electrons created soon after the disruption collide with the plasma facing components damaging this surface and producing a shower of debris that traverses the toroidal vessel and falls over the inner bumper limiter.

  7. Liquid metal MHD heat transfer investigations apllied to fusion Tokamak reactor cooling ducts

    NASA Astrophysics Data System (ADS)

    Sviridov, V. G.; Ivochkin, Yu. P.; Razuvanov, N. G.; Zhilin, V. G.; Genin, L. G.; Ivanova, O. N.; Averianov, K. V.

    2003-12-01

    The liquid metal heat transfer experimental investigations were carried out at the joint MPEI-IIHI magnetohydrodynamic (MHD) complex. The united scientific group examines the liquid metal flow in a horizontal heated tube without and under a longitudinal or a transverse magnetic field. Various configurations of the applied heat flux were taken into consideration. All these cases correspond to the flow in the Tokamak fusion reactor blanket or divertor. Temperature fields, temperature fluctuations field, heat transfer intensities were measured. Strong influence of thermogravitational convection was observed in a horizontal heated tube. Depending on the MHD-configuration, magnetic fields (MF) can enhance or weaken this effect. Tables 2, Figs 5, Refs 8.

  8. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  9. Experimental study of thermal crisis in connection with Tokamak reactor high heat flux components

    NASA Astrophysics Data System (ADS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G. P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-04-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology.

  10. Implications of the DD fuel cycle on tokamak reactor technology considerations

    SciTech Connect

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Evans, K. Jr.; Jung, J.; Mattas, R.; Misra, B.; Smith, D.L.; Stevens, H.C.; Turner, L.R.

    1980-01-01

    The initial results of a study to develop a conceptual design of a DD tokamak reactor are reported in this paper. A comprehensive steady-state power and particle balance code has been used to survey a variety of design points. A reference set of parameters (R/sub 0/ = 8.6 m, B/sub max/ = 14 T, ..beta.. = 11%, P/sub th/ = 2500 to 2700 MW) has been selected for further study. The concepts of rf driven currents and a limiter-vacuum system have been examined and appear to offer promise. However, the penalty of rf current drive, in terms of recirculating power, is relatively larger for DD than for DT reactors. Blanket engineering studies have emphasized the optimization of energy multiplication and inboard blanket thickness.

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

    SciTech Connect

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

    1987-11-01

    The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D 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. 23 refs., 9 figs., 4 tabs.

  12. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    SciTech Connect

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  13. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    SciTech Connect

    Rule, K.; Scott, J.; Larson, S.

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  14. Safety in the ARIES-III D- sup 3 He tokamak reactor design

    SciTech Connect

    Herring, J.S.; Dolan, T.J.

    1991-01-01

    The ARIES-3 reactor study is an extensive examination of the viability of a D-{sup 3}He-fueled commercial tokamak power reactor. Because neutrons are produced only through side reactions, the reactor has the significant advantages of 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. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The ARIES-3 reactor uses an organic coolant to cool the first wall, shield and divertor. The organic coolant has a low vapor pressure at the operating temperature required for good thermal efficiency. Radiation damage requires processing the coolant to remove and crack radiolytic products that would otherwise foul cooling surfaces. The cracking process produces waste, which must be disposed of through incineration or burial. We estimated the offsite doses due to incineration at five candidate locations. The plasma confinement requirements for a D-{sup 3}He reactor are much more challenging than those for a D-T reactor. Thus, the demands on the divertor are more severe, particularly during a disruption. We explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. We also modeled a loss-of-cooling accident in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. We analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps. For our reference design, the tritium will be burned in the plasma. These results re-emphasize the need for low activation materials and advanced divertor designs, even in reactors using advanced fuels.

  15. Safety in the ARIES-3 D-He-3 tokamak reactor design

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Dolan, T. J.

    1991-11-01

    The ARIES-3 reactor study is an extensive examination of the viability of a D-He-3 He-fueled commercial tokamak power reactor. Because neutrons are produced only through side reactions, the reactor has the significant advantages of 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. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The ARIES-3 reactor uses an organic coolant to cool the first wall, shield and divertor. The organic coolant has a low vapor pressure at the operating temperature required for good thermal efficiency. Radiation damage requires processing the coolant to remove and crack radiolytic products that would otherwise foul cooling surfaces. The cracking process produces waste, which must be disposed of through incineration or burial. We estimated the offsite doses due to incineration at five candidate locations. The plasma confinement requirements for a D-He-3 reactor are much more challenging than those for a D-T reactor. Thus, the demands on the divertor are more severe, particularly during a disruption. We explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. We also modeled a loss-of-cooling accident in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. We analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps. For our reference design, the tritium will be burned in the plasma. These results re-emphasize the need for low activation materials and advanced divertor designs, even in reactors using advanced fuels.

  16. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    SciTech Connect

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  17. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  18. Modeling of neutral hydrogen velocities in the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Stotler, D. P.; Skinner, C. H.; Budny, R. V.; Ramsey, A. T.; Ruzic, D. N.; Turkot, R. B., Jr.

    1996-11-01

    Monte Carlo neutral transport simulations of hydrogen velocities in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] are compared with experiment using the Doppler-broadened Balmer-α spectral line profile. Good agreement is obtained under a range of conditions, validating the treatment of charge exchange, molecular dissociation, surface reflection, and sputtering in the neutral gas code DEGAS [D. Heifetz et al., J. Comput. Phys. 46, 309 (1982)]. A residual deficiency of 10-100 eV neutrals in most of the simulations indicates that further study of the energetics of H+2 dissociation for electron energies in excess of 100 eV is needed.

  19. Mirror test for International Thermonuclear Experimental Reactor at the JET tokamak: An overview of the program

    SciTech Connect

    Rubel, M. J.; De Temmerman, G.; Coad, J. P.; Vince, J.; Drake, J. R.; Le Guern, F.; Murari, A.; Pitts, R. A.; Walker, C.; JET-EFDA Contributors

    2006-06-15

    Metallic mirrors will be essential components of all optical spectroscopy and imaging systems for plasma diagnosis that will be used at the next-step magnetic fusion experiment, International Thermonuclear Experimental Reactor (ITER). Any change of the mirror performance, in particular, reflectivity, will influence the quality and reliability of detected signals. At the instigation of the ITER Design Team, a dedicated technical and experimental activity aiming at the assessment of mirror surface degradation as a result of exposure to the plasma has been initiated on the JET tokamak. This article provides a comprehensive overview of the mirror test program, including design details of the mirror samples and their supports, their locations within JET, and the issue of optical characterization of the mirrors both before and after exposure. The postexposure characterization is particularly challenging in JET as a consequence of an environment in which both tritium and beryllium are present.

  20. Dielectronic satellite spectra of hydrogenlike iron from TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Decaux, V. ); Bitter, M.; Hsuan, H.; von Goeler, S.; Hill, K.W.; Hulse, R.A.; Taylor, G.; Park, H. . Plasma Physics Lab.); Bhalla, C.P. . Dept. of Physics)

    1990-08-01

    Spectra of hydrogenlike iron, Fe26, have been observed from Tokamak Fusion Test Reactor (TFTR) plasmas with a high-resolution crystal spectrometer. The experimental arrangement permits simultaneous observation of the Fe26 Ly-{alpha}{sub 1}and Ly-{alpha}{sub 2} lines and the associated dielectronic satellites, which are due to transitions 1snl-2pnl{prime} with n {ge} 2, as well as the heliumlike 1s{sup 2}({sup 1}S{sub 0}){minus}1s4p({sup 1}P{sub 1})and both hydrogenlike Ly-{beta}{sub 1} and Ly-{beta}{sub 2} lines from chromium. Relative wavelengths and line intensities can be determined very accurately. The spectral data are in very good agreement with theoretical calculations. The observed spectra have also been used to estimate the total dielectronic recombination rate coefficient of Fe26. 30 refs., 4 figs., 3 tabs.

  1. Neutron dosimetry qualification experiments for the Tokamak Fusion Test Reactor Lithium Blanket Module program

    SciTech Connect

    Tsang, F.Y.; Harker, Y.D.; Anderi, R.A.; Nigg, D.W.; Jassby, D.L.

    1986-11-01

    The Tokamak Fusion Test Reactor (TFTR) Lithium Blanket module (LBM) program is a first-of-kind neutronics experiment involving a toroidal fusion neutron source. Qualification experiments have been conducted to develop primary measurement techniques and verify dosimetry materials that will be used to characterize the neutron environment inside and on the surfaces of the LBM. The deuterium-tritium simulation experiments utilizing a 14-MeV neutron generator and a fusion blanket mockup facility at the Idaho National Engineering Laboratory are described. Results and discussions are presented that identify the quality and limitations of the measured integral reaction data, including the minimum fluence requirement for the TFTR experiment and the use of such data in neutron spectrum adjustment and in predicting integral performance parameters, e.g., tritium production.

  2. Summary of Apollo; A D- sup 3 He tokamak reactor design

    SciTech Connect

    Kulcinski, G.L.; Blanchard, T.P.; El-Guebaly, L.A.; Emert, G.A.; Khater, H.Y.; Maynard, C.W.; Mogahed, E.A.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J. . Fusion Technology Inst.)

    1992-07-01

    In this paper, the key features of Apollo, a conceptual D-{sup 3}He tokamak reactor for commercial electricity production, are summarized. The 1000-MW (electric) design utilizes direct conversion of transport, neutron, and bremsstrahlung radiation power. The direct conversion method uses reactants, and the thermal conversion cycle uses an organic coolant. Apollo operates in the first-stability regime, with a major radius of 7.89 m, a peak magnetic field on the toroidal field coils of 19.3 T, a 53-MA plasma current, and a 6.7% beta value. The low neutron production of the D-{sup 3}He fuel cycle greatly reduces the radiation damage rate and allows a full-lifetime first wall and structure made of standard steels with only slight modifications to reduce activation levels.

  3. Development of integrated insulation joint for cooling pipe in tokamak reactor

    NASA Astrophysics Data System (ADS)

    Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro

    1994-08-01

    In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel / ceramics / stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured.

  4. Anomalous Loss of DT Alpha Particles in the Tokamak Fusion Test Reactor

    SciTech Connect

    Herrmann, Hans W.

    1997-06-01

    Princeton's Tokamak Fusion Test Reactor (TFTR) is the first experimental fusion device to routinely use tritium to study the deuterium-tritium (DT) fusion reaction,allowing the first systematic study of DT alpha particles in tokamak plasmas. A crucial aspect of alpha-particle physics is the fraction of alphas that escape from the plasma, particularly since these energetic particles can do severe damage to the first wall of a reactor. An escaping alpha collector probe has been developed for TFTR's DT phase. Energy distributions of escaping alphas have been determined by measuring the range of alpha-particles implanted into nickel foils located within the alpha collector. Results at 1.0 MA of plasma current are in good agreement with predictions for first orbit alpha loss. Results at 1.8 MA, however, show a significant anomalous loss of partially thermalized alphas (in addition to the expected first orbit loss), which is not observed with the lost alpha scintillator detectors in DT plasmas, but does resemble the anomalous "delayed" loss seen in DD plasmas. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations. An experiment designed to study the effect of plasma major radius shifts on alpha-particle loss has led to a better understanding of alpha-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing alpha-particles forced to move toward higher magnetic field during an inward major radius shift (i.e. compression) would mirror and become trapped particles, leading to increased alpha loss. Such an effect was looked for during the shift experiment, however, no significant changes in alpha loss to the 90 degree lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an alpha-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized alpha-particles near the

  5. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  6. Review of deuterium-tritium results from the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    McGuire, K. M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J. W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, Cris W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M.; Bell, M. G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C. S.; Cheng, C. Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D. S.; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P. C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Fujita, T.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Hawryluk, R. J.; Heidbrink, W.; Herrmann, H. W.; Hill, K. W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Loughlin, M. J.; Machuzak, J.; Majeski, R.; Mansfield, D. K.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M. P.; McChesney, J.; McCormack, B.; McCune, D. C.; McKee, G.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Parks, P.; Paul, S. F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Scillia, R.; Scott, S. D.; Semenov, I.; Senko, T.; Sesnic, S.; Sissingh, R.; Skinner, C. H.; Snipes, J.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Tighe, W.; Timberlake, J. R.; Tobita, K.; Towner, H. H.; Tuszewski, M.; von Halle, A.; Vannoy, C.; Viola, M.; von Goeler, S.; Voorhees, D.; Walters, R. T.; Wester, R.; White, R.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Winston, J.; Wright, K.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zavereev, V.; Zweben, S. J.

    1995-06-01

    After many years of fusion research, the conditions needed for a D-T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D-T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m-3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D-T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed

  7. A Compact Nuclear Fusion Reactor for Space Flights

    NASA Astrophysics Data System (ADS)

    Nastoyashchiy, Anatoly F.

    2006-05-01

    A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products — charged particles. The energy balance considerably improves, as synchrotron radiation turn out "captured" in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

  8. Spent fuel utilization in a compact traveling wave reactor

    SciTech Connect

    Hartanto, Donny; Kim, Yonghee

    2012-06-06

    In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

  9. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade

    NASA Astrophysics Data System (ADS)

    Arredondo Parra, R.; Moreno Quicios, R.; Ploeckl, B.; Birkenmeier, G.; Herrmann, A.; Kocsis, G.; Laggner, F. M.; Lang, P. T.; Lunt, T.; Macian-Juan, R.; Rohde, V.; Sellmair, G.; Szepesi, T.; Wolfrum, E.; Zeidner, W.; Neu, R.

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 1019 atoms (0.21 mg) to 1.64 × 1020 atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet.

  10. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade.

    PubMed

    Arredondo Parra, R; Moreno Quicios, R; Ploeckl, B; Birkenmeier, G; Herrmann, A; Kocsis, G; Laggner, F M; Lang, P T; Lunt, T; Macian-Juan, R; Rohde, V; Sellmair, G; Szepesi, T; Wolfrum, E; Zeidner, W; Neu, R

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 10(19) atoms (0.21 mg) to 1.64 × 10(20) atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet. PMID:26931850

  11. Calculations of alpha particle loss for reversed magnetic shear in the Tokamak Fusion Test Reactor

    SciTech Connect

    Redi, M.H.; White, R.B.; Batha, S.H.; Levinton, F.M.; McCune, D.C.

    1997-03-01

    Hamiltonian coordinate, guiding center code calculations of the toroidal field ripple loss of alpha particles from a reversed shear plasma predict both total alpha losses and ripple diffusion losses to be greater than those from a comparable non-reversed magnetic shear plasma in the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. High central q is found to increase alpha ripple losses as well as first orbit losses of alphas in the reversed shear simulations. A simple ripple loss model, benchmarked against the guiding center code, is found to work satisfactorily in transport analysis modelling of reversed and monotonic shear scenarios. Alpha ripple transport on TFTR affects ions within r/a=0.5, not at the plasma edge. The entire plasma is above threshold for stochastic ripple loss of alpha particles at birth energy in the reversed shear case simulated, so that all trapped 3.5 MeV alphas are lost stochastically or through prompt losses. The 40% alpha particle loss predictions for TFTR suggest that reduction of toroidal field ripple will be a critical issue in the design of a reversed shear fusion reactor.

  12. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  13. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    SciTech Connect

    Bell, M.G.; Beer, M.; Batha, S.

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} {approx} 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions.

  14. Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2015-11-01

    The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.

  15. MIRI: A multichannel far-infrared laser interferometer for electron density measurements on TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Mansfield, D.K.; Park, H.K.; Johnson, L.C.; Anderson, H.M.; Chouinard, R.; Foote, V.S.; Ma, C.H.; Clifton, B.J.

    1987-07-01

    A ten-channel far-infrared laser interferometer which is routinely used to measure the spatial and temporal behavior of the electron density profile on the TFTR tokamak is described and representative results are presented. This system has been designed for remote operation in the very hostile environment of a fusion reactor. The possible expansion of the system to include polarimetric measurements is briefly outlined. 13 refs., 8 figs.

  16. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    SciTech Connect

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-19

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  17. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    NASA Astrophysics Data System (ADS)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-01

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  18. Development of Compact Quasi-Axisymmetric Stellarator Reactor Configurations

    SciTech Connect

    L.P. Ku; M. Zarnstorff; R.B. White; W.A. Cooper; R. Sanchez; H. Neilson; J.A. Schmidt

    2003-09-19

    We have started to examine the reactor potential of quasi-axisymmetric (QA) stellarators with an integrated approach that includes systems evaluation, engineering considerations, and plasma and coil optimizations. In this paper, we summarize the progress made so far in developing QA configurations with reduced alpha losses while retaining good MHD stability properties. The minimization of alpha losses is achieved by directly targeting the collisionless orbits to prolong the average resident times. Configurations with an overall energy loss rate of {approx}10% or less, including collisional contributions, have been found. To allow remotely maintaining coils and machine components in a reactor environment, there is a desire to simplify to the extent possible the coil design. To this end, finding a configuration that is optimized not only for the alpha confinement and MHD stability but also for the good coil and reactor performance, remains to be a challenging task.

  19. Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1983-11-01

    Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles.

  20. Evaluation of weldments in type 21-6-9 stainless steel for compact ignition tokamak structural applications, phase 1

    NASA Astrophysics Data System (ADS)

    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.

  1. A compact breed and burn fast reactor using spent nuclear fuel blanket

    SciTech Connect

    Hartanto, D.; Kim, Y.

    2012-07-01

    A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

  2. Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.

    1994-04-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet{sup {trademark}} system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of {approximately} 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of {alpha}-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined {alpha}-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed.

  3. On the power and size of tokamak fusion pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.; Hugill, J.; Buxton, P. F.

    2015-03-01

    It is generally accepted that the route to fusion power involves large devices of ITER scale or larger. However, we show, contrary to expectations, that for steady state tokamaks operating at fixed fractions of the density and beta limits, the fusion gain, Qfus, depends mainly on the absolute level of the fusion power and the energy confinement, and only weakly on the device size. Our investigations are carried out using a system code and also by analytical means. Further, we show that for the two qualitatively different global scalings that have been developed to fit the data contained in the ITER ELMy H-mode database, i.e. the normally used beta-dependent IPB98y2 scaling and the alternative beta-independent scalings, the power needed for high fusion performance differs substantially, typically by factors of three to four. Taken together, these two findings imply that lower power, smaller, and hence potentially lower cost, pilot plants and reactors than currently envisaged may be possible. The main parameters of a candidate low power (˜180 MW), high Qfus (˜5), relatively small (˜1.35 m major radius) device are given.

  4. Preparations for deuterium-tritium experiments on the Tokamak Fusion Test Reactor*

    NASA Astrophysics Data System (ADS)

    Hawryluk, R. J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J. W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C. W.; Barnes, G.; Batchelor, D. B.; Bateman, G.; Batha, S.; Baylor, L. A.; Beer, M.; Bell, M. G.; Biglow, T. S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C. Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D. S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P. C.; Ernst, D.; Faunce, J.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R. C.; Golian, T.; Gorelenkov, N.; Gouge, M. J.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Heidbrink, W.; Hermann, H. W.; Hill, K. W.; Hirshman, S.; Hoffman, D. J.; Hosea, J.; Hulse, R. A.; Hsuan, H.; Jaeger, E. F.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Laughlin, M. J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D. E.; Marchlik, M.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M. P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D. C.; McGuire, K. M.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Milora, S. L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Paul, S. F.; Pavlov, Yu. I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Pitcher, S.; Popovichev, S.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Sabbagh, S. A.; Sasao, M.; Scharer, J.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Scillia, R.; Scott, S. D.; Senko, T.; Sissingh, R.; Skinner, C.; Snipes, J.; Snook, P.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Timberlake, J. R.; Towner, H. H.; Ulrickson, M.; von Halle, A.; Vannoy, C.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Wright, K.; Wong, D.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yeun, A.; Yoshikawa, S.; Young, K. M.; Zakharov, L.; Zarnstorff, M. C.; Zweben, S. J.

    1994-05-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium-tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed.

  5. Anomalous fast ion losses at high β on the tokamak fusion test reactor

    SciTech Connect

    Fredrickson, E. D.; Bell, M. G.; Budny, R. V.; Darrow, D. S.; White, R.

    2015-03-15

    This paper describes experiments carried out on the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] to investigate the dependence of β-limiting disruption characteristics on toroidal field strength. The hard disruptions found at the β-limit in high field plasmas were not found at low field, even for β's 50% higher than the empirical β-limit of β{sub n} ≈ 2 at high field. Comparisons of experimentally measured β's to TRANSP simulations suggest anomalous loss of up to half of the beam fast ions in the highest β, low field shots. The anomalous transport responsible for the fast ion losses may at the same time broaden the pressure profile. Toroidal Alfvén eigenmodes, fishbone instabilities, and Geodesic Acoustic Modes are investigated as possible causes of the enhanced losses. Here, we present the first observations of high frequency fishbones [F. Zonca et al., Nucl. Fusion 49, 085009 (2009)] on TFTR. The interpretation of Axi-symmetric Beam-driven Modes as Geodesic Acoustic Modes and their possible correlation with transport barrier formation are also presented.

  6. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  7. Thin blanket design for MINIMARS - A compact tandem mirror fusion reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Sawan, M.E.; El-Guebaly, L.A.; Wittenberg, L.J.; Corradini, M.L.; Vogelsang, W.F.; Kulcinski, G.L.

    1986-11-01

    Recent fusion power reactor designs have shown a trend toward lower power, lower cost, higher mass utilization compact configurations with inherent safety, in order to improve the economic aspects of fusion and make them more competitive with other energy sources. Since the blanket thickness directly impacts the size and mass of the remaining reactor components, it is prudent to minimize its thickness while ensuring adequate neutronic and thermal performance. This paper describes the blanket for the MINI-MARS compact tandem mirror fusion power reactor. The blanket which utilizes HT-9 ferritic steel structure, LiPb breeder, Be multiplier/moderator and He gas cooling is only 17 cm thick and is backed up by a steel reflector. Helium gas cools the blanket and reflector in series and the outlet temperature of 575/sup 0/C gives a gross thermal power cycle efficiency of 42.7%.

  8. Shielding analysis of a small compact space nuclear reactor

    NASA Astrophysics Data System (ADS)

    Woodrow, Lee L., Jr.

    1987-08-01

    The SP-100 reactor, currently in its developmental stage, has a layered tungsten-lithium hydride shield. Studies indicate that this shield configuration is the lightest weight shield. This configuration and three other shielding concepts were analyzed to determine the lightest shield and to determine the shield configuration with the smallest volume. The other concepts were a boron carbide-beryllium layered shield, and a lithium hydride-beryllium shield. FEMP2D and FEMP1D codes were used in this analysis. These codes were developed at Sandia National Laboatory (SNL), using the input from another code, RFCC, which produced enery dependent dose conversion factors, and determined the shields' ability to attenuate the neutron and gamma radiation to permissible dose limits. The results of this analysis show that the lithium hydribe-tungsten layered shield was indeed the lightest weight shield. Volume, not weight, may be the driving factor in determining the shield configuration.

  9. Surface analysis of 1984/1985 Tokamak Fusion Test Reactor limiter tiles

    SciTech Connect

    Pontau, A.E.; Wampler, W.R.; Mills, B.E.; Doyle, B.L.; Wright, A.F.; Ulrickson, M.A.; LaMarche, P.H.; Dylla, H.F.; Fukuda, S.

    1986-05-01

    Bare POCO AXF-5Q graphite tiles were installed as the plasma-facing surface of the Tokamak Fusion Test Reactor (TFTR) movable limiter for a series of approx. 2700 high power discharges (600 with up to 6 MW neutral beams). During this operating phase, erosion and deposition processes modified the surface of the limiter. In the regions of the most intense plasma contact, which reached temperatures over 2400/sup 0/C, only small amounts of metallic impurities (<5 x 10/sup 16/ atoms/cm/sup 2/) are observed during subsequent beta backscattering and proton-induced x-ray emission (PIXE) analysis. Also observed in these regions are several small areas of surface crazing, and scattered droplets of metals (approx. 1 mm diam) presumed to originate from melted internal hardware. In regions more removed from direct plasma contact, thicker metallic deposits are observed (approx. 2 x 10/sup 18/ atoms/cm/sup 2/). There are more of these deposits on the side facing the electron drift direction than on the ion drift side. The metallic deposits are composed of approx. 55% Ni, 25% Cr, and 20% Fe and several trace elements. Near-surface retained deuterium levels vary from 10/sup 16/ to 10/sup 18//cm/sup 2/ along one ion-side/electron-side pair of tiles. Deuterium from TFTR operations in the bulk of a tile may have been as much as 5 ppm. Hydrogen in the tile was approx.1 at. %. Surface roughness measured outside the intense plasma contact region is much more than the original POCO AXF-5Q graphite, whereas within the plasma contact region, tile surfaces are somewhat smoother than they were initially. Two of the tiles in high heat flux regions broke during operation. The resulting geometry change led to areas of enhanced erosion and nearby redeposition of carbon.

  10. Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

    SciTech Connect

    Budny, R.V.; Alper, B.; Borba, D.; Cordey, J.G.; Ernst, D.R.; Gowers, C.

    2001-02-02

    First results of gyrokinetic analysis of JET [Joint European Torus] ELMy [Edge Localized Modes] H-mode [high-confinement modes] plasmas are presented. ELMy H-mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER [International Thermonuclear Experimental Reactor]. Relatively high performance for long duration has been achieved and the scaling appears to be favorable. It will be necessary to sustain low Z(subscript eff) and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition between Type I to III ELMs at the heating power threshold. Linear gyrokinetic analysis gives the growth rate, gamma(subscript lin) of the fastest growing modes. The flow-shearing rate omega(subscript ExB) and gamma(subscript lin) are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high |gamma(subscript ExB)|/gamma(subscript lin) near the top of the pedestal for high confinement.

  11. Neoclassical Simulations of Fusion Alpha Particles in Pellet Charge Exchange Experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    Batha, S.H.; Budny, R.V.; Darrow, D.S.; Levinton, F.M.; Redi, M.H.; et al

    1999-02-01

    Neoclassical simulations of alpha particle density profiles in high fusion power plasmas on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 5 (1998) 1577] are found to be in good agreement with measurements of the alpha distribution function made with a sensitive active neutral particle diagnostic. The calculations are carried out in Hamiltonian magnetic coordinates with a fast, particle-following Monte Carlo code which includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss and collisional effects. New global loss and confinement domain calculations allow an estimate of the actual alpha particle densities measured with the pellet charge exchange diagnostic.

  12. New interpretation of alpha-particle-driven instabilities in deuterium-tritium experiments on the Tokamak Fusion Test Reactor.

    PubMed

    Nazikian, R; Kramer, G J; Cheng, C Z; Gorelenkov, N N; Berk, H L; Sharapov, S E

    2003-09-19

    The original description of alpha particle driven instabilities in the Tokamak Fusion Test Reactor in terms of toroidal Alfvén eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the antiballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time. PMID:14525368

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

    SciTech Connect

    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-{sup 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.

  14. Alpha Particle-Driven Toroidal Alfven Eigenmodes in Tokamak Fusion Test Reactor Deuterium-Tritium Plasmas: Theory and Experiments

    SciTech Connect

    Budny, R.; Chang, Z.; Fu, G.Y.; Nazikian, R.

    1998-07-09

    The toroidal Alfvén eigenmodes (TAE) in the Tokamak Fusion Test Reactor [K. Young, et al., Plasma Phys. Controlled Fusion 26, 11 (1984)]deuterium-tritium plasmas are analyzed using the NOVA-K code [C.Z. Cheng, Phys. Reports 211, 1 (1992)]. The theoretical results are compared with the experimental measurements in detail. In most cases, the theory agrees with the observations in terms of mode frequency, mode structure, and mode stability. However, one mode with toroidal mode number n = 2 is observed to be poloidally localized on the high field side of the magnetic axis with a mode frequency substantially below the TAE frequency.

  15. A New Interpretation of Alpha-particle-driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    R. Nazikian; G.J. Kramer; C.Z. Cheng; N.N. Gorelenkov; H.L. Berk; S.E. Sharapov

    2003-03-26

    The original description of alpha-particle-driven instabilities in the Tokamak Fusion Test Reactor (TFTR) in terms of Toroidal Alfvin Eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the anti-ballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time.

  16. ECH tokamak

    SciTech Connect

    Firestone, M.A.; Mau, T.K.; Conn, R.W.

    1985-04-01

    A small steady-state tokamak capable of producing power in the 100 to 300 MWe range and relying on electron cyclotron RF heating (ECH) for both heating and current drive is described. Working in the first MHD stability regime for tokamaks, the approach adheres to the recently discovered maximum beta limit. An appropriate figure of merit is the ratio of the fusion power to absorbed RF power. Efficient devices are feasible at both small and large values of fusion power, thereby pointing to a development path for an attractive commercial fusion reactor.

  17. Collaboration on Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Text Reactor. Final report

    SciTech Connect

    Intrator, T.

    2000-06-01

    This proposal was peer reviewed and funded as a Collaboration on ''Low Phase Speed Radio Frequency Current Drive Experiments at the Tokamak Fusion Test Reactor''. The original plans we had were to carry out the collaboration proposal by including a post doctoral scientist stationed at PPPL. In response to a 60+% funding cut, all expenses were radically pruned. The post doctoral position was eliminated, and the Principal Investigator (T. Intrator) carried out the brunt of the collaboration. Visits to TFTR enabled T. Intrator to set up access to the TFTR computing network, database, and get familiar with the new antennas that were being installed in TFTR during an up to air. One unfortunate result of the budget squeeze that TFTR felt for its last year of operation was that the experiments that we specifically got funded to perform were not granted run time on TFTR., On the other hand we carried out some modeling of the electric field structure around the four strap direct launch Ion Bernstein Wave (IBW) antenna that was operated on TFTR. This turned out to be a useful exercise and shed some light on the operational characteristics of the IBW antenna and its coupling to the plasma. Because of this turn of events, the project was renamed ''Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Test Reactor''.

  18. Nuclear aspects of tokamak fusion test reactor (TFTR) diagnostics and instrumentation

    SciTech Connect

    Young, K.M.

    1982-01-01

    There are five principal aspects of the nuclear radiation from the high temperature plasmas of TFTR on its plasma diagnostic equipment. i) Important information about the plasma properties to be obtained from measurement of the neutrons, or other fusion reaction products. ii) Experimental studies to give design data for future tokamak devices and their instrumentation. iii) Transient noise or damage effects on the array of detectors for the collection of physics data about the plasma. iv) The effect of tritium on detectors that necessarily are in vacuum, directly connected to the tokamak vacuum vessel. v) Damage of diagnostic components mounted close to the vacuum vessel. Each of these topics will be addressed after a brief description of the TFTR tokamak and its radiation environment.

  19. A Compact Torus Fusion Reactor Utilizing a Continuously Generated Strings of CT's. The CT String Reactor, CTSR.

    SciTech Connect

    Hartman, C W; Reisman, D B; McLean, H S; Thomas, J

    2007-05-30

    A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal field opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall.

  20. Development and analysis of a compact low-conversion ratio fast burner reactor.

    SciTech Connect

    Smith, M. A.; Hill, R. N.; Nuclear Engineering Division

    2006-05-12

    This report explores design options for compact fast burner reactors that can achieve low conversion ratios. Operational characteristics and whole-core reactivity coefficients are generated and contrasted with low conversion ratio designs of previous studies. A compact core point design is then selected and detailed reactivity coefficients are displayed and discussed. The effectiveness of fast spectrum systems for actinide transmutation has been well documented. The key advantage of the fast spectrum resides in the severely reduced capture/fission ratios. this inhibits the production of the higher actinides that dominate the long-term radiotoxicity of nuclear waste. In conventional fast burner studies, the transmutation rate was limited by constraints placed on the fuel composition. In an earlier phase of this study the entire range of fuel compositions (including non-uranium fuel) was explored to assess the performance and safety limits of fast burner reactor systems. In this report, similar fuel compositions are utilized for application in compact configurations to achieve conversion ratios below 0.5.

  1. Spherical tokamaks with plasma centre-post

    NASA Astrophysics Data System (ADS)

    Ribeiro, Celso

    2013-10-01

    The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

  2. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    NASA Astrophysics Data System (ADS)

    Vdovin, V.

    2014-02-01

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

  3. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect

    Vdovin, V.

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  4. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    NASA Astrophysics Data System (ADS)

    Maise, George; Powell, James; Paniagua, John

    2006-01-01

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from ~20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator is ~3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  5. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  6. Concept for Dismantling the Reactor Vessel and the Biological Shield of the Compact Sodium-Cooled Nuclear Reactor Facility (KNK)

    SciTech Connect

    Hillebrand, I.; Benkert, J.

    2002-02-27

    The Compact Sodium-cooled Nuclear Reactor Facility (KNK) was an experimental nuclear power plant of 20 MW electric power erected on the premises of the Karlsruhe Research Center. The plant was initially run as KNK I with a thermal core between 1971 and 1974 and then, between 1977 and 1991, with a fast core as the KNK II fast breeder plant. Under the decommissioning concept, the plant is to be decommissioned completely to green field conditions at the end of 2005 in ten steps, i.e. under the corresponding ten decommissioning permits. To this day, nine decommissioning permits have been issued, the first one in 1993 and the most recent one, number nine, in 2001. The decommissioning and demolition activities covered by decommissioning permits 1 to 7 have been completed. Under the 8th Decommissioning Permit, the components of the primary system and the rotating reactor top shield are to be removed by late 2001. Under the 9th Decommissioning Permit, the reactor vessel with its internals, the primary shield, and the biological shield are to be dismantled. The residual sodium volume in the reactor vessel was estimated to amount to approx. 30 l. The maximum Co-60 activation is on the order of 107-108 Bq/g; the maximum dose rate in the middle of the vessel was measured in April 1997 to be 55 Sv/h. The difficulty involved especially in dismantling KNK, on the one hand, is posed by the residual sodium in the plant, which determines the choice of neither wet nor thermical techniques to be used in disassembly. Another difficulty is caused by the depth of activation by fast neutrons, as a result of which not only the reactor vessel proper, but also the entire primary shield (60 cm of grey cast iron) and large parts of the biological shield must be disassembled and disposed of under remote control.

  7. Thermionic plasma injection for the Lockheed Martin T4 Compact Fusion Reactor experiment

    NASA Astrophysics Data System (ADS)

    Heinrich, Jonathon

    2015-11-01

    Lockheed Martin's Compact Fusion Reactor (CFR) concept relies on diamagnetic confinement in a magnetically encapsulated linear ring cusp geometry. Plasma injection into cusp field configurations requires careful deliberation. Previous work has shown that axial injection via a plasma gun is capable of achieving high-beta conditions in cusp configurations. We present a pulsed, high power thermionic plasma source and the associated magnetic field topology for plasma injection into the caulked-cusp magnetic field. The resulting plasma fueling and cross-field diffusion is discussed.

  8. Final Report for Monitoring of Reactor Antineutrinos with Compact Germanium Detectors

    SciTech Connect

    Orrell, John L.; Collar, J. I.

    2009-07-01

    This 2008 NCMR project has pursued measurement of the antineutrino-nucleus coherent scattering interaction using a low-energy threshold germanium gamma-ray spectrometer of roughly one-half kilogram total mass. These efforts support development of a compact system for monitoring the antineutrino emission from nuclear reactor cores. Such a monitoring system is relevant to nuclear safeguards and nuclear non-proliferation in general by adding a strong method for assuring quantitative material balance of special nuclear material in the nuclear fuel cycle used in electricity generation.

  9. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    SciTech Connect

    Schriener, T. M.; El-Genk, M. S.

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  10. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    SciTech Connect

    Redi, M.H.; Batha, S.H.; Budny, R.V.; Darrow, D.S.; Levinton, F.M.; McCune, D.C.; Medley, S.S.; Petrov, M.P.; von Goeler, S.; White, R.B.; Zarnstorff, M.C.; Zweben, S.J.; TFTR Team

    1999-07-01

    Neoclassical simulations of alpha particle density profiles in high fusion power plasmas on the Tokamak Fusion Test Reactor [Phys. Plasmas {bold 5}, 1577 (1998)] are found to be in good agreement with measurements of the alpha distribution function made with a sensitive active neutral particle diagnostic. The calculations are carried out in Hamiltonian magnetic coordinates with a fast, particle-following Monte Carlo code which includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss and collisional effects. New calculations show that monotonic shear alpha particles are virtually unaffected by toroidal field ripple. The calculations show that in reversed shear the confinement domain is not empty for trapped alphas at birth and allow an estimate of the actual alpha particle densities measured with the pellet charge exchange diagnostic. {copyright} {ital 1999 American Institute of Physics.}

  11. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    SciTech Connect

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.; Chan, A.A.; England, A.C.; Hendel, H.W.; Medley, S.S.; Nieschmidt, E.; Roquemore, A.L.; Scott, S.D.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and /sup 3/He ions, respectively. When the plasma was compressed, the d(d,n)/sup 3/He fusion reaction rate increased a factor of five, and the /sup 3/He(d,p)/sup 4/He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling.

  12. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    NASA Astrophysics Data System (ADS)

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-08-01

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor.

  13. What we should do for transition from current tokamaks to fusion-fission reactor

    NASA Astrophysics Data System (ADS)

    Mirnov, S.

    2012-06-01

    The Russian fission community places several heavy demands to quality of fusion neutron source for the first step of investigation of minority transmutations ("burning") and breading of nuclear fuel. They are: the steady state regime of neutron production (not rare 80% of main operation time), the total power on neutron flux should be not lower than 20MW with surface neutron load not lower than 0.2MW/m2. Between the current fusion devices: mirror traps, reverse field pinches, stellarators, spherical torus and tokamaks only lasts have today the some probability to fulfill in the near future these hard demands. Two well known DT-tokamaks with neutron power production higher 10MW - TFTR and JET-had maximal neutron load approximately 0.1MW/m2 only in transient (with time scale lower 1s) regimes. The quasi steady state neutron emission regime (˜5MW, 5sec) was performed in JET with mean surface neutron load lower than 0.025MW/m2 only. In this communication it will be discussed the main needs of JET scale tokamak improvement for increase on neutron load up to 0.2MW/m2. They are: decrease of Zeff by ECRH and lithium use as plasma facing components, the increase of energy of steady state neutral injectors up to 150-170keV (tritium), the He removal and creation of closed loop of DT fuel circulation.

  14. TIBER II: Tokamak Ignition/Burn Experimental Reactor: 1986 status report

    SciTech Connect

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

    1986-10-23

    Several chapters are presented that cover the following areas: (1) physics basis; (2) current drive; (3) compact divertors; (4) neutron shielding; (5) high-current density, radiation-tolerant magnets; and (6) costs. (MOW)

  15. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    SciTech Connect

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  16. Electron cyclotron heating of a tokamak reactor at down-shifted frequencies

    SciTech Connect

    Fidone, I.; Giruzzi, G.; Mazzucato, E.

    1985-01-01

    The absorption of electron cyclotron waves in a hot and dense tokamak plasma is investigated for the case of the extraordinary mode for outside launching. It is shown that, for electron temperatures T/sub e/ greater than or equal to 5 keV, strong absorption occurs for oblique propagation at frequencies significantly below the electron gyrofrequency at the plasma center. A new density dependence of the wave absorption is found which is more favorable for plasma heating than the familiar n/sub e//sup -1/ scaling.

  17. Overview of the STARFIRE reference commercial tokamak fusion power reactor design

    SciTech Connect

    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.

  18. Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

    SciTech Connect

    R.V. Budny

    2000-11-15

    ELMy H-mode plasmas form the basis of conservative performance predictions for tokalmak reactors of the size of ITER. Relatively high performace for long durations has been achieved and the scaling is favorable. It will be necessary to sustain low Zeff and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one in which the degradation occurs with an intense gas puff, and the other with a spontaneous transition at the heating power threshold from Type I to III ELMs. Linear gryokinetic analysis gives the growth rate, glin of the fastest growing mode. Our results indicate that the flow-shearing rate wExB and glin are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high *wExB*/glin near the top of the pedestal for high confinement.

  19. Particle confinement property in the cusp-mirror field of a compact fusion reactor

    NASA Astrophysics Data System (ADS)

    Zhu, Limin; Liu, Haifeng; Wang, Xianqu

    2016-09-01

    The cusp-mirror magnetic structure in a compact fusion reactor (CFR) is investigated to understand the properties of the particle confinement for the first time. Compared with a cascade magnetic mirror device, its advanced performance is shown by means of test particle simulations. Some interesting results are obtained as follows: the adiabatic region and non-adiabatic region are found in the CFR’s magnetic configuration. In the non-adiabatic region, due to the magnetic field-free space existing, the ions are scattered stochastically and are not directly guided into the loss cone, unlike the particles around the fixed magnetic lines in the adiabatic region, which decrease the ion loss fraction. The CFR’s configuration, combining advantages of cusp-magnetic configuration and mirror-magnetic configuration, leads to confine particles longer than cascade magnetic mirror’s. This phenomenon may be relevant to the construction of advanced magnetic-confinement devices.

  20. Compact Stellarator Path to DEMO

    NASA Astrophysics Data System (ADS)

    Lyon, J. F.

    2007-11-01

    Issues for a DEMO reactor are sustaining an ignited/high-Q plasma in steady state, avoiding disruptions and large variations in power flux to the wall, adequate confinement of thermal plasma and alpha-particles, control of a burning plasma, particle and power handling, etc. Compact stellarators have key advantages -- steady-state high-plasma-density operation without external current drive or disruptions, stability without a close conducting wall or active feedback systems, and low recirculating power -- in addition to moderate plasma aspect ratio, good confinement, and high-beta potential. The ARIES-CS study established that compact stellarators can be competitive with tokamaks as reactors. Many of the issues for a compact stellarator DEMO can be answered using results from large tokamaks, ITER D-T experiments and fusion materials, technology and component development programs, in addition to stellarators in operation, under construction or in development. However, a large next-generation stellarator will be needed to address some physics issues: size scaling and confinement at higher parameters, burning plasma issues, and operation with a strongly radiative divertor. Technology issues include simpler coils, structure, and divertor fabrication, and better cost information.

  1. Ignition in near term D-/sup 3/He tokamak reactors: Appendix B

    SciTech Connect

    Emmert, G.A.; Deng, B.Q.

    1987-01-01

    The prospects for achieving breakeven and ignition in near term ETR type tokamaks under D/He-3 relevant conditions are considered. Using present scaling laws for beta in the first stability regime, it is found that CIT may be close to breakeven with the presently planned toroidal magnetic field system, if the ASDEX H-mode scaling law is used. With Kaye-Goldston scaling, Q = .22 can be attained, but this requires an excessive amount of rf heating power. Larger devices, such as NET/INTOR, can ignite with ASDEX H-mode scaling with an increase of the toroidal field by 20% and removal of the blanket and reduction of the inboard shield to that required for D/He-3. 5 refs., 4 figs., 2 tabs.

  2. Atomic data and level populations of highly ionized Ti for tokamak plasmas. [reactor chamber alloy constituent

    NASA Technical Reports Server (NTRS)

    Bhatia, A. K.; Feldman, U.; Doschek, G. A.

    1980-01-01

    The paper presents calculations of electron impact collision strengths and spontaneous radiative decay rates for titanium ions of the LiI through FI isoelectronic sequences for transitions between levels of the 2S(2)2p(k), 2s2p(k+1), and 2p(k+2) configurations. From these atomic data, excitation-rate coefficients are calculated along with level populations for these three configurations. The calculations of level populations include the effects of proton excitation, and are carried out at electron temperatures and densities typical of tokamak plasmas. Wavelengths of forbidden and intersystem lines are given, and a synthetic spectrum is presented for a typical temperature and density.

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

    SciTech Connect

    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.

  4. Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

    SciTech Connect

    Bartel, N.; Chen, M.; Utgikar, V. P.; Sun, X.; Kim, I. -H.; Christensen, R.; Sabharwall, P.

    2015-04-04

    A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimum combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.

  5. Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

    DOE PAGESBeta

    Bartel, N.; Chen, M.; Utgikar, V. P.; Sun, X.; Kim, I. -H.; Christensen, R.; Sabharwall, P.

    2015-04-04

    A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

  6. Real-time loop-mediated isothermal DNA amplification in compact disc micro-reactors.

    PubMed

    Santiago-Felipe, Sara; Tortajada-Genaro, Luis A; Carrascosa, Javier; Puchades, Rosa; Maquieira, Ángel

    2016-05-15

    An integrated device composed of micro-reactors embedded onto compact discs is proposed for real-time targeted DNA determination. The method principle is based on in-disc loop-mediated isothermal amplification (iD-LAMP) and quantitative optical read-out by a disc drive. In the presence of a target, the turbidimetric or colorimetric properties of reaction solution change, and the transmitted intensity of the disc drive laser modifies according to reaction yield. Monitoring real-time curves allowed the quantitative determination of DNA template amounts. The best amplification/detection results were obtained with micro-reactors (2mm diameter and 1.1mm in depth) drilled on a digital video disc (DVD) and detection based on the colorimetric mode. As proof-of-concept, the assay was applied to detect pathogenic bacteria Salmonella spp. and to identify bovine meat in food samples. Ninety-six samples were simultaneously analysed in 15 min, with high selectivity and sensitivity (5 CFU/mL and 10 µg/g for bacteria and meat, respectively). The in-disc results were comparable to those obtained by conventional LAMP or qPCR approaches. The developed device allows low sample and reagent consumption (3 µL of reaction), portability, ease-of-use, and rapid low-cost high-throughput analyses. PMID:26716424

  7. Liquid immersion blanket design for use in a compact modular fusion reactor

    NASA Astrophysics Data System (ADS)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  8. Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

    NASA Technical Reports Server (NTRS)

    Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

    2011-01-01

    Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology

  9. Progress in Compact Toroid Experiments

    SciTech Connect

    Dolan, Thomas James

    2002-09-01

    The term "compact toroids" as used here means spherical tokamaks, spheromaks, and field reversed configurations, but not reversed field pinches. There are about 17 compact toroid experiments under construction or operating, with approximate parameters listed in Table 1.

  10. RAMI Analysis for Designing and Optimizing Tokamak Cooling Water System (TCWS) for the ITER's Fusion Reactor

    SciTech Connect

    Ferrada, Juan J; Reiersen, Wayne T

    2011-01-01

    U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment

  11. Distributions of alpha particles escaping to the wall during sawtooth oscillations in the Tokamak Fusion Test Reactor

    SciTech Connect

    Kolesnichenko, Y.I.; Lutsenko, V.V.; White, R.B.; Yakovenko, Y.V.; Zweben, S.J.

    1999-04-01

    It has been observed experimentally in deuterium{endash}tritium shots of the Tokamak Fusion Test Reactor (TFTR) [D. J. Grove and D. M. Meade, Nucl. Fusion {bold 25}, 1167 (1985)] that crashes of sawtooth oscillations may result in very inhomogeneous flux of alpha particles to the wall. To explain this phenomenon, both theoretical analysis and numerical simulation have been carried out. It is concluded that the {open_quotes}crash-induced prompt loss,{close_quotes} i.e., the orbital loss of marginally trapped particles arising because of the crash-induced orbit transformation of circulating particles, is responsible for the flux near the bottom of the vessel, whereas the crash-induced stochastic diffusion of moderately trapped particles explains the large signal near the equatorial plane of the torus. The calculated poloidal distributions of the integral alpha flux are in reasonable agreement with experimental data. The energy spectrum of the escaping particles has also been calculated, which can be used for diagnostics of the crash type. {copyright} {ital 1999 American Institute of Physics.} thinsp

  12. Soft-computing approach to plasma evolution tracking in tokamak reactors

    NASA Astrophysics Data System (ADS)

    Morabito, Francesco C.

    1997-10-01

    Qualitative information about the structure of a mapping can surely be of help in learning a mapping by a collection of input-output pairs. However, there are conditions in which time and some other constraints make guessing the only plausible means for interpreting data. In this paper, the problem of the plasma boundary reconstruction in 'Tokamak' nuclear fusion rectors is assessed. The problem is formulated as an inverse 'identification' problem and the mapping is derived by a properly generated database of simulated experiments. Real data coming from experiments are also available to validate both numerically generated data and extracted model. The identification problem is solved for two different databases by using neural networks and more conventional models. The introduction of techniques derived from soft computing is shown to improve the performance in various respects. Dynamic identification systems appear to be rather demanding also for such systems, for the need of rapidly interpreting real time data for discharge control. Soft computing approaches may yet yield some low cost ways to take decisions during plasma evolution. The approximate analysis of experimental data could also improve the knowledge on the particular problem allowing an evolution of the knowledge base. Experimental data related to ASDEX-Upgrade machine are presented in this work and preliminary processed. Soft computing techniques also allow to simply get ideas about two other interesting problems in plasma engineering, namely, the fault tolerance and the minimization of the number of sensors.

  13. Designing a tokamak fusion reactor—How does plasma physics fit in?

    NASA Astrophysics Data System (ADS)

    Freidberg, J. P.; Mangiarotti, F. J.; Minervini, J.

    2015-07-01

    This paper attempts to bridge the gap between tokamak reactor design and plasma physics. The analysis demonstrates that the overall design of a tokamak fusion reactor is determined almost entirely by the constraints imposed by nuclear physics and fusion engineering. Virtually, no plasma physics is required to determine the main design parameters of a reactor: a , R 0 , B 0 , T i , T e , p , n , τ E , I . The one exception is the value of the toroidal current I , which depends upon a combination of engineering and plasma physics. This exception, however, ultimately has a major impact on the feasibility of an attractive tokamak reactor. The analysis shows that the engineering/nuclear physics design makes demands on the plasma physics that must be satisfied in order to generate power. These demands are substituted into the well-known operational constraints arising in tokamak physics: the Troyon limit, Greenwald limit, kink stability limit, and bootstrap fraction limit. Unfortunately, a tokamak reactor designed on the basis of standard engineering and nuclear physics constraints does not scale to a reactor. Too much current is required to achieve the necessary confinement time for ignition. The combination of achievable bootstrap current plus current drive is not sufficient to generate the current demanded by the engineering design. Several possible solutions are discussed in detail involving advances in plasma physics or engineering. The main contribution of the present work is to demonstrate that the basic reactor design and its plasma physics consequences can be determined simply and analytically. The analysis thus provides a crisp, compact, logical framework that will hopefully lead to improved physical intuition for connecting plasma physic to tokamak reactor design.

  14. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  15. QUANTITATIVE HOMOGENEITY AND IN-CONTACT PARTICLES OF HIGH TEMPERATURE REACTORS (HTR) COMPACTS DETERMINATION VIA X-RAY TOMOGRAPHY

    SciTech Connect

    Lecomte, G.; Letang, J. M.; Tisseur, D.; Banchet, J.; Vitali, M. P.

    2008-02-28

    In AREVA Nuclear Power's High Temperature Reactor (HTR) design called ANTARES, fuel consists of compacts composed of few thousands millimetric quasi-spherical particles dispersed in a graphite matrix. Compact homogeneity, defined as the homogeneous particles spatial distribution in the matrix, as well as the possibility of obtaining particles in contact, need to be assessed since they condition the thermo-mechanical behavior of the nuclear fuel under irradiation. In this paper, image and data processing algorithms are developed to do so, based on X-Ray tomographic images.

  16. Quantitative Homogeneity and In-Contact Particles of High Temperature Reactors (htr) Compacts Determination via X-Ray Tomography

    NASA Astrophysics Data System (ADS)

    Lecomte, G.; Tisseur, D.; Létang, J. M.; Banchet, J.; Vitali, M. P.

    2008-02-01

    In AREVA Nuclear Power's High Temperature Reactor (HTR) design called ANTARES, fuel consists of compacts composed of few thousands millimetric quasi-spherical particles dispersed in a graphite matrix. Compact homogeneity, defined as the homogeneous particles spatial distribution in the matrix, as well as the possibility of obtaining particles in contact, need to be assessed since they condition the thermo-mechanical behavior of the nuclear fuel under irradiation. In this paper, image and data processing algorithms are developed to do so, based on X-Ray tomographic images.

  17. Anomalous loss of DT alpha particles in the Tokamak Fusion Test Reactor

    SciTech Connect

    Herrmann, H.W.

    1997-09-01

    An escaping alpha collector probe has been developed for TFTR`s DT phase. Energy distributions of escaping alphas have been determined by measuring the range of {alpha}-particles implanted into nickel foils located within the alpha collector. Results at 1.0 MA of plasma current are in good agreement with predictions for first orbit alpha loss. Results at 1.8 MA, however, show a significant anomalous loss of partially thermalized alphas (in addition to the expected first orbit loss), which is not observed with the lost alpha scintillator detectors in DT plasmas, but does resemble the anomalous delayed loss seen in DD plasmas. None of the candidate explanations proposed thus far are fully consistent with the anomalous loss observations. An experiment designed to study the effect of plasma major radius shifts on {alpha}-particle loss has led to a better understanding of {alpha}-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing {alpha}-particles forced to move toward higher magnetic field during an inward major radius shift (i.e., compression) would mirror and become trapped particles, leading to increased alpha loss. Such an effect was looked for during the shift experiment, however, no significant changes in alpha loss to the 90{degree} lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an {alpha}-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized {alpha}-particles near the passing/trapped boundary was observed to occur between inward and outward shifts at an intermediate value of plasma current (1.4 MA). This anomalous loss feature is not yet understood.

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

    SciTech Connect

    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.

  19. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    SciTech Connect

    El-Guebaly, L.A.; Sviatoslavsky, I.N.

    1987-01-01

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m/sup 2/ neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10/sup 19/ n/cm/sup 2/. In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H/sub 2/O/LiNO/sub 3/ layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is <95/sup 0/C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of approx.2 were found at the hot spots. 2 refs., 6 figs., 2 tabs.

  20. Nuclear design methodology for analyzing ultra high temperature highly compact ternary carbide reactor

    NASA Astrophysics Data System (ADS)

    Gouw, Reza Raymond

    Recent studies at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) have demonstrated the feasibility of fabricating solid solutions of ternary carbide fuels such as (U,Zr,Nb)C, (U,Zr,Ta)C, (U,Zr,Hf)C and (U,Zr,W)C. The necessity for accurate nuclear design analysis of these ternary carbides in highly compact nuclear space systems prompted the development of nuclear design methodology for analyzing these systems. This study will present the improvement made in the high temperature nuclear cross-sections. It will show the relation between Monte Carlo and Deterministic calculations. It will prove the significant role of the energy spectrum in the multigroup nuclear cross-sections generation in the highly-thermalized-nuclear system. The nuclear design methodology will address several issues in the homogenization of a nuclear system, such as energy spectrum comparison between a heterogeneous system and homogeneous system. It will also address several key points in the continuous and multigroup nuclear cross-sections generation. The study will present the methodology of selecting broad energy group structures. Finally, a comparison between the Monte Carlo and Deterministic methods will be performed for the Square-Lattice Honeycomb Nuclear Space Reactor. In the comparison calculations, it will include the system characterization calculations, such as energy spectrum comparison, 2-D power distributions, temperature coefficient analysis, and water submersion accident analysis.

  1. Thin blanket design for MINIMARS - a compact tandem mirror fusion reactor

    SciTech Connect

    Sviatoslavsky, I.N.; Sawan, M.E.; El-Guebaly, L.A.; Wittenberg, L.J.; Corradini, M.L.; Vogelsang, W.F.; Kulcinski, G.L.

    1986-01-01

    A primary goal in the MINIMARS fusion power reactor design is to achieve the lowest possible cost of electricity and highest mass utilization while maintaining credibility and passive safety. Because the blanket impacts many components, reducing its thickness-while achieving adequate breeding and a high energy multiplication-was of prime importance. The MINIMARS blanket is a helium-gas-cooled design using 17Li-83Pb (LiPb) breeder, HT-9 structure, and beryllium moderator/multiplier. The helium gas is contained in small tubes that are immersed in a close-packed matrix of beryllium balls and LiPb. The result is a compact blanket only 18 cm thick in which only the tubes are operated in a stressed condition, but the blanket structure is designed to withstand a helium gas leak in one of the tubes. By circulating the helium gas from the blanket into the reflector, the reflector energy is recovered at a high temperature giving a gross power cycle efficiency of 42.7% while maintaining a low interface temperature between the breeding material and structure.

  2. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  3. Analysis of alpha particle-driven toroidal Alfv{acute e}n eigenmodes in Tokamak Fusion Test Reactor deuterium{endash}tritium experiments

    SciTech Connect

    Fu, G.Y.; Cheng, C.Z.; Budny, R.; Chang, Z.; Darrow, D.S.; Fredrickson, E.; Mazzucato, E.; Nazikian, R.; Wong, K.L.; Zweben, S.

    1996-11-01

    The toroidal Alfv{acute e}n eigenmodes (TAE) are calculated to be stable in the presently obtained deuterium{endash}tritium plasmas in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. {bold 26}, 11 (1984)]. However, the core localized TAE mode can exist and is less stable than the global TAE modes. The beam ion Landau damping and the radiative damping are the two main stabilizing mechanisms in the present calculation. In future deuterium{endash}tritium experiments, the alpha-driven TAE modes are predicted to occur with a weakly reversed shear profile. {copyright} {ital 1996 American Institute of Physics.}

  4. Systems code assessment of innovations, major design drivers, and minimum sizes of INTOR (International Tokamak Reactor) and ETR-like designs

    SciTech Connect

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

    1987-10-01

    System studies of next-generation superconducting tokamaks are presented here. These studies include examining design changes suggested for the International Tokamak Reactor (INTOR) as a means of reducing the size and simplifying the device and assessing the impact of a series of more aggressive design assumptions suggested in recent Engineering Test Reactor (ETR) studies. Also, a set of candidate design points offering small machines (major radius = 4 m) with a relatively conservative mix of design assumptions is proposed. Some of the design assumptions found to have a major effect on the minimum size are TF coil current density, noninductive current drive, plasma elongation and edge q, plasma temperature for current drive, maximum allowable plasma beta, the minimum required wall load, and assumptions on fixed radial dimensions such as shield thickness, gaps, and plasma scrapeoff. Some design assumptions with less impact on the device size are the OH coil current density, PF configuration (limiter/divertor), and plasma current level. 22 refs., 5 figs., 6 tabs.

  5. ADX - Advanced Divertor and RF Tokamak Experiment

    NASA Astrophysics Data System (ADS)

    Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

    2015-11-01

    The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

  6. A physics study for negative void reactivity in compact supercritical CO{sub 2}-cooled fast reactor

    SciTech Connect

    Kim, Y.; Hartanto, D.; Lee, J. I.

    2013-07-01

    A compact S-CO{sub 2}-cooled fast reactor which has negative Coolant Void Reactivity (CVR) has been investigated. A negative CVR is important for the gas cooled fast reactor as an inherent safety mechanism to prevent the sudden positive reactivity insertion when the loss of coolant accident happens. An alternative solution to reduce the CVR is investigated in this study by using O-17 instead of O-16 in UO{sub 2} fuel. By using O-17 in the fuel, it is found that the CVR can even be negative. Impacts of the radial reflector on the CVR are also evaluated for the small SCO{sub 2} cooled fast reactor in this study. We have considered a pure lead (Pb) reflector and a lead magnesium eutectic (LME) reflector as alternative radial reflectors of the S-CO 2-cooled fast reactor. It has been shown that, with the LME radial reflector, the CVR can be negative, while the pure lead reflector provides a slightly positive CVR. (authors)

  7. Bootstrapped tokamak with oscillating field current drive

    SciTech Connect

    Weening, R.H. )

    1993-07-01

    A magnetic helicity conserving mean-field Ohm's law is used to study bootstrapped tokamaks with oscillating field current drive. The Ohm's law leads to the conclusion that the tokamak bootstrap effect can convert the largely alternating current of oscillating field current drive into a direct toroidal plasma current. This plasma current rectification is due to the intrinsically nonlinear nature of the tokamak bootstrap effect, and suggests that it may be possible to maintain the toroidal current of a tokamak reactor by supplementing the bootstrap current with oscillating field current drive. Steady-state tokamak fusion reactors operating with oscillating field current drive could provide an alternative to tokamak reactors operating with external current drive.

  8. Three-dimensional analysis of tokamaks and stellarators

    PubMed Central

    Garabedian, Paul R.

    2008-01-01

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

  9. Three-dimensional analysis of tokamaks and stellarators.

    PubMed

    Garabedian, Paul R

    2008-09-16

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

  10. Preliminary Design For Conventional and Compact Secondary Heat Exchanger in a Molten Salt Reactor

    SciTech Connect

    Piyush Sabharwall; Mike Patterson; Ali Siahpush; Eung Soo Kim

    2012-07-01

    The strategic goal of the Advance Reactors such as AHTR is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors

  11. Tritium catalyzed deuterium tokamaks

    SciTech Connect

    Greenspan, E.; Miley, G.H.; Jung, J.; Gilligan, J.

    1984-04-01

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the /sup 3/He from the D(D,n)/sup 3/He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general).

  12. A Compact Torus Fusion Reactor Utilizing a Continuously Generated String of CT's. The CT String Reactor, CTSR

    NASA Astrophysics Data System (ADS)

    Hartman, Charles W.; Reisman, David B.; McLean, Harry S.; Thomas, John

    2008-06-01

    A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal field opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall. 60 /proj/ads/abstracts/absload/springerxml.pl* /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9121: total 28 28 10894_2007_Article_9121.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9122: total 16 16 10894_2007_Article_9122.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9123: total 12 12 10894_2007_Article_9123.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9124: total 36 36 10894_2007_Article_9124.xml.meta /proj/ads/abstracts/sources/SPRINGER/files/JOU=10894/VOL=2008.27/ISU=3/ART=9125: total 32 32 10894_2007_Article_9125.xml.meta /proj

  13. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator.

    SciTech Connect

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, Thomas W; Dix, D.; El-GuebalyUniv. Wisco, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D. G.; Zarnstorff, M. C.

    2011-01-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  14. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    NASA Astrophysics Data System (ADS)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  15. Toroidal rotation and radial electric field driven by the lower-hybrid-wave in a tokamak fusion reactor

    SciTech Connect

    Wang Shaojie

    2011-10-15

    A theoretical model is proposed to interpret the counter-current rotation driven by the lower-hybrid-wave observed in the tokamak lower-hybrid-wave parallel current drive experiments. It is found that ions absorb the toroidal momentum indirectly from the wave through collisional friction with the resonant electrons that directly take the momentum from the wave through Landau resonance. This momentum coupling pumps out the ions to produce a negative radial electric field and makes the plasma rotate in the counter-current direction.

  16. Calculation of the absolute detection efficiency of a moderated /sup 235/U neutron detector on the Tokamak Fusion Test Reactor

    SciTech Connect

    Ku, L.P.; Hendel, H.W.; Liew, S.L.

    1989-02-01

    Neutron transport simulations have been carried out to calculate the absolute detection efficiency of a moderated /sup 235/U neutron detector which is used on the TFTR as a part of the primary fission detector diagnostic system for measuring fusion power yields. Transport simulations provide a means by which the effects of variations in various shielding and geometrical parameters can be explored. These effects are difficult to study in calibration experiments. The calculational model, benchmarked against measurements, can be used to complement future detector calibrations, when the high level of radioactivity resulting from machine operation may severely restrict access to the tokamak. We present a coupled forward-adjoint algorithm, employing both the deterministic and Monte Carlo sampling methods, to model the neutron transport in the complex tokamak and detector geometries. Sensitivities of the detector response to the major and minor radii, and angular anisotropy of the neutron emission are discussed. A semi-empirical model based on matching the calculational results with a small set of experiments produces good agreement (+-15%) for a wide range of source energies and geometries. 20 refs., 6 figs., 4 tabs.

  17. Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

    DOE PAGESBeta

    Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; Roney, T. J.; Morrell, S. R.

    2016-02-05

    In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

  18. Stabilized Spheromak Fusion Reactors

    SciTech Connect

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  19. Design and optimization of Artificial Neural Networks for the modelling of superconducting magnets operation in tokamak fusion reactors

    NASA Astrophysics Data System (ADS)

    Froio, A.; Bonifetto, R.; Carli, S.; Quartararo, A.; Savoldi, L.; Zanino, R.

    2016-09-01

    In superconducting tokamaks, the cryoplant provides the helium needed to cool different clients, among which by far the most important one is the superconducting magnet system. The evaluation of the transient heat load from the magnets to the cryoplant is fundamental for the design of the latter and the assessment of suitable strategies to smooth the heat load pulses, induced by the intrinsically pulsed plasma scenarios characteristic of today's tokamaks, is crucial for both suitable sizing and stable operation of the cryoplant. For that evaluation, accurate but expensive system-level models, as implemented in e.g. the validated state-of-the-art 4C code, were developed in the past, including both the magnets and the respective external cryogenic cooling circuits. Here we show how these models can be successfully substituted with cheaper ones, where the magnets are described by suitably trained Artificial Neural Networks (ANNs) for the evaluation of the heat load to the cryoplant. First, two simplified thermal-hydraulic models for an ITER Toroidal Field (TF) magnet and for the ITER Central Solenoid (CS) are developed, based on ANNs, and a detailed analysis of the chosen networks' topology and parameters is presented and discussed. The ANNs are then inserted into the 4C model of the ITER TF and CS cooling circuits, which also includes active controls to achieve a smoothing of the variation of the heat load to the cryoplant. The training of the ANNs is achieved using the results of full 4C simulations (including detailed models of the magnets) for conventional sigmoid-like waveforms of the drivers and the predictive capabilities of the ANN-based models in the case of actual ITER operating scenarios are demonstrated by comparison with the results of full 4C runs, both with and without active smoothing, in terms of both accuracy and computational time. Exploiting the low computational effort requested by the ANN-based models, a demonstrative optimization study has been

  20. The effect of plasma density on the behavior of the plasma boundary in helium discharges in the Tokamak Fusion Test Reactor

    SciTech Connect

    Kilpatrick, S.J.; Pitcher, C.S.; Bell, M.G.; Manos, D.M.; Mansfield, D.K.; Nyberg, I.; Ramsey, A.T.; Stratton, B.C.

    1990-03-01

    Langmuir probe measurements have been made across the plasma boundary in helium discharges of varying central electron densities in the Tokamak Fusion Test Reactor (TFTR). Radial profiles of ion saturation current density, I{sub s}, electron density, n{sub e}, and electron temperature, T{sub e}, were achieved by small displacements of the plasma relative to a stationary probe. Scrape-off lengths outside the plasma boundary decreased somewhat with increasing plasma density. Inside the plasma, I{sub s}, n{sub e}, and T{sub e} rose more steeply for the higher densities than the scrape-off lengths would predict. The floating potential, V{sub f}, decreased sharply inside the boundary as expected, with some density dependence. I{sub s} and n{sub e} increased with plasma density, while T{sub e} decreased a few eV. Measurements of C II emission at the inner wall combined with probe data support a relatively constant edge T{sub e} with increasing plasma density and a constant ratio between the fluxes of helium and carbon ions. 14 refs., 5 figs.

  1. Measurements of the radial structure and poloidal spectra of toroidal Alfvén eigenmodes in the Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Durst, R. D.; Fonck, R. J.; Wong, K. L.; Cheng, C. Z.; Fredrickson, E. D.; Paul, S. F.

    1992-11-01

    Toroidal Alfvén eigenmodes (TAE) have been excited by tangential neutral beam injection in the Tokamak Fusion Test Reactor (TFTR) [Proceedings of the Thirteenth International Conference on Plasma Physics and Controlled Nuclear Fusion Research, 1990, Washington, D.C. (International Atomic Energy Agency, Vienna, 1990), Vol. I, p. 9]. Beam emission spectroscopy (BES) has been used to study the radial structure and the poloidal power spectra of these modes. Radial profiles show a global, standing wave structure with a node near r/a=0.6 and a maximum displacement of about 5-10 mm. The cross-phase profiles and the power spectra both imply that the mode is composed of a mixture of components with various poloidal and toroidal mode numbers, as expected for the TAE. Measurements of the poloidal mode spectrum via BES show good agreement with theoretical simulations performed by a nonvariational, kinetic magnetohydrodynamics stability code (nova-k [Cheng, Phys. Rep. 211, 1 (1992)]). In particular, the dominant harmonics in the poloidal spectrum obey the expected relation m+1/2≊q(r)n.

  2. Shielding analysis of a small compact space nuclear reactor. Final report

    SciTech Connect

    Woodrow, L.L.

    1987-08-01

    The SP-100 reactor concept, currently in its developmental stage, has layered tungsten - lithium hydride shield. Studies indicate that this shield configuration is the lightest weight shield. This configuration and three other shielding concepts were analyzed to determine the lightest shield and to determine the shield configuration with the smallest volume. The other three concepts were a boron carbide - beryllium layered shield, and a lithium hydride - beryllium shield. FEMP2D and FEMP1D codes were used in this analysis. These codes were developed at Sandia National Laboratory (SNL), using the input from another code, RFCC, which produced energy-dependent dose conversion factors, and determined the shield's ability to attenuate the neutron and gamma radiation to permissible dose limits. The results of this analysis show that the lithium hydride - tungsten layered shield was indeed the lightest weight shield. However, a boron carbide - tungsten shield was calculably volume constrained. Therefore volume, not weight, may be the driving factor in determining the shield configuration.

  3. The design of the Tokamak Physics Experiment (TPX)

    NASA Astrophysics Data System (ADS)

    Schmidt, J. A.; Thomassen, K. I.; Goldston, R. J.; Neilson, G. H.; Nevins, W. M.; Sinnis, J. C.; Andersen, P.; Bair, W.; Barr, W. L.; Batchelor, D. B.; Baxi, C.; Berg, G.; Bernabei, S.; Bialek, J. M.; Bonoli, P. T.; Boozer, A.; Bowers, D.; Bronner, G.; Brooks, J. N.; Brown, T. G.; Bulmer, R.; Butner, D.; Campbell, R.; Casper, T.; Chaniotakis, E.; Chaplin, M.; Chen, S. J.; Chin, E.; Chrzanowski, J.; Citrolo, J.; Cole, M. J.; Dahlgren, F.; Davis, F. C.; Davis, J.; Davis, S.; Diatchenko, N.; Dinkevich, S.; Feldshteyn, Y.; Felker, B.; Feng, T.; Fenstermacher, M. E.; Fleming, R.; Fogarty, P. J.; Fragetta, W.; Fredd, E.; Gabler, M.; Galambos, J.; Gohar, Y.; Goranson, P. L.; Greenough, N.; Grisham, L. R.; Haines, J.; Haney, S.; Hassenzahl, W.; Heim, J.; Heitzenroeder, P. J.; Hill, D. N.; Hodapp, T.; Houlberg, W. A.; Hubbard, A.; Hyatt, A.; Jackson, M.; Jaeger, E. F.; Jardin, S. C.; Johnson, J.; Jones, G. H.; Juliano, D. R.; Junge, R.; Kalish, M.; Kessel, C. E.; Knutson, D.; LaHaye, R. J.; Lang, D. D.; Langley, R. A.; Liew, S.-L.; Lu, E.; Mantz, H.; Manickam, J.; Mau, T. K.; Medley, S.; Mikkelsen, D. R.; Miller, R.; Monticello, D.; Morgan, D.; Moroz, P.; Motloch, C.; Mueller, J.; Myatt, L.; Nelson, B. E.; Neumeyer, C. L.; Nilson, D.; O'Conner, T.; Pearlstein, L. D.; Peebles, W. A.; Pelovitz, M.; Perkins, F. W.; Perkins, L. J.; Petersen, D.; Pillsbury, R.; Politzer, P. A.; Pomphrey, N.; Porkolab, M.; Posey, A.; Radovinsky, A.; Raftopoulis, S.; Ramakrishnan, S.; Ramos, J.; Rauch, W.; Ravenscroft, D.; Redler, K.; Reiersen, W. T.; Reiman, A.; Reis, E.; Rewoldt, G.; Richards, D. J.; Rocco, R.; Rognlien, T. D.; Ruzic, D.; Sabbagh, S.; Sapp, J.; Sayer, R. O.; Scharer, J. E.; Schmitz, L.; Schnitz, J.; Sevier, L.; Shipley, S. E.; Simmons, R. T.; Slack, D.; Smith, G. R.; Stambaugh, R.; Steill, G.; Stevenson, T.; Stoenescu, S.; Onge, K. T. St.; Stotler, D. P.; Strait, T.; Strickler, D. J.; Swain, D. W.; Tang, W.; Tuszewski, M.; Ulrickson, M. A.; VonHalle, A.; Walker, M. S.; Wang, C.; Wang, P.; Warren, J.; Werley, K. A.; West, W. P.; Williams, F.; Wong, R.; Wright, K.; Wurden, G. A.; Yugo, J. J.; Zakharov, L.; Zbasnik, J.

    1993-09-01

    The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.

  4. An Assessment of the Penetrations in the First Wall Required for Plasma Measurments for Control of an Advanced Tokamak Plasma Demo

    SciTech Connect

    Kenneth M. Young

    2010-02-22

    A Demonstration tokamak (Demo) is an essential next step toward a magnetic-fusion based reactor. One based on advanced-tokamak (AT) plasmas is especially appealing because of its relative compactness. However, it will require many plasma measurements to provide the necessary signals to feed to ancillary systems to protect the device and control the plasma. This note addresses the question of how much intrusion into the blanket system will be required to allow the measurements needed to provide the information required for plasma control. All diagnostics will require, at least, the same shielding designs as planned for ITER, while having the capability to maintain their calibration through very long pulses. Much work is required to define better the measurement needs and the quantity and quality of the measurements that will have to be made, and how they can be integrated into the other tokamak structures.

  5. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor

    SciTech Connect

    Singh, M. J.; De Esch, H. P. L.

    2010-01-15

    This paper describes the physics design of a 100 keV, 60 A H{sup -} accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

  6. R&D around a photoneutralizer-based NBI system (Siphore) in view of a DEMO Tokamak steady state fusion reactor

    NASA Astrophysics Data System (ADS)

    Simonin, A.; Achard, Jocelyn; Achkasov, K.; Bechu, S.; Baudouin, C.; Baulaigue, O.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; de Esch, H. P. L.; Fiorucci, D.; Fubiani, G.; Furno, I.; Futtersack, R.; Garibaldi, P.; Gicquel, A.; Grand, C.; Guittienne, Ph.; Hagelaar, G.; Howling, A.; Jacquier, R.; Kirkpatrick, M. J.; Lemoine, D.; Lepetit, B.; Minea, T.; Odic, E.; Revel, A.; Soliman, B. A.; Teste, P.

    2015-11-01

    Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η  >  60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (~3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

  7. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor.

    PubMed

    Singh, M J; De Esch, H P L

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated. PMID:20113091

  8. Construction and operation of parallel electric and magnetic field spectrometers for mass/energy resolved multi-ion charge exchange diagnostics on the Tokamak Fusion Test Reactor

    SciTech Connect

    Medley, S.S.; Roquemore, A.L.

    1998-07-01

    A novel charge exchange spectrometer using a dee-shaped region of parallel electric and magnetic fields was developed at the Princeton Plasma Physics Laboratory for neutral particle diagnostics on the Tokamak Fusion Test Reactor (TFTR). The E{parallel}B spectrometer has an energy range of 0.5{le}Athinsp(amu)Ethinsp(keV){le}600 and provides mass-resolved energy spectra of H{sup +}, D{sup +}, and T{sup +} (or {sup 3}He{sup +}) ion species simultaneously during a single discharge. The detector plane exhibits parallel rows of analyzed ions, each row containing the energy dispersed ions of a given mass-to-charge ratio. The detector consists of a large area microchannel plate (MCP) which is provided with three rectangular, semicontinuous active area strips, one coinciding with each of the mass rows for detection of H{sup +}, D{sup +}, and T{sup +} (or {sup 3}He{sup +}) and each mass row has 75 energy channels. To suppress spurious signals attending operation of the plate in the magnetic fringe field of the spectrometer, the MCP was housed in a double-walled iron shield with a wire mesh ion entrance window. Using an accelerator neutron generator, the MCP neutron detection efficiency was measured to be 1.7{times}10{sup {minus}3} and 6.4{times}10{sup {minus}3}thinspcounts/neutron/cm{sup 2} for 2.5 MeV-DD and 14 MeV-DT neutrons, respectively. The design and calibration of the spectrometer are described in detail, including the effect of MCP exposure to tritium, and results obtained during high performance D{endash}D operation on TFTR are presented to illustrate the performance of the E{parallel}B spectrometer. The spectrometers were not used during D{endash}T plasma operation due to the cost of providing the required radiation shielding. {copyright} {ital 1998 American Institute of Physics.}

  9. Burn Control Mechanisms in Tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  10. Linear optimal control of tokamak fusion devices

    SciTech Connect

    Kessel, C.E.; Firestone, M.A.; Conn, R.W.

    1989-05-01

    The control of plasma position, shape and current in a tokamak fusion reactor is examined using linear optimal control. These advanced tokamaks are characterized by non up-down symmetric coils and structure, thick structure surrounding the plasma, eddy currents, shaped plasmas, superconducting coils, vertically unstable plasmas, and hybrid function coils providing ohmic heating, vertical field, radial field, and shaping field. Models of the electromagnetic environment in a tokamak are derived and used to construct control gains that are tested in nonlinear simulations with initial perturbations. The issues of applying linear optimal control to advanced tokamaks are addressed, including complex equilibrium control, choice of cost functional weights, the coil voltage limit, discrete control, and order reduction. Results indicate that the linear optimal control is a feasible technique for controlling advanced tokamaks where the more common classical control will be severely strained or will not work. 28 refs., 13 figs.

  11. Tokamak power systems studies, FY 1985

    SciTech Connect

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  12. Superconducting magnet system for the TPX Tokamak

    SciTech Connect

    Hassenzahl, W.V.; Chaplin, M.R.; Heim, J.R.

    1993-09-15

    The Tokamak Physics Experiment (TPX) will be the first Tokamak using superconducting magnets for both the poloidal and toroidal field. It is designed for advanced Tokamak physics experiments in steady-state and long-pulse operation. The TPX superconducting magnets use an advanced cable-in-conduit conductor (CICC) design similar to that developed in support of the International Thermonuclear Experimental Reactor (ITER). The toroidal field magnets provide 4.0 T at 2.25 m with a stored energy of 1.05 GJ. The poloidal field magnets provide 18.0 V-s to ohmically start and control long burns of a 2.0 MA plasma.

  13. ITER tokamak device

    NASA Astrophysics Data System (ADS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    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, 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, fueling, 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., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  14. Tokamak and RFP ignition requirements

    SciTech Connect

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (n{tau}{sub 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{sub RAD} {approximately} 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 n{tau}{sub 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.

  15. Advanced fusion MHD power conversion using the CFAR (compact fusion advanced Rankine) cycle concept

    SciTech Connect

    Hoffman, M.A.; Campbell, R.; Logan, B.G.; Lawrence Livermore National Lab., CA )

    1988-10-01

    The CFAR (compact fusion advanced Rankine) cycle concept for a tokamak reactor involves the use of a high-temperature Rankine cycle in combination with microwave superheaters and nonequilibrium MHD disk generators to obtain a compact, low-capital-cost power conversion system which fits almost entirely within the reactor vault. The significant savings in the balance-of-plant costs are expected to result in much lower costs of electricity than previous concepts. This paper describes the unique features of the CFAR cycle and a high- temperature blanket designed to take advantage of it as well as the predicted performance of the MHD disk generators using mercury seeded with cesium. 40 refs., 8 figs., 3 tabs.

  16. Ideal magnetohydrodynamic simulations of low beta compact toroid injection into a hot strongly magnetized plasma

    SciTech Connect

    Liu, Wei; Hsu, Scott; Li, Hui

    2009-01-01

    We present results from three-dimensional ideal magnetohydrodynamic simulations of low {beta} compact toroid (CT) injection into a hot strongly magnetized plasma, with the aim of providing insight into CT fueling of a tokamak with parameters relevant for ITER (International Thermonuclear Experimental Reactor). A regime is identified in terms of CT injection speed and CT-to-background magnetic field ratio that appears promising for precise core fueling. Shock-dominated regimes, which are probably unfavorable for tokamak fueling, are also identified. The CT penetration depth is proportional to the CT injection speed and density. The entire CT evolution can be divided into three stages: (1) initial penetration, (2) compression in the direction of propagation and reconnection, and (3) coming to rest and spreading in the direction perpendicular to injection. Tilting of the CT is not observed due to the fast transit time of the CT across the background plasma.

  17. Compact stellarators with modular coils

    PubMed Central

    Garabedian, P. R.

    2000-01-01

    Compact stellarator designs with modular coils and only two or three field periods are now available; these designs have both good stability and quasiaxial symmetry providing adequate transport for a magnetic fusion reactor. If the bootstrap current assumes theoretically predicted values a three field period configuration is optimal, but if that net current turns out to be lower, a device with two periods and just 12 modular coils might be better. There are also attractive designs with quasihelical symmetry and four or five periods whose properties depend less on the bootstrap current. Good performance requires that there be a satisfactory magnetic well in the vacuum field, which is a property lacking in a stellarator-tokamak hybrid that has been proposed for a proof of principle experiment. In this paper, we present an analysis of stability for these configurations that is based on a mountain pass theorem asserting that, if two solutions of the problem of magnetohydrodynamic equilibrium can be found, then there has to be an unstable solution. We compare results of our theory of equilibrium, stability, and transport with recently announced measurements from the large LHD experiment in Japan. PMID:10899993

  18. OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS

    SciTech Connect

    LIN-LIU,YR; STAMBAUGH,RD

    2002-11-01

    OAK A271 OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS. The dependence of the ideal ballooning {beta} limit on aspect ratio, A, and elongation {kappa} is systematically explored for nearly 100% bootstrap current driven tokamak equilibria in a wide range of the shape parameters (A = 1.2-7.0, {kappa} = 1.5-6.0 with triangularity {delta} = 0.5). The critical {beta}{sub N} is shown to be optimal at {kappa} = 3.0-4.0 for all A studied and increases as A decreases with a dependence close to A{sup -0.5}. The results obtained can be used as a theoretical basis for the choice of optimum aspect ratio and elongation of next step burning plasma tokamaks or tokamak reactors.

  19. Control of Dust Inventory in Tokamaks

    SciTech Connect

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Pitcher, C. S.; Taylor, N.; Furlan, J.

    2008-09-07

    Particles with sizes ranging from 100 nm to 100 {mu}m are produced in tokamaks by the interaction of the plasma with the first wall materials and divertor. Dust has not yet been of a major concern in existing tokamaks mainly because their quantities are small and these devices are not nuclear facilities. However, in ITER and in future reactors, they could represent operational and potential safety issues. The aim of this paper is thus to describe the dust creation processes in the tokamak environment. The diagnostics and removal techniques that are needed to be implemented to measure and minimise the dust inventory are also presented. The integration of these techniques into a tokamak environment is also discussed.

  20. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  1. Ripple induced trapped particle loss in tokamaks

    SciTech Connect

    White, R.B.

    1996-05-01

    The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks.

  2. MHD stability of tokamak plasmas

    SciTech Connect

    Chance, M.S. Sun, Y.C.; Jardin, S.C.; Kessel, C.E.; Okabayashi, M.

    1992-08-01

    This paper will give an overview of the some of the methods which are used to simulate the ideal MHD properties of tokamak plasmas. A great deal of the research in this field is necessarily numerical and the substantial progress made during the past several years has roughly paralleled the continuing availability of more advanced supercomputers. These have become essential to accurately model the complex configurations necessary for achieving MHD stable reactor grade conditions. Appropriate tokamak MHD equilibria will be described. Then the stability properties is discussed in some detail, emphasizing the difficulties of obtaining stable high {beta} discharges in plasmas in which the current is mainly ohmically driven and thus demonstrating the need for tailoring the current and pressure profiles of the plasma away from the ohmic state. The outline of this paper will roughly follow the physics development to attain the second region of stability in the PBX-M device at The Princeton Plasmas Physics Laboratory.

  3. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

    2002-03-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

  4. NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Aug 15, 2002 to Nov. 15, 2002) - DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Lawrence Townsend; Martin Williamson; Rupy Sawhney; Jacob Fife

    2002-12-15

    The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. This report covers the ninth quarter of the project. The three reactor concept teams have completed initial plant concept development, evaluation and layout. A significant design effort has proceeded with substantial change and evolution from original ideas. The concepts have been reviewed by the industry participants and improvements have been implemented. The third phase, industrial engineering simulation of reactor fabrication has begun.

  5. Comparison of tokamak burn cycle options

    SciTech Connect

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options.

  6. Development of a repetitive compact torus injector

    NASA Astrophysics Data System (ADS)

    Onchi, Takumi; McColl, David; Dreval, Mykola; Rohollahi, Akbar; Xiao, Chijin; Hirose, Akira; Zushi, Hideki

    2013-10-01

    A system for Repetitive Compact Torus Injection (RCTI) has been developed at the University of Saskatchewan. CTI is a promising fuelling technology to directly fuel the core region of tokamak reactors. In addition to fuelling, CTI has also the potential for (a) optimization of density profile and thus bootstrap current and (b) momentum injection. For steady-state reactor operation, RCTI is necessary. The approach to RCTI is to charge a storage capacitor bank with a large capacitance and quickly charge the CT capacitor bank through a stack of integrated-gate bipolar transistors (IGBTs). When the CT bank is fully charged, the IGBT stack will be turned off to isolate banks, and CT formation/acceleration sequence will start. After formation of each CT, the fast bank will be replenished and a new CT will be formed and accelerated. Circuits for the formation and the acceleration in University of Saskatchewan CT Injector (USCTI) have been modified. Three CT shots at 10 Hz or eight shots at 1.7 Hz have been achieved. This work has been sponsored by the CRC and NSERC, Canada.

  7. Burn Control Mechanisms in Tokamaks

    NASA Astrophysics Data System (ADS)

    Hill, M. A.; Stacey, W. M.

    2015-11-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

  8. Tokamak Systems Code

    SciTech Connect

    Reid, R.L.; Barrett, R.J.; Brown, T.G.; Gorker, G.E.; Hooper, R.J.; Kalsi, S.S.; Metzler, D.H.; Peng, Y.K.M.; Roth, K.E.; Spampinato, P.T.

    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.

  9. Microbial and Physicochemical Characteristics of Compact Anaerobic Ammonium-Oxidizing Granules in an Upflow Anaerobic Sludge Blanket Reactor

    PubMed Central

    Ni, Bing-Jie; Hu, Bao-Lan; Fang, Fang; Xie, Wen-Ming; Kartal, Boran; Liu, Xian-Wei; Sheng, Guo-Ping; Jetten, Mike; Zheng, Ping; Yu, Han-Qing

    2010-01-01

    Anaerobic ammonium oxidation (anammox) is a promising new process to treat high-strength nitrogenous wastewater. Due to the low growth rate of anaerobic ammonium-oxidizing bacteria, efficient biomass retention is essential for reactor operation. Therefore, we studied the settling ability and community composition of the anaerobic ammonium-oxidizing granules, which were cultivated in an upflow anaerobic sludge blanket (UASB) reactor seeded with aerobic granules. With this seed, the start-up period was less than 160 days at a NH4+-N removal efficiency of 94% and a loading rate of 0.064 kg N per kg volatile suspended solids per day. The formed granules were bright red and had a high settling velocity (41 to 79 m h−1). Cells and extracellular polymeric substances were evenly distributed over the anaerobic ammonium-oxidizing granules. The high percentage of anaerobic ammonium-oxidizing bacteria in the granules could be visualized by fluorescent in situ hybridization and electron microscopy. The copy numbers of 16S rRNA genes of anaerobic ammonium-oxidizing bacteria in the granules were determined to be 4.6 × 108 copies ml−1. The results of this study could be used for a better design, shorter start-up time, and more stable operation of anammox systems for the treatment of nitrogen-rich wastewaters. PMID:20190088

  10. Compact, Lightweight Adsorber and Sabatier Reactor for CO2 Capture and Reduction for Consumable and Propellant Production

    NASA Technical Reports Server (NTRS)

    Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Busby, Stacy A.; Abney, Morgan B.; Perry, Jay L.; Knox, James C.

    2012-01-01

    The utilization of CO2 to produce (or recycle) life support consumables, such as O2 and H2O, and to generate propellant fuels is an important aspect of NASA's concept for future, long duration planetary exploration. One potential approach is to capture and use CO2 from the Martian atmosphere to generate the consumables and propellant fuels. Precision Combustion, Inc. (PCI), with support from NASA, continues to develop its regenerable adsorber technology for capturing CO2 from gaseous atmospheres (for cabin atmosphere revitalization and in-situ resource utilization applications) and its Sabatier reactor for converting CO2 to methane and water. Both technologies are based on PCI's Microlith(R) substrates and have been demonstrated to reduce size, weight, and power consumption during CO2 capture and methanation process. For adsorber applications, the Microlith substrates offer a unique resistive heating capability that shows potential for short regeneration time and reduced power requirements compared to conventional systems. For the Sabatier applications, the combination of the Microlith substrates and durable catalyst coating permits efficient CO2 methanation that favors high reactant conversion, high selectivity, and durability. Results from performance testing at various operating conditions will be presented. An effort to optimize the Sabatier reactor and to develop a bench-top Sabatier Development Unit (SDU) will be discussed.

  11. Magnetic confinement experiment. I: Tokamaks

    SciTech Connect

    Goldston, R.J.

    1995-08-01

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 2.5x greater than ELM`ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices.

  12. Simulation of the alpha particle heating and the helium ash source in an International Thermonuclear Experimental Reactor-like tokamak with an internal transport barrier

    SciTech Connect

    Ye, Lei Guo, Wenfeng; Xiao, Xiaotao; Dai, Zongliang; Wang, Shaojie

    2014-12-15

    A guiding center orbit following code, which incorporates a set of non-singular coordinates for orbit integration, was developed and applied to investigate the alpha particle heating in an ITER-like tokamak with an internal transport barrier. It is found that a relatively large q (safety factor) value can significantly broaden the alpha heating profile in comparison with the local heating approximation; this broadening is due to the finite orbit width effects; when the orbit width is much smaller than the scale length of the alpha particle source profile, the heating profile agrees with the source profile, otherwise, the heating profile can be significantly broadened. It is also found that the stagnation particles move to the magnetic axis during the slowing-down process, thus the effect of stagnation orbits is not beneficial to the helium ash removal. The source profile of helium ash is broadened in comparison with the alpha source profile, which is similar to the heating profile.

  13. Modeling of compact loop antennas

    SciTech Connect

    Baity, F.W.

    1987-01-01

    A general compact loop antenna model which treats all elements of the antenna as lossy transmission lines has been developed. In addition to capacitively-tuned resonant double loop (RDL) antennas the model treats stub-tuned resonant double loop antennas. Calculations using the model have been compared with measurements on full-scale mockups of resonant double loop antennas for ATF and TFTR in order to refine the transmission line parameters. Results from the model are presented for RDL antenna designs for ATF, TFTR, Tore Supra, and for the Compact Ignition Tokamak (CIT).

  14. Modeling of compact loop antennas

    NASA Astrophysics Data System (ADS)

    Baity, F. W.

    A general compact loop antenna model which treats all elements of the antenna as lossy transmission lines has been developed. In addition to capacitively-tuned resonant double loop (RDL) antennas the model treats stub-tuned resonant double loop antennas. Calculations using the model have been compared with measurements on full-scale mockups of resonant double loop antennas for ATF and TFTR in order to refine the transmission line parameters. Results from the model are presented for RDL antenna designs for ATF, TFTR, Tore Supra, and for the Compact Ignition Tokamak (CIT).

  15. Modeling of compact loop antennas

    NASA Astrophysics Data System (ADS)

    Baity, F. W.

    1987-09-01

    A general compact loop antenna model which treats all elements of the antenna as lossy transmission lines has been developed. In addition to capacitively tuned resonant double loop (RDL) antennas, the model treats sub-tuned RDL antennas. Calculations using the model have been compared with measurements on full-scale mock-ups of RDL antennas for ATF and TFTR in order to refine the transmission line parameters. Results from the model are presented for RDL antenna designs for ATF, TFTR, Tore Supra, and the Compact Ignition Tokamak (CIT).

  16. ORNL compact loop antenna design for TFTR and Tore Supra

    SciTech Connect

    Taylor, D.J.; Baity, F.W.; Bryan, W.E.; Hoffman, D.J.; McIlwain, R.L. ); Ray, J.M. )

    1987-01-01

    The goal supplemental ion cyclotron resonance heating (ICRH) of fusion plasma is to deliver power at high efficiencies deep within the plasma. The technology for fast-wave ICRH has reached the point of requiring proof-of-performance'' demonstration of specific antenna configurations of specific antenna configurations and their mechanical adequacy for operating in a fusion environment. Oak Ridge National Laboratory (ORNL) has developed the compact loop antenna concept based on a resonant double loop (RDL) configuration for use in both Tokamak Fusion Test Reactor (TFTR) and the Tore Supra ICRH programs. A description and a comparison of the technologies developed in the two designs are presented. The electrical circuit and the mechanical philosophy employed are the same for both antennas, but different operating environments result in substantial differences in the design of specific components. The ORNL TFTR antenna is designed to deliver 4 MW over a 2-s pulse, and the ORNL Tore Supra antenna is designed for 4 MW and essentially steady-state conditions. The TFTR design embodies the first operations compact RDL antenna, and the Tore Supra antenna extends the technology to an operational duty cycle consistent with reactor-relevant applications. 7 refs., 5 figs.

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

    SciTech Connect

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

    1995-04-01

    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.

  18. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  19. Critical issues for modeling dust transport in tokamaks.

    PubMed

    Bacharis, Minas; Coppins, Michael; Allen, John E

    2010-08-01

    Dust produced in tokamaks is an important issue for fusion. Dust particles can introduce health and safety issues when in the same time can have an impact on reactor performance. Apart from the associated problems there are also potential benefits that make the better understanding of their behavior important. In this work the dust transport code Dust in TOKamakS will be used to explore the effect that variations in the plasma background and the physical model, describing the plasma-dust interaction, have on their predicted trajectories. PMID:20866922

  20. Neoclassical diffusion of heavy impurities in a rotating tokamak plasma

    SciTech Connect

    Wong, K.L.; Cheng, C.Z.

    1987-08-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle simulation is carried out and the results offer a qualitative explanation for some experimental data from the Tokamak Test Reactor (TFTR). 13 refs., 2 figs.

  1. Environmental Assessment and Finding of No Significant Impact: The National Compact Stellarator Experiment at the Princeton Plasma Physics Laboratory

    SciTech Connect

    N /A

    2002-10-25

    If the United States is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels, the eventual depletion of fissionable uranium used in present-day nuclear reactors, and the limitations of solar, hydro and wind alternatives. Fusion energy, the power source of the sun and other stars, has the potential to become a major source of energy for the future. Power from fusion would provide substantially reduced environmental impacts as compared with current forms of energy generation. Thus, the United States and other countries around the world continue to pursue development of fusion energy as one of a number of potential power sources for the long term. Fusion research, using various machine configurations, has been proceeding since the early 1950's, and significant progress has been achieved in performance and in understanding of the underlying physics. For most of this period, fusion machines called stellarators and tokamaks, which are toroidal (doughnut-shaped) devices, have been most frequently used to conduct experiments for producing controlled nuclear fusion. It is now desirable to take a next step in the fusion development program, by providing an experimental device to investigate the attractiveness of a compact stellarator as the basis for a fusion power reactor. This concept has the potential to build upon advances in understanding of stellarators and tokamaks, and to combine the best features of both. The goal is to build a compact stellarator that would be smaller than conventional stellarators and operate more efficiently than previous tokamaks. Such a device would broaden our understanding of magnetic fusion science while contributing to the development of a potentially attractive fusion reactor solution that may have cost advantages over other fusion concepts.

  2. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0. 8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    SciTech Connect

    Dickens, J.K.; Hill, N.W.; Hou, F.S.; McConnell, J.W.; Spencer, R.R.; Tsang, F.Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

  3. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    NASA Astrophysics Data System (ADS)

    Dickens, J. K.; Hill, N. W.; Hou, F. S.; McConnell, J. W.; Spencer, R. R.; Tsang, F. Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

  4. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  5. Broadband measurements of electron cyclotron emission in TFTR (Tokamak Fusion Test Reactor) using a quasi-optical light collection system and a polarizing Michelson interferometer

    SciTech Connect

    Stauffer, F.J.; Boyd, D.A.; Cutler, R.C.; Diesso, M.; McCarthy, M.P.; Montague, J.; Rocco, R.

    1988-04-01

    For the past three years, a Fourier transform spectrometer diagnostic system, employing a fast-scanning polarizing Michelson interferometer, has been operating on the TFTR tokamak at Princeton Plasma Physics Laboratory. It is used to measure the electron cyclotron emission spectrum over the range 2.5 to 18 cm/sup /minus/1/ (75-540 GHz) with a resolution of 0.123 cm/sup /minus/1/(3.7 GHz), at a rate of 72 spectra per second. The quasi-optical system for collecting the light and transporting it through the interferometer to the detector has been designed using the concepts of both Gaussian and geometrical optics in order to produce a system that is efficient over the entire spectral range. The commerical Michelson interferometer was custom-made for this project and is at the state of the art for this type of specialized instrument. Various pre-installation and post-installation tests of the optical system and the interferometer were performed and are reported here. An error propagation analysis of the absolute calibration process is given. Examples of electron cyclotron emission spectra measured in two polarization directions are given, and electron temperature profiles derived from each of them are compared. 34 refs., 17 figs.

  6. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  7. Adult Compacts.

    ERIC Educational Resources Information Center

    Further Education Unit, London (England).

    This bulletin focuses on adult compacts, three-way agreements among employers, potential employees, and trainers to provide the right kind of quality training to meet the employers' requirements. Part 1 is an executive summary of a report of the Adult Compacts Project, which studied three adult compacts in Birmingham and Loughborough, England, and…

  8. Advanced tokamak research on the DIII-D tokamak

    SciTech Connect

    Chan, V.S.

    1994-01-01

    The objective of the planned research in advanced tokamak development on DIII-D at General Atomics, San Diego, USA. is to establish improved tokamak operation through significant improvements in the stability factor, confinement quality, and bootstrap current fraction using localized radio frequency (rf) current profile control, rf and neutral beam heating for pressure profile control, as well as control of plasma rotation and optimization of the plasma boundary conditions. Recent research results in H-mode confinement, modifications of current profiles to achieve higher confinement and higher {beta}, a new regime of improved confinement (VH-mode), and rf noninductive current drive are encouraging. In this talk, arguments will be presented supporting the need for improved performance in tokamak reactors. Experimentally observed advanced performance regimes on DIII-D will be discussed. Confinement improvement up to H = 4, where H is the ratio of energy confinement time to the ITER89-P scaling H{triple_bond} {Tau}{sub E}/{Tau}{sub E-ITER89-P}, has been achieved. In other discharges {beta}{sub N} = {beta}/(I/aB),[%-m{center_dot}{Tau}/MA] {approx_gt} 6 has been obtained. These values have so far been achieved transiently and independently. Techniques, will be described which can extend the high performance to quasi-steady-state and sustain the high H and {beta}{sub N} values simultaneously. Two high performance regimes, one in first stable regime and the other in second stable regime, have been simulated br self-consistently evolving a magnetohydrodynamic (MHD) equilibrium-transport code. Finally, experimental program plans and outstanding important physics issues will be discussed.

  9. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  10. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    NASA Astrophysics Data System (ADS)

    Azizov, E. A.

    2012-02-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.

  11. Final Report (1994 to 1996) Diagnostic of the Spatial and Velocity Distribution of Alpha Particles in Tokamak Fusion Reactor using Beat-wave Generated Lower Hybrid Wave

    SciTech Connect

    Hwang, D.Q.; Horton, R.D.; Evans, R.W.

    1999-06-03

    The alpha particles in a fusion reactor play a key role in the sustaining the fusion reaction. It is the heating provided by the alpha particles that help a fusion reactor operating in the ignition regime. It is, therefore, essential to understand the behavior of the alpha population both in real space and velocity space in order to design the optimal confinement device for fusion application. Moreover, the alphas represent a strong source of free energy that may generate plasma instabilities. Theoretical studies has identified the Toroidal Alfven Eigenmode (TAE) as an instability that can be excited by the alpha population in a toroidal device. Since the alpha has an energy of 3.5 MeV, a good confinement device will retain it in the interior of the plasma. Therefore, alpha measurement system need to probe the interior of a high density plasma. Due to the conducting nature of a plasma, wave with frequencies below the plasma frequency can not penetrate into the interior of the plasma where the alphas reside. This project uses a wave that can interact with the perpendicular motion of the alphas to probe its characteristics. However, this wave (the lower hybrid wave) is below the plasma frequency and can not be directly launched from the plasma edge. This project was designed to non-linearly excite the lower hybrid in the interior of a magnetized plasma and measure its interaction with a fast ion population.

  12. Initial results of systems analysis of the ETR/ITER (Engineering Test Reactor/International Thermonuclear Experimental Reactor) design space

    SciTech Connect

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

    1987-01-01

    Preliminary versions of the Engineering Test Reactor (ETR) systems code TETRA (Tokamak Engineering Test Reactor Analysis), which determines design solutions by the method of constrained optimization, are used to characterize the International Thermonuclear Experimental Reactor (ITER) and its design parameter space. We find that the physics objectives of high ignition margin and high plasma current lead to minimum size at relatively low aspect ratios (A = 2.5-3.0), while the engineering objective of high neutron wall load (W/sub L/ /approx gt/ 1.0 MW/m/sup 2/) leads to minimum size at higher A (/approximately/3.5). For minimum-size ITERs, the optimal toroidal field coil (TFC) designs fall within a narrow range of maximum fields (10-11 T) with R varying over only a few percent despite a factor of two change in the winding pack current density J/sub wp/. The major radius of the design is found to be sensitive to changes in elongation, inboard distances (such as plasma scrape-off), inductive flux capability, plasma temperature, beta limit, and ignition margin. A preliminary characterization of the US ITER designs with plasma current I/sub p/ > 15 MA and R < 4.5 m has been obtained by combining the engineering assumptions for devices such as the Tokamak Ignition/Burn Engineering Reactor (TIBER) with the physics assumptions for devices such as the Compact Ignition Tokamak (CIT) and the Next European Torus (NET). These devices can accommodate a range of full- to reduced-bore, driven (Q < 10), steady-state plasmas for the engineering phase that produces high neutron wall load and fluence. 12 refs., 4 figs., 3 tabs.

  13. Improvement of tokamak performance by injection of electrons

    SciTech Connect

    Ono, Masayuki

    1992-12-01

    Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas.

  14. Localized Ballooning Modes in Compact Quasiaxially Symmetric Stellarators

    SciTech Connect

    M.H. Redi; J. Canik; R.L. Dewar; E.D. Fredrickson; W.A. Cooper; J.L. Johnson; S. Klasky

    2001-06-14

    Understanding of ballooning mode stability boundaries may lead to performance improvement of toroidal devices through control of plasma disruptions. Toroidally localized ballooning modes have been found as precursors to high-beta plasma disruptions on the Tokamak Fusion Test Reactor (TFTR) arising in conditions of n=1 kink mode asymmetry. Recent optimization has shown that magnetohydrodynamic (MHD) stability as well as good particle confinement are likely to be achievable in the National Compact Stellarator Experiment (NCSX), a compact, quasiaxially symmetric stellarator (QAS) for values of the plasma near beta = 4%. The configuration, with a major radius of 1.42 m, an aspect ratio of 4.4, a toroidal magnetic field 1.2-1.7 T and 6 MW of neutral-beam heating, is stable to MHD instabilities, and is expected to be limited by high-n kink and ballooning modes. This paper describes the ballooning eigenvalue isosurfaces for NCSX, the first step in an examination of the kinetic stabilization of the ballooning beta limit using a hybrid WKB approach.

  15. Modular tokamak magnetic system

    DOEpatents

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  16. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  17. Prospects for toroidal fusion reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.D.

    1994-06-01

    Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, {approximately}2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges.

  18. VIBRATION COMPACTION

    DOEpatents

    Hauth, J.J.

    1962-07-01

    A method of compacting a powder in a metal container is described including the steps of vibrating the container at above and below the resonant frequency and also sweeping the frequency of vibration across the resonant frequency several times thereby following the change in resonant frequency caused by compaction of the powder. (AEC)

  19. Tritium Retention and Removal in Tokamaks

    SciTech Connect

    Skinner, Charles H.

    2009-02-19

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy. Tritium is an important source term in safety assessments, it is expensive and in short supply. Tritium can be continuously retained in a tokamak by codeposition with eroded carbon or beryllium and JET and TFTR with carbon plasma facing components showed a tritium retention level that would be unacceptable in ITER or future fusion reactors. Asdex-U and Alcator C-mod have shown reduced hydrogenic retention with tungsten clad and molybdenum plasma facing components. Once the tritium inventory approaches the administrative limit, tritium must be removed to permit continued D-T plasma operations. Several candidate techniques are being considered and need to be proven at a relevant speed and efficiency in contemporary tokamaks. Projections for ITER are discussed.

  20. The physics of tokamak start-up

    SciTech Connect

    Mueller, D.

    2013-05-15

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  1. A low aspect ratio tokamak transmutation system

    NASA Astrophysics Data System (ADS)

    Qiu, L. J.; Wu, Y. C.; Xiao, B. J.; Xu, Q.; Huang, Q. Y.; Wu, B.; Chen, Y. X.; Xu, W. N.; Chen, Y. P.; Liu, X. P.

    2000-03-01

    A low aspect ratio tokamak transmutation system is proposed as an alternative application of fusion energy on the basis of a review of previous studies. This system includes: (1) a low aspect ratio tokamak as fusion neutron driver, (2) a radioactivity-clean nuclear power system as blanket, and (3) a novel concept of liquid metal centre conductor post as part of the toroidal field coils. In the conceptual design, a driver of 100 MW fusion power under 1 MW/m2 neutron wall loading can transmute the amount of high level waste (including minor actinides and fission products) produced by ten standard pressurized water reactors of 1 GW electrical power output. Meanwhile, the system can produce tritium on a self-sustaining basis and an output of about 2 GW of electrical energy. After 30 years of operation, the biological hazard potential level of the whole system will decrease by two orders of magnitude.

  2. Nonambipolar transport by trapped particles in tokamaks.

    PubMed

    Park, Jong-Kyu; Boozer, Allen H; Menard, Jonathan E

    2009-02-13

    Small nonaxisymmetric perturbations of the magnetic field can greatly change the performance of tokamaks through nonambipolar transport. A number of theories have been developed, but the predictions were not consistent with experimental observations in tokamaks. This Letter provides a resolution, with a generalized analytic treatment of the nonambipolar transport. It is shown that the discrepancy between theory and experiment can be greatly reduced by two effects: (1) the small fraction of trapped particles for which the bounce and precession rates resonate; (2) the nonaxisymmetric variation in the field strength along the perturbed magnetic field lines rather than along the unperturbed magnetic field lines. The expected sensitivity of the International Thermonuclear Experimental Reactor to nonaxisymmetries is also discussed. PMID:19257595

  3. The physics of tokamak start-upa)

    NASA Astrophysics Data System (ADS)

    Mueller, D.

    2013-05-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

  4. ADX: a high field, high power density, advanced divertor and RF tokamak

    NASA Astrophysics Data System (ADS)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  5. Catalytic reactor

    DOEpatents

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  6. Tokamak foundation in USSR/Russia 1950-1990

    NASA Astrophysics Data System (ADS)

    Smirnov, V. P.

    2010-01-01

    In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

  7. Resistive edge mode instability in stellarator and tokamak geometries

    SciTech Connect

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-15

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  8. Texas Experimental Tokamak

    SciTech Connect

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported.

  9. Completely bootstrapped tokamak

    SciTech Connect

    Weening, R.H. ); Boozer, A.H. )

    1992-01-01

    Numerical simulations of the evolution of large-scale magnetic fields have been developed using a mean-field Ohm's law. The Ohm's law is coupled to a {Delta}{prime} stabilty analysis and a magnetic island growth equation in order to simulate the behavior of tokamak plasmas that are subject to tearing modes. In one set of calculations, the magnetohydrodynamic (MHD)-stable regime of the tokamak is examined via the construction of an {ital l}{sub {ital i}} -{ital q}{sub {ital a}} diagram. The results confirm previous calculations that show that tearing modes introduce a stability boundary into the {ital l}{sub {ital i}} -{ital q}{sub {ital a}} space. In another series of simulations, the interaction between tearing modes and the bootstrap current is investigated. The results indicate that a completely bootstrapped tokamak may be possible, even in the absence of any externally applied loop voltage or current drive.

  10. First Results of the ETE Spherical Tokamak

    NASA Astrophysics Data System (ADS)

    Ludwig, G. O.; del Bosco, E.; Ferreira, J. G.; Barroso, J. J.; Berni, L. A.; Oliveira, R. M.

    2001-10-01

    First tokamak plasma discharges were obtained in the ETE Spherical Tokamak Experiment (Experimento Tokamak Esferico) by November 2000. ETE is a medium size machine (major radius R=0.30m) with a compact design and good access for diagnostics. During the first phase of operation a plasma current of 200kA (t=15ms) will be produced in a 1.5 aspect ratio configuration with a toroidal magnetic field up to 0.4T. The ultimate values are limited by mechanical stresses in the joints of the toroidal field coil (B<0.8T), and by stresses and heating in the solenoid (0.24Wb, 30kA, 0.2s) for a maximum plasma current of about 400kA. Presently the experiments are focused on plasma formation, vacuum conditioning and diagnostics implementation. The machine was constructed in accordance with stringent design specifications. The assembly has an overall precision better than 2mm. Vacuum conditioning is being improved with baking, glow discharge cleaning and usual tokamak operation. Breakdown is easily obtained even without the pre-ionization provided by a hot filament electron emitter and by an ultraviolet lamp. Preliminary measurements of stray magnetic fields were carried out and eddy current effects are being modeled. Energy of the capacitor banks is being continuously increased to achieve the design parameters. A fast neutral lithium beam probe for plasma edge studies and a 6.7GHz high-power monotron for pre-ionization and electron cyclotron resonance plasma heating experiments are under development.

  11. Formation of a 100-kA tokamak discharge in the Princeton large torus by lower hybrid waves

    SciTech Connect

    Jobes, F.; Stevens, J.; Bell, R.; Bernabei, S.; Cavallo, A.; Chu, T.K.; Cohen, S.; Denne, B.; Efthimion, P.; Hinnov, E.

    1984-03-01

    The development of non-inductive current drive is of great importance in establishing the tokamak as a long-pulse or steady-state fusion reactor. Lower hybrid waves, carrying 200 kW of power at 800 MHz, have been launched into the PLT tokamak to initiate and drive the discharge current to a level in excess of 100 kA.

  12. Evaluation of the operational parameters for NBI-driven fusion in low-gain tokamaks with two-component plasma

    NASA Astrophysics Data System (ADS)

    Chirkov, A. Yu.

    2015-09-01

    Low gain (Q ~ 1) fusion plasma systems are of interest for concepts of fusion-fission hybrid reactors. Operational regimes of large modern tokamaks are close to Q  ≈  1. Therefore, they can be considered as prototypes of neutron sources for fusion-fission hybrids. Powerful neutral beam injection (NBI) can support the essential population of fast particles compared with the Maxwellial population. In such two-component plasma, fusion reaction rate is higher than for Maxwellian plasma. Increased reaction rate allows the development of relatively small-size and relatively inexpensive neutron sources. Possible operating regimes of the NBI-heated tokamak neutron source are discussed. In a relatively compact device, the predictions of physics of two-component fusion plasma have some volatility that causes taking into account variations of the operational parameters. Consequent parameter ranges are studied. The feasibility of regimes with Q  ≈  1 is shown for the relatively small and low-power system. The effect of NBI fraction in total heating power is analyzed.

  13. A Midsize Tokamak As Fast Track To Burning Plasmas

    SciTech Connect

    E. Mazzucato

    2010-07-14

    This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.

  14. Solenoid-free plasma start-up in spherical tokamaks

    NASA Astrophysics Data System (ADS)

    Raman, R.; Shevchenko, V. F.

    2014-10-01

    The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

  15. Proposals for an influential role of small tokamaks in mainstream fusion physics and technology research

    NASA Astrophysics Data System (ADS)

    Van Oost, G.; Del Bosco, E.; Gryaznevich, M. P.; Malaquias, A.; Mank, G.

    2006-12-01

    Small tokamaks may significantly contribute to the better understanding of phenomena in a wide range of fields such as plasma confinement and energy transport; plasma stability in different magnetic configurations; plasma turbulence and its impact on local and global plasma parameters; processes at the plasma edge and plasma-wall interaction; scenarios of additional heating and non-inductive current drive; new methods of plasma profile and parameter control; development of novel plasma diagnostics; benchmarking of new numerical codes and so on. Furthermore, due to the compactness, flexibility, low operation costs and high skill of their personnel small tokamaks are very convenient to develop and test new materials and technologies, which because of the risky nature cannot be done in large machines without preliminary studies. Small tokamaks are suitable and important for broad international cooperation, providing the necessary environment and manpower to conduct dedicated joint research programmes. In addition, the experimental work on small tokamaks is very appropriate for the education of students, scientific activities of post-graduate students and for the training of personnel for large tokamaks. All these tasks are well recognised and reflected in documents and understood by the large tokamak teams. Recent experimental results will be presented of contributions to mainstream fusion physics and technology research on small tokamaks involved in the IAEA Coordinated Research Project "Joint Research using small tokamaks", started in 2004.

  16. Tearing Modes in Tokamaks

    NASA Astrophysics Data System (ADS)

    White, R. B.

    2008-05-01

    This lecture gives a basic introduction to magnetic £elds, magnetic surface destruction, toroidal equilibrium and tearing modes in a tokamak, including the linear and nonlinear development of these modes and their modi£cation by current drive and bootstrap current, and sawtooth oscillations and disruptions.

  17. Ureilite compaction

    NASA Astrophysics Data System (ADS)

    Walker, D.; Agee, C. B.

    1988-03-01

    Ureilite meteorites show the simple mineralogy and compact recrystallized textures of adcumulate rock or melting residues. A certain amount of controversy exists about whether they are in fact adcumulate rocks or melting residues and about the nature of the precursor liquid or solid assemblage. The authors undertook a limited experimental study which made possible the evaluation of the potential of the thermal migration mechanism (diffusion on a saturation gradient) for forming ureilite-like aggregates from carbonaceous chondrite precursors. They find that the process can produce compact recrystallized aggregates of silicate crystals which do resemble the ureilities and other interstitial-liquid-free adcumulate rocks in texture.

  18. 20 years of research on the Alcator C-Mod tokamak

    SciTech Connect

    Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; and others

    2014-11-15

    The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

  19. Transport in gyrokinetic tokamaks

    SciTech Connect

    Mynick, H.E.; Parker, S.E.

    1995-01-01

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ``gyrokinetic tokamak`` is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/{rho}{sub s} {approx_gt} 64) with minor radius, with current, and with a/{rho}{sub s} are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (N{sub k} {approximately} 10) of k dominate the transport, and for each, only a handful (N{sub p} {approximately} 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients M{sub kpq} governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time.

  20. High Beta Tokamaks

    SciTech Connect

    Cowley, S.

    1998-11-14

    Perhaps the ideal tokamak would have high {beta} ({beta} {approx}> 1) and classical confinement. Such a tokamak has not been found, and we do not know if one does exist. We have searched for such a possibility, so far without success. In 1990, we obtained analytic equilibrium solutions for large aspect ratio tokamaks at {beta} {approx} {Omicron}(1) [1]. These solutions and the extension at high {beta} poloidal to finite aspect ratio [2] provided a basis for the study of high {beta} tokamaks. We have shown that these configurations can be stable to short scale MHD modes [3], and that they have reduced neoclassical transport [4]. Microinstabilities (such as the {del}T{sub i} mode) seem to be stabilized at high {beta} [5] - this is due to the large local shear [3] and the magnetic well. We have some concerns about modes associated with the compressional branch which may appear at high {beta}. Bill Dorland and Mike Kotschenreuther have studied this issue and our concerns may be unfounded. It is certainly tantalizing, especially given the lowered neoclassical transport values, that these configurations could have no microinstabilities and, one could assume, no anomalous transport. Unfortunately, while this work is encouraging, the key question for high {beta} tokamaks is the stability to large scale kink modes. The MHD {beta} limit (Troyon limit) for kink modes at large aspect ratio is problematically low. There is ample evidence from computations that the limit exists. However, it is not known if stable equilibria exist at much higher {beta}--none have been found. We have explored this question in the asymptotic high {beta} poloidal limit. Unfortunately, we are unable to find stable equilibrium and also unable to show that they don't exist. The results of these calculations will be published when a more definitive answer is found.

  1. Considerations of the high magnetic field tokamak path on the approach to fusion energy

    NASA Astrophysics Data System (ADS)

    Marmar, Earl

    2015-11-01

    This tutorial will review the physics basis, and its applications, for high magnetic field, compact visions of steady-state pilot plants and fusion reactors. This includes: energy and particle confinement; transport barriers; heating and current drive; scrape-off layer and divertor physics including implications for power handling, and ash/impurity control. The development of new technologies, particularly high-temperature, high critical magnetic field superconducting materials opens a new opportunity to consider the leverage of on-axis magnetic fields of 10T or more, enabling the feasibility of smaller sized devices on the path to fusion energy, including a pilot plant which could produce hundreds of megawatts of net electricity in a 10T tokamak with major radius of order 3 meter. Incorporating jointed magnetic coils, also made feasible by the high temperature superconductors, can dramatically improve flexibility of experimental superconducting facilities, and ultimately maintainability for reactor systems. Steady-state requires high bootstrap fraction, combined with efficient off-axis current drive, and existing and new approaches for RF sustainment will be covered, including Lower Hybrid Current Drive (both from the low- and high-field side), ECCD, and fast-wave techniques. External torque drive from neutral beams, routinely used in most present-day experiments to enhance confinement and suppress instabilities, will be weak or absent in reactors. Alternative, RF-based flow drive, using mode-converted ICRF waves will be discussed. All reactor concepts have extraordinary power handling requirements, combined with stringent limits on PFC erosion and impurity sources; the current state of the art in divertor configurations will be compared with emerging and new concepts, including snowflake, x-point, x-divertor and liquid metals, to meet these challenges. Supported by USDOE.

  2. Development and investigations of compact heat-transfer equipment for a nuclear power station equipped with a high-temperature gas-cooled reactor

    NASA Astrophysics Data System (ADS)

    Golovko, V. F.; Dmitrieva, I. V.; Kodochigov, N. G.; Bykh, O. A.

    2013-07-01

    The project of a nuclear power station the reactor coolant system of which includes a high-temperature gas-cooled reactor combined with a gas-turbine energy conversion unit supposes the use of high-efficient gas-cycle-based heat-transfer equipment. An analysis aimed at selecting the optimal heat-transfer surfaces is presented together with the results from their calculated and experimental investigation. The design features of recuperators arranged integrally with end and intermediate coolers and placed in a vertical sealed high-pressure vessel of limited sizes are considered.

  3. Prospects and status of low-aspect-ratio tokamaks

    SciTech Connect

    Peng, Y.K.M.

    1994-12-31

    The prospects for the low-aspect-ratio (A) tokamak to fulfill the requirements of viable fusion power plants are considered relative to the present status in data and modeling. Desirable physics and design features for an attractive Blanket Test Facility and power reactors are estimated for low-A tokamaks based on calculations improved with the latest data from small pioneering experiments. While these experiments have confirmed some of the recent predictions for low-A, they also identify the remaining issues that require verification before reliable projections can be made for these deuterium-tritium applications. The results show that the low-A regime of small size, modest field, and high current offers a path complementary to the standard and high A tokamaks in developing the full potential of fusion power.

  4. Orbit effects on impurity transport in a rotating tokamak plasma

    SciTech Connect

    Wong, K.L.; Cheng, C.Z.

    1988-05-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster with a higher bounce frequency, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle orbits near the surface of a rotating tokamak are also analyzed. Orbit effects indicate that more impurities can penetrate into a plasma rotating with counter-beam injection. Particle simulation is carried out with realistic experimental parameters and the results are in qualitative agreement with some experimental observations in the Tokamak Fusion Test Reactor (TFTR). 19 refs., 15 figs.

  5. The Thomson scattering systems of the ASDEX upgrade tokamak

    SciTech Connect

    Murmann, H.; Goetsch, S.; Roehr, H.; Salzmann, H.; Steuer, K.H. )

    1992-10-01

    The Thomson scattering system of the ASDEX upgrade (AUG) tokamak is described. One of the main objectives of AUG is to investigate plasma wall interaction in reactor relevant discharges with a magnetic divertor. The very successful Nd:YAG scattering system developed for its predecessor ASDEX, has been upgraded to give higher spatial and temporal resolution, reliability, and flexibility to different discharge conditions. The system consists of two independently operating devices, each using a cluster of six lasers: One measures the electron temperature and density along three possible vertical chords alternatively through the magnetic axis, or the inner or outer boundary layer; a second chord in the equatorial plane will always cover the magnetic center even in the case of considerable Shafranov shifts. An additional compact spectrometer has been designed for measurements with high radial resolution in the equatorial plane across the separatrix. A third system, using the laser beams for the vertical arrangement once again, has been designed for profile measurements in the energy deposition zone 2 cm above the outer divertor plate. Each laser is run at a repetition rate of 20 Hz and 1 J per pulse. A variety of synchronization modes are available, e.g., 20 Hz/6 J, 120 Hz/1 J etc., or repetitive bursts at 20 Hz. In this case the minimum delay between two pulses is presently limited to {approx}30 {mu}s by the existing data acquisition. This mode will be used for investigating fast phenomena such as sawteeth or disruptive instabilities. During the time intervals between the laser pulses the bremsstrahlung radiation (line integral) will be measured by the Thomson scattering detection system to calculate {ital Z}{sub eff}.

  6. Power supplies and quench protection for the Tokamak Physics Experiment

    SciTech Connect

    Neumeyer, C.L.

    1994-07-01

    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). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems.

  7. Excitation of large-{kappa}{sub {theta}} ion-Bernstein waves in tokamaks

    SciTech Connect

    Valeo, E.J.; Fisch, N.J.

    1994-09-01

    The mode-converted ion-Bernstein wave excited in tokamaks is shown to exhibit certain very interesting behavior, including the attainment of very small poloidal phase velocities, the reversal of poloidal direction, and up-down asymmetries in propagation and damping. Because of these effects, this wave holds promise for channeling {alpha}-particle power to ions, something that would make a tokamak fusion reactor far more attractive than presently envisioned.

  8. Compact accelerator

    DOEpatents

    Caporaso, George J.; Sampayan, Stephen E.; Kirbie, Hugh C.

    2007-02-06

    A compact linear accelerator having at least one strip-shaped Blumlein module which guides a propagating wavefront between first and second ends and controls the output pulse at the second end. Each Blumlein module has first, second, and third planar conductor strips, with a first dielectric strip between the first and second conductor strips, and a second dielectric strip between the second and third conductor strips. Additionally, the compact linear accelerator includes a high voltage power supply connected to charge the second conductor strip to a high potential, and a switch for switching the high potential in the second conductor strip to at least one of the first and third conductor strips so as to initiate a propagating reverse polarity wavefront(s) in the corresponding dielectric strip(s).

  9. Energy confinement in tokamaks

    SciTech Connect

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston.

  10. Polarization spectroscopy of tokamak plasmas

    SciTech Connect

    Wroblewski, D.

    1991-09-01

    Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs.

  11. Compact magnetograph

    NASA Technical Reports Server (NTRS)

    Title, A. M.; Gillespie, B. A.; Mosher, J. W.

    1982-01-01

    A compact magnetograph system based on solid Fabry-Perot interferometers as the spectral isolation elements was studied. The theory of operation of several Fabry-Perot systems, the suitability of various magnetic lines, signal levels expected for different modes of operation, and the optimal detector systems were investigated. The requirements that the lack of a polarization modulator placed upon the electronic signal chain was emphasized. The PLZT modulator was chosen as a satisfactory component with both high reliability and elatively low voltage requirements. Thermal control, line centering and velocity offset problems were solved by a Fabry-Perot configuration.

  12. Experimental studies of compact toroids

    SciTech Connect

    Not Available

    1991-01-01

    The Berkeley Compact Toroid Experiment (BCTX) device is a plasma device with a Marshall-gun generated, low aspect ratio toroidal plasma. The device is capable of producing spheromak-type discharges and may, with some modification, produce low-aspect ratio tokamak configurations. A unique aspect of this experimenal devie is its large lower hybrid (LH) heating system, which consists of two 450MHz klystron tubes generating 20 megawatts each into a brambilla-type launching structure. Successful operation with one klystron at virtually full power (18 MW) has been accomplished with 110 {mu}s pulse length. A second klystron is currently installed in its socket and magnet but has not been added to the RF drive system. This report describes current activities and accomplishments and describes the anticipated results of next year's activity.

  13. First neutron spectrometry measurement at the HL-2A Tokamak

    NASA Astrophysics Data System (ADS)

    Yuan, Xi; Zhang, Xing; Xie, Xu-Fei; Chen, Zhong-Jing; Peng, Xing-Yu; Fan, Tie-Shuan; Chen, Jin-Xiang; Li, Xiang-Qing; Yuan, Guo-Liang; Yang, Qing-Wei; Yang, Jin-Wei

    2013-12-01

    A compact neutron spectrometer based on the liquid scintillator is presented for neutron energy spectrum measurements at the HL-2A Tokamak. The spectrometer was well characterized and a fast digital pulse shape discrimination software was developed using the charge comparison method. A digitizer data acquisition system with a maximum frequency of 1 MHz can work under an environment with a high count rate at HL-2A Tokamak. Specific radiation and magnetic shielding for the spectrometer were designed for the neutron spectrum measurement at the HL-2A Tokamak. For pulse height spectrum analysis, dedicated numerical simulation utilizing NUBEAM combined with GENESIS was performed to obtain the neutron energy spectrum. Subsequently, the transportation process from the plasma to the detector was evaluated with Monte Carlo calculations. The distorted neutron energy spectrum was folded with the response matrix of the liquid scintillation spectrometer, and good consistency was found between the simulated and measured pulse height spectra. This neutron spectrometer based on a digital acquisition system could be well adopted for the investigation of the auxiliary heating behavior and the fast-ion related phenomenon on different tokamak devices.

  14. Transmutation of nuclear waste with a low-aspect-ratio tokamak neutron source

    NASA Astrophysics Data System (ADS)

    Hong, Bong Guen; Moon, Se Youn

    2014-10-01

    The transmutation characteristics of transuranics (TRUs) in a transmutation reactor based on a LAR (Low-aspect-ratio) tokamak as a neutron source are investigated. The optimum radial build of a transmutation reactor is found by using a coupled analysis of the tokamak systems and the neutron transport. The dependences of the transmutation characteristics on the aspect ratio A in the range of 1.5 to 2.5 and on the fusion power in the range of 150 to 500 MW are investigated. An equilibrium fuel cycle is developed for effective transmutation, and show that with one unit of the transmutation reactor based on the LAR tokamak producing fusion power in the range of a few hundred MWs, up to 3 PWRs (1.0 GWe capacity) can be supported with a burn-up fraction larger than 50%.

  15. Co3O4-based honeycombs as compact redox reactors/heat exchangers for thermochemical storage in the next generation CSP plants

    NASA Astrophysics Data System (ADS)

    Pagkoura, Chrysoula; Karagiannakis, George; Halevas, Eleftherios; Konstandopoulos, Athanasios G.

    2016-05-01

    Over the last years, several research groups have focused on developing efficient thermochemical heat storage (THS) systems, in-principle capable of being coupled with next generation high temperature Concentrated Solar Power plants. Among systems studied, the Co3O4/CoO redox system is a promising candidate. Currently, research efforts extend beyond basic level identification of promising materials to more application-oriented approaches aiming at validation of THS performance at pilot scale reactors. The present work focuses on the investigation of cobalt oxide based honeycomb structures as candidate reactors/heat exchangers to be employed for such purposes. In the evaluation conducted and presented here, cobalt oxide-based structures with different composition and geometrical characteristics were subjected to redox cycles in the temperature window between 800 and 1000°C under air flow. Basic aspects related to redox performance of each system are briefly discussed but the main focus lies on the evaluation of the segments structural stability after multi-cyclic operation. The latter is based on macroscopic visual observation and also supplemented by pre- (i.e. fresh samples) and post-characterization (i.e. after long term exposure) of extruded honeycombs via combined mercury porosimetry and SEM analysis.

  16. Thermally excited proton spin-flip laser emission in tokamaks

    SciTech Connect

    Arunasalam, V.; Greene, G.J.

    1993-07-01

    Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser.

  17. FEM (Free Electron Maser) for tokamak: Final report

    SciTech Connect

    Not Available

    1987-01-01

    This paper studies the feasibility of a microwave source for heating a tokamak reactor. The free electron maser (FEM) shows great promise for being this source. The topics covered in this paper are microwave generation with FEM, efficiency enhancement, parameter scaling, space charge scaling, beam energy spread and efficiency scaling, electron beam line with energy recovery, achromatic bend, multi-stage depressed voltage electron beam collector, and development plans. 12 refs., 10 figs., 5 tabs. (LSP)

  18. The tokamak as a neutron source

    SciTech Connect

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs.

  19. Mathematical modeling plasma transport in tokamaks

    SciTech Connect

    Quiang, Ji

    1995-12-31

    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{sup 20}/m{sup 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%.

  20. Dust Measurements in Tokamaks

    SciTech Connect

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  1. Sawtooth oscillation in tokamaks

    SciTech Connect

    Park, W.; Monticello, D.A.

    1989-03-01

    A three-dimensional nonlinear toroidal full MHD code, MH3D, has been used to study sawtooth oscillations in tokamaks. The profile evolution during the sawtooth crash phase compares well with experiment, but only if neoclassical resistivity is used in the rise phase. (Classical resistivity has been used in most of the previous theoretical sawtooth studies.) With neoclassical resistivity, the q value at the axis drops from 1 to about 0.8 before the crash phase, and then resets to 1 through a Kadomtsev-type complete reconnection process. This ..delta..q/sub 0/ approx. = 0.2 is much larger than ..delta..q/sub o/ approx. = 0.01, which is obtained if classical resistivity is used. The current profile is strongly peaked at the axis with a flat region around the singular surface, and is similar to the Textor profile. To understand this behavior, approximate formulas for the time behavior of current and q values are derived. A functional dependence of sawtooth period scaling is also derived. A semi-empirical scaling is found which fits the experimental data from various tokamaks. Some evidence is presented which indicates that the fast crash time is due to enhanced effective resistivity inside the singular current sheet, generated by, e.g., microinstability and electron parallel viscosity with stochastic fields at the x-point. 16 refs., 5 figs.

  2. Tritium Experience in Large Tokamaks: Application to ITER

    SciTech Connect

    Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

    1998-05-01

    Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R{ampersand}D avenues to address them. This paper summarizes the presentations, discussions, and recommendations.

  3. Proof-of-Principle Experiment for Compact, Energy Efficient Neutron Source: Enabling Technology for Radioactive Waste Transmutation or Sub-Critical Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Hershcovitch, Ady; Roser, Thomas; Santarius, John

    2013-10-01

    A novel neutron source is proposed for radioactive waste transmutation or sub-critical nuclear reactors; it's based on injecting 125 keV deuterium beam through 1-inch tube filled with magnetized tritium plasma to generate 14 MeV D-T neutrons. T target thickness is chosen to slow the D ions to 75 keV. At the opposite end of the tube D ion energy is recovered. Each ion source and tube forms a module. Larger systems can be formed from multiple units. As a D beam propagates through T plasma, it is slowed down by plasma electrons, which are consequently heated. Electron temperature rises until heating is balanced by energy losses. Equilibrium electron temperature is the crucial parameter, since higher temperature, leads to lower drag on the ion beam; therefore, larger target thickness is needed to slow deuterons to 75 keV; with consequently higher neutron yield. A proof of principle experiment, to determine the equilibrium electron temperature, can be perform by injecting 62.5 keV hydrogen beam into hydrogen plasma target and measure the equilibrium electron temperature with Thomson scattering. To reduce electron equilibration with target ions, electron pre-heating can be done rather efficiently with 2.45 GHz microwaves. Supported by USDOE under Contract No. DE-AC02-98CH10886.

  4. ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

    NASA Astrophysics Data System (ADS)

    Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

    2014-10-01

    The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

  5. Observation of finite-. beta. MHD phenomena in tokamaks

    SciTech Connect

    McGuire, K.M.

    1984-09-01

    Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1 internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.

  6. Alternative fusion concepts and the prospects for improved reactors

    NASA Astrophysics Data System (ADS)

    Krakowski, R. A.

    1985-05-01

    Past trends, present status, and future directions in the search for an improved fusion reactor are reviewed, and promising options available to both the principle tokamak and other supporting concept are summarized.

  7. Argonne Plasma Engineering Experiment (APEX) Tokamak

    SciTech Connect

    Norem, J.H.; Balka, L.J.; Kulovitz, E.E.; Magill, S.R.; McGhee, D.G.; Moretti, A.; Praeg, W.F.

    1981-03-01

    The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials.

  8. ITER tokamak buildings and equipment layout

    SciTech Connect

    Ahlfeld, C.E.; Dilling, D.A.; Ishimoto, Kazuyuki; Tanaka, Eiichi; Stoner, S.

    1996-12-31

    The International Thermonuclear Experimental Reactor (ITER) design has evolved to a level of maturity that has enabled the building designers to define the major dimensions and characteristics of the cluster of buildings that contain the tokamak and adjacent support equipment. Three-dimensional building models developed in a CATIA database provide the framework for the equipment layout. This article describes the preliminary layout of all major pieces of equipment, large bore pipes, ducts, busbars and other services. It is anticipated that some features of the layout will change as equipment design is advanced and future decisions are made, but these changes are not expected to alter the basic building design and any necessary changes are facilitated by the 3-D CATIA models. 1 ref., 6 figs.

  9. Causes of major tokamak disruptions

    SciTech Connect

    White, R.B.; Monticello, D.A.

    1980-07-01

    The nonlinear saturation theory of the tearing mode is used to examine the necessary conditions for the occurrence of a major tokamak disruption. The results are compared with full three-dimensional numerical simulations, and with experimental data.

  10. Multidimensional optimization of fusion reactors using heterogenous codes and engineering software

    NASA Astrophysics Data System (ADS)

    Hartwig, Zachary; Olynyk, Geoffrey; Whyte, Dennis

    2012-10-01

    Magnetic confinement fusion reactors are tightly coupled systems. The parameters under a designer's control, such as magnetic field, wall temperature, and blanket thickness, simultaneously affect the behavior, performance, and components of the reactor, leading to complex tradeoffs and design optimizations. In addition, the engineering analyses require non-trivial, self-consistent inputs, such as reactor geometry, to ensure high fidelity between the various physics and engineering design codes. We present a framework for analysis and multidimensional optimization of fusion reactor systems based on the coupling of heterogeneous codes and engineering software. While this approach is widely used in industry, most code-coupling efforts in fusion have been focused on plasma and edge physics. Instead, we use a simplified plasma model to concentrate on how fusion neutrons and heat transfer affect the design of the first wall, breeding blanket, and magnet systems. The framework combines solid modeling, neutronics, and engineering multiphysics codes and software, linked across Windows and Linux clusters. Initial results for optimizing the design of a compact, high-field tokamak reactor based on high-temperature demountable superconducting coils and a liquid blanket are presented.

  11. Compaction behavior of roller compacted ibuprofen.

    PubMed

    Patel, Sarsvatkumar; Kaushal, Aditya Mohan; Bansal, Arvind Kumar

    2008-06-01

    The effect of roller compaction pressure on the bulk compaction of roller compacted ibuprofen was investigated using instrumented rotary tablet press. Three different roller pressures were utilized to prepare granules and Heckel analysis, Walker analysis, compressibility, and tabletability were performed to derive densification, deformation, course of volume reduction and bonding phenomenon of different pressure roller compacted granules. Nominal single granule fracture strength was obtained by micro tensile testing. Heckel analysis indicated that granules prepared using lower pressure during roller compaction showed lower yield strength. The reduction in tabletability was observed for higher pressure roller compacted granules. The reduction in tabletability supports the results of granule size enlargement theory. Apart from the granule size enlargement theory, the available fines and relative fragmentation during compaction is responsible for higher bonding strength and provide larger areas for true particle contact at constant porosity for lower pressure roller compacted granules. Overall bulk compaction parameters indicated that granules prepared by lower roller compaction pressure were advantageous in terms of tabletability and densification. Overall results suggested that densification during roller compaction affects the particle level properties of specific surface area, nominal fracture strength, and compaction behavior. PMID:18280716

  12. Ceramic powder compaction

    SciTech Connect

    Glass, S.J.; Ewsuk, K.G.; Mahoney, F.M.

    1995-12-31

    With the objective of developing a predictive model for ceramic powder compaction we have investigated methods for characterizing density gradients in ceramic powder compacts, reviewed and compared existing compaction models, conducted compaction experiments on a spray dried alumina powder, and conducted mechanical tests and compaction experiments on model granular materials. Die filling and particle packing, and the behavior of individual granules play an important role in determining compaction behavior and should be incorporated into realistic compaction models. These results support the use of discrete element modeling techniques and statistical mechanics principals to develop a comprehensive model for compaction, something that should be achievable with computers with parallel processing capabilities.

  13. Diagnostics for hybrid reactors

    NASA Astrophysics Data System (ADS)

    Orsitto, Francesco Paolo

    2012-06-01

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  14. Diagnostics for hybrid reactors

    SciTech Connect

    Orsitto, Francesco Paolo

    2012-06-19

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  15. LIDAR Thomson scattering for advanced tokamaks. Final report

    SciTech Connect

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-03-18

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured.

  16. Three-dimensional equilibria in axially symmetric tokamaks.

    PubMed

    Garabedian, Paul R

    2006-12-19

    The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of alpha particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158

  17. Three-dimensional equilibria in axially symmetric tokamaks

    PubMed Central

    Garabedian, Paul R.

    2006-01-01

    The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of α particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158

  18. [alpha]-particle transport-driven current in tokamaks

    SciTech Connect

    Heikkinen, J.A. ); Sipilae, S.K. )

    1995-03-01

    It is shown that the radial transport of fusion-born energetic [alpha] particles, induced by electrostatic waves traveling in one poloidal direction, is directly connected to a net momentum of [alpha] particles in the toroidal direction in tokamaks. Because the momentum change is almost independent of toroidal velocity, the energy required for the momentum generation remains small on an [alpha]-particle population sustained by an isotropic time-independent source. By numerical toroidal Monte Carlo calculations it is shown that the current carried by [alpha] particles in the presence of intense well penetrated waves can reach several mega-amperes in reactor-sized tokamaks. The current obtained can greatly exceed the neoclassical bootstrap current of the [alpha] particles.

  19. Tokamak Plasma Flows Induced by Local RF Forces

    NASA Astrophysics Data System (ADS)

    Chen, Jiale; Gao, Zhe

    2015-10-01

    The tokamak plasma flows induced by the local radio frequency (RF) forces in the core region are analyzed. The effective components of local RF forces are composed of the momentum absorption term and the resonant parallel momentum transport term (i.e. the parallel component of the resonant ponderomotive forces). Different momentum balance relations are employed to calculate the plasma flows depending on different assumptions of momentum transport. With the RF fields solved from RF simulation codes, the toroidal and poloidal flows by these forces under the lower hybrid current drive and the mode conversion ion cyclotron resonance heating on EAST-like plasmas are evaluated. supported by National Natural Science Foundation of China (Nos. 11405218, 11325524, 11375235 and 11261140327), in part by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB111002, 2013GB112001 and 2013GB112010), and the Program of Fusion Reactor Physics and Digital Tokamak with the CAS “One-Three-Five” Strategic Planning

  20. Limiter/vacuum system for plasma impurity control and exhaust in tokamaks

    SciTech Connect

    Abdou, M.; Brooks, J.; Mattas, R.

    1980-01-01

    A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification.

  1. Energy deposition in STARFIRE reactor components

    SciTech Connect

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  2. Simplified compact containment BWR plant

    SciTech Connect

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-07-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  3. Moving Divertor Plates in a Tokamak

    SciTech Connect

    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.

  4. Resistive instabilities in tokamaks

    SciTech Connect

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed.

  5. Status of tokamak experiments

    SciTech Connect

    Wolf, G.H.

    1996-03-01

    Plasma-wall interaction, heat removal and ash exhaust have emerged as the dominant problems still to be solved in order to achieve ignition and - even more difficult - to maintain a state of self-sustained thermo-nuclear burn. This is of particular delicacy, since those operational regimes which yield the best energy confinement correspond to an even better particle confinement and confinement of impurities, which then tend to accumulate in the plasma core and to result in disruption or degradation of the tokamak discharge. Therefore, plasma-wall interaction, heat removal and particle exhaust will determine not only the structure and configuration of the plasma edge region, of the wall system and of the materials facing the plasma, but also the final choice of useful confinement regimes. Moreover, the potential effect of powerful {alpha}-particle heating on plasma stability and confinement has to be kept below critical values. For the latter requirement, a final answer can only be obtained in an ITER-type device where ignition and burn will become accessible. 72 refs., 12 figs.

  6. Protection of tokamak plasma facing components by a capillary porous system with lithium

    NASA Astrophysics Data System (ADS)

    Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

    2015-08-01

    Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

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

    NASA Astrophysics Data System (ADS)

    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.

  8. Compact Process Development at Babcock & Wilcox

    SciTech Connect

    Eric Shaber; Jeffrey Phillips

    2012-03-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of compaction trials have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel at packing fractions exceeding 46% by volume. Results from these trials are included. The scale-up effort is nearing completion with the process installed and operable using nuclear fuel materials. Final process testing is in progress to certify the process for manufacture of qualification test fuel compacts in 2012.

  9. Continuous tokamak operation with an internal transformer

    SciTech Connect

    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.

  10. Tokamak coordinate conventions: COCOS

    NASA Astrophysics Data System (ADS)

    Sauter, O.; Medvedev, S. Yu.

    2013-02-01

    Dealing with electromagnetic fields, in particular current and related magnetic fields, yields "natural" physical vector relations in 3-D. However, when it comes to choosing local coordinate systems, the "usual" right-handed systems are not necessarily the best choices, which means that there are several options being chosen. In the magnetic fusion community such a difficulty exists for the choices of the cylindrical and of the toroidal coordinate systems. In addition many codes depend on knowledge of an equilibrium. In particular, the Grad-Shafranov axisymmetric equilibrium solution for tokamak plasmas, ψ, does not depend on the sign of the plasma current Ip nor that of the magnetic field B0. This often results in ill-defined conventions. Moreover the sign, amplitude and offset of ψ are of less importance, since the free sources in the equation depend on the normalized radial coordinate. The signs of the free sources, dp/dψ and dF2/dψ (p being the pressure, ψ the poloidal magnetic flux and F=RBφ), must be consistent to generate the current density profile. For example, RF and CD calculations (Radio Frequency heating and Current Drive) require an exact sign convention in order to calculate a co- or counter-CD component. It is shown that there are over 16 different coordinate conventions. This paper proposes a unique identifier, the COCOS convention, to distinguish between the 16 most-commonly used options. Given the present worldwide efforts towards code integration, the proposed new index COCOS defining uniquely the COordinate COnventionS required as input by a given code or module is particularly useful. As codes use different conventions, it is useful to allow different sign conventions for equilibrium code input and output, equilibrium being at the core of any calculations in magnetic fusion. Additionally, given two different COCOS conventions, it becomes simple to transform between them. The relevant transformations are described in detail.

  11. Compact reactor for onboard hydrogen generation

    NASA Technical Reports Server (NTRS)

    Brabbs, T. A.

    1980-01-01

    Hydrogen, chemically stored as methanol, is promising internal-combustion fuel. Methanol is readily obtainable from natural products such as wood, compost, or various organic wastes. Steam reformation of methanol as source for hydrogen is relatively simple operation.

  12. Transport theory for energetic alpha particles in finite aspect ratio tokamaks with broken symmetry

    NASA Astrophysics Data System (ADS)

    Shaing, K. C.; Schlutt, M.; Lai, A. L.

    2016-02-01

    Transport theory for the energetic alpha particles in finite aspect ratio tokamaks with broken symmetry is developed for the case where the slowing down collision operator dominates. The transport fluxes in the 1 /ν and superbanana plateau regimes are derived. Here, ν is the typical collision frequency. They can be used in modeling the energy loss of the alpha particles in thermonuclear fusion reactors. Numerical realizations of the superbanana orbits of alpha particles in tokamaks with broken symmetry are also presented. The existence of the superbananas corroborates the predictions of the theories presented here and elsewhere.

  13. Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

    NASA Astrophysics Data System (ADS)

    Freidberg, J.; Mangiarotti, F.; Minervini, J.

    2014-10-01

    The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

  14. Helium Refrigerator Design for Pulsed Heat Load in Tokamaks

    SciTech Connect

    Kuendig, A.; Schoenfeld, H.

    2006-04-27

    Nuclear fusion reactors of the Tokamak type will be operated in a pulsed mode requiring the helium refrigerator to remove periodically large heat loads in time steps of approximately one hour. What are the necessary steps for a refrigerator to cope with such load variations?A series of numerical simulations has been performed indicating the possibility of an active refrigerator control with low exergetic losses. A basic comparison is made between the largest existing refrigerator sizes and the size required to service for example the ITER requirements.

  15. Compaction Scale Up and Optimization of Cylindrical Fuel Compacts for the Next Generation Nuclear Plant

    SciTech Connect

    Jeffrey J. Einerson; Jeffrey A. Phillips; Eric L. Shaber; Scott E. Niedzialek; W. Clay Richardson; Scott G. Nagley

    2012-10-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of designed experiments have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel. Results from these experiments are included. The scale-up effort is nearing completion with the process installed and operational using nuclear fuel materials. The process is being certified for manufacture of qualification test fuel compacts for the AGR-5/6/7 experiment at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL).

  16. Profile control of advanced tokamak plasmas in view of continuous operation

    NASA Astrophysics Data System (ADS)

    Mazon, D.

    2015-07-01

    The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named 'advanced scenarios' are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated 'bootstrap' current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.

  17. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    SciTech Connect

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs.

  18. Colliding Beam Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  19. Optimized profiles for improved confinement and stability in the DIII-D tokamak

    SciTech Connect

    Taylor, T.S.; St. John, H.; Turnbull, A.D.

    1995-02-01

    Simultaneous achievement of high energy confinement, {tau}{sub E}, and high plasma beta, {beta}, leads to an economically attractive compact tokamak fusion reactor. High confinement enhancement, H = {tau}{sub E}/{tau}{sub E-ITER89P} = 4, and high normalized beta {beta}{sub N} = {beta}/(I/aB) = 6%-m-T/MA, have been obtained in DIII-D experimental discharges. These improved confinement and/or improved stability limits are observed in several DIII-D high performance operational regimes: VH-mode, high {ell}{sub i} H-mode, second stable core, and high beta poloidal. The authors have identified several important features of the improved performance in these discharges: details of the plasma shape, toroidal rotation or ExB flow profile, q profile and current density profile, and pressure profile. From the improved physics understanding of these enhanced performance regimes, they have developed operational scenarios which maintain the essential features of the improved confinement and which increase the stability limits using localized current profile control. The stability limit is increased by modifying the interior safety factor profile to be nonmonotonic with high central q, while maintaining the edge current density consistent with the improved transport regimes and the high edge bootstrap current. They have calculated high beta equilibria with {beta}{sub N} = 6.5, stable to ideal n = 1 kinks and stable to ideal ballooning modes. The safety factor at the 95% flux surface is 6, the central q value is 3.9 and the minimum in q is 2.6. The current density profile is maintained by the natural profile of the bootstrap current, and a modest amount of electron cyclotron current drive.

  20. Mouse Embryo Compaction.

    PubMed

    White, M D; Bissiere, S; Alvarez, Y D; Plachta, N

    2016-01-01

    Compaction is a critical first morphological event in the preimplantation development of the mammalian embryo. Characterized by the transformation of the embryo from a loose cluster of spherical cells into a tightly packed mass, compaction is a key step in the establishment of the first tissue-like structures of the embryo. Although early investigation of the mechanisms driving compaction implicated changes in cell-cell adhesion, recent work has identified essential roles for cortical tension and a compaction-specific class of filopodia. During the transition from 8 to 16 cells, as the embryo is compacting, it must also make fundamental decisions regarding cell position, polarity, and fate. Understanding how these and other processes are integrated with compaction requires further investigation. Emerging imaging-based techniques that enable quantitative analysis from the level of cell-cell interactions down to the level of individual regulatory molecules will provide a greater understanding of how compaction shapes the early mammalian embryo. PMID:27475854

  1. Microwave Tokamak Experiment: Overview and status

    SciTech Connect

    Not Available

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs.

  2. A charged fusion product diagnostic for a spherical tokamak

    NASA Astrophysics Data System (ADS)

    Perez, Ramona Leticia Valenzuela

    Designs for future nuclear fusion power reactors rely on the ability to create a stable plasma (hot ionized gas of hydrogen isotopes) as a medium with which to sustain nuclear fusion reactions. My dissertation work involves designing, constructing, testing, installing, operating, and validating a new diagnostic for spherical tokamaks, a type of reactor test facility. Through detecting charged particles emitted from the plasma, this instrument can be used to study fusion reaction rates within the plasma and how they are affected by plasma perturbations. Quantitatively assessing nuclear fusion reaction rates at specific locations inside the plasma and as a function of time can provide valuable data that can be used to evaluate theory-based simulations related to energy transport and plasma stability. The Proton Detector (PD), installed in the Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy (CCFE) in Abingdon, England, was the first instrument to experimentally detect 3 MeV Protons and 1 MeV Tritons created from deuterium- deuterium (hydrogen isotopes) nuclear fusion reactions inside a spherical tokamak's plasma. The PD consists of an array of particle detectors with a protective housing and the necessary signal conditioning electronics and readout. After several years of designing (which included simulations for detector orientations), fabricating, and testing the PD, it was installed in MAST and data were collected over a period of two months in the summer of 2013. Proton and triton rates as high as 200 kHz were measured and an initial radial profile of these fusion reaction rates inside the plasma was extracted. These results will be compared to a complementary instrument at MAST as well as theory-based simulations and form the knowledge basis for developing a larger future instrument. The design and performance of all instrument components (electrical, computational, mechanical), and subsequent data analysis methods and results are

  3. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    NASA Astrophysics Data System (ADS)

    Hughes, Paul E.

    It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

  4. STARFIRE: a commercial tokamak fusion power plant study

    SciTech Connect

    Not Available

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on 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 and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  5. Density limit disruptions in tokamaks

    NASA Astrophysics Data System (ADS)

    Kleva, Robert G.; Drake, J. F.

    1991-02-01

    Magnetohydrodynamic simulations are presented which reproduce the rapid drop in the central temperature observed during density limit disruptions in tokamaks. The loss of central confinement is triggered by edge radiation which destabilizes a q=1 kink mode. A bubble of cold plasma is injected from the edge into the center by the q=1 kink. This bubble bears a striking resemblance to the cold plasma that is observed to move from the edge into the center during density limit disruptions on the JET tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 433], initiating the loss of central confinement. The bubble profile produced by the q=1 kink is unstable to a broad spectrum of modes which progressively reduce the magnetic shear between the q=2 surface and the center. The q=2 mode then grows across the center, broadening the current and throwing the hot plasma to the wall.

  6. Comprehensive numerical modelling of tokamaks

    SciTech Connect

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-03

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell.

  7. Reversible DNA compaction.

    PubMed

    González-Pérez, Alfredo

    2014-01-01

    In this review we summarize and discuss the different methods we can use to achieve reversible DNA compaction in vitro. Reversible DNA compaction is a natural process that occurs in living cells and viruses. As a result these process long sequences of DNA can be concentrated in a small volume (compacted) to be decompacted only when the information carried by the DNA is needed. In the current work we review the main artificial compacting agents looking at their suitability for decompaction. The different approaches used for decompaction are strongly influenced by the nature of the compacting agent that determines the mechanism of compaction. We focus our discussion on two main artificial compacting agents: multivalent cations and cationic surfactants that are the best known compacting agents. The reversibility of the process can be achieved by adding chemicals like divalent cations, alcohols, anionic surfactants, cyclodextrins or by changing the chemical nature of the compacting agents via pH modifications, light induced conformation changes or by redox-reactions. We stress the relevance of electrostatic interactions and self-assembly as a main approach in order to tune up the DNA conformation in order to create an on-off switch allowing a transition between coil and compact states. The recent advances to control DNA conformation in vitro, by means of molecular self-assembly, result in a better understanding of the fundamental aspects involved in the DNA behavior in vivo and serve of invaluable inspiration for the development of potential biomedical applications. PMID:24444152

  8. Current Trends of Blanket Research and Deveopment in Japan 3.Blanket Designs in Fusion Power Reactors

    NASA Astrophysics Data System (ADS)

    Sagara, Akio; Enoeda, Mikio; Nishio, Satoshi; Kozaki, Yasuji

    The main functions of the blanket in fusion power reactors are basically independent of the type of magnetic fusion reactor (tokamak, helical, etc.) and inertia fusion. However, from technical point of view, many candidate designs of blanket have been proposed depending on the particular reactor concepts. Their main features are characterized for the recent typical designs, and key issues are defined.

  9. Magnetic island formation in tokamaks

    SciTech Connect

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs.

  10. Gyrosheath near the tokamak edge

    SciTech Connect

    Hazeltine, R.D.; Xiao, H.; Valanju, P.M.

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results.

  11. Neoclassical magnetic microislands in tokamaks

    SciTech Connect

    Kovalishen, E.A.; Mikhailovskii, A.B.; Botov, P.V.; Shirokov, M.S.; Konovalov, S.V.; Tsypin, V.S.; Galvao, R.M.O.

    2005-09-15

    Possibility of existence of neoclassical magnetic microislands (island width smaller than the ion Larmor radius) in a tokamak in the banana regime is shown. The rotation frequency of such islands is found. It is shown that for the case of positive electron temperature gradient, the bootstrap current destabilizes the microislands while the polarization current leads to their stabilization. Maximally possible neoclassical microisland width is estimated.

  12. Fast wave current drive experiment on the DIII-D tokamak

    SciTech Connect

    Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Lohr, J.; Luce, T.C.; Mayberry, M.J.; Prater, R. ); Porkolab, M. ); Baity, F.W.; Goulding, R.H.; Hoffman, J.D. ); James, R.A. ); Kawash

    1992-06-01

    One method of radio-frequency heating which shows theoretical promise for both heating and current drive in tokamak plasmas is the direct absorption by electrons of the fast Alfven wave (FW). Electrons can directly absorb fast waves via electron Landau damping and transit-time magnetic pumping when the resonance condition {omega} {minus} {kappa}{sub {parallel}e}{upsilon}{sup {parallel}e} = O is satisfied. Since the FW accelerates electrons traveling the same toroidal direction as the wave, plasma current can be generated non-inductively by launching FW which propagate in one toroidal direction. Fast wave current drive (FWCD) is considered an attractive means of sustaining the plasma current in reactor-grade tokamaks due to teh potentially high current drive efficiency achievable and excellent penetration of the wave power to the high temperature plasma core. Ongoing experiments on the DIII-D tokamak are aimed at a demonstration of FWCD in the ion cyclotron range of frequencies (ICRF). Using frequencies in the ICRF avoids the possibility of mode conversion between the fast and slow wave branches which characterized early tokamak FWCD experiments in the lower hybrid range of frequencies. Previously on DIII-D, efficient direct electron heating by FW was found using symmetric (non-current drive) antenna phasing. However, high FWCD efficiencies are not expected due to the relatively low electron temperatures (compared to a reactor) in DIII-D.

  13. Compaction behavior of isomalt after roll compaction.

    PubMed

    Quodbach, Julian; Mosig, Johanna; Kleinebudde, Peter

    2012-01-01

    The suitability of the new isomalt grade galenIQ™ 801 for dry granulation and following tableting is evaluated in this study. Isomalt alone, as well as a blend of equal parts with dibasic calcium phosphate, is roll compacted and tableted. Particle size distribution and flowability of the granules and friability and disintegration time of the tablets are determined. Tensile strength of tablets is related to the specific compaction force during roll compaction and the tableting force. In all cases, the tensile strength increases with raising tableting forces. The specific compaction force has a different influence. For isomalt alone the tensile strength is highest for tablets made from granules prepared at 2 kN/cm and 6 kN/cm and decreases at higher values, i.e., >10 kN/cm. Tensile strength of the blend tablets is almost one third lower compared to the strongest tablets of pure isomalt. Friability of pure isomalt tablets is above the limit. Disintegration time is longest when the tensile strength is at its maximum and decreases with higher porosity and lower tensile strengths. Isomalt proves to be suitable for tableting after roll compaction. Even though the capacity as a binder might not be as high as of other excipients, it is a further alternative for the formulation scientist. PMID:24300366

  14. Compaction Behavior of Isomalt after Roll Compaction

    PubMed Central

    Quodbach, Julian; Mosig, Johanna; Kleinebudde, Peter

    2012-01-01

    The suitability of the new isomalt grade galenIQ™ 801 for dry granulation and following tableting is evaluated in this study. Isomalt alone, as well as a blend of equal parts with dibasic calcium phosphate, is roll compacted and tableted. Particle size distribution and flowability of the granules and friability and disintegration time of the tablets are determined. Tensile strength of tablets is related to the specific compaction force during roll compaction and the tableting force. In all cases, the tensile strength increases with raising tableting forces. The specific compaction force has a different influence. For isomalt alone the tensile strength is highest for tablets made from granules prepared at 2 kN/cm and 6 kN/cm and decreases at higher values, i.e., >10 kN/cm. Tensile strength of the blend tablets is almost one third lower compared to the strongest tablets of pure isomalt. Friability of pure isomalt tablets is above the limit. Disintegration time is longest when the tensile strength is at its maximum and decreases with higher porosity and lower tensile strengths. Isomalt proves to be suitable for tableting after roll compaction. Even though the capacity as a binder might not be as high as of other excipients, it is a further alternative for the formulation scientist. PMID:24300366

  15. The TITAN reversed-field-pinch fusion reactor study

    SciTech Connect

    Not Available

    1990-01-01

    This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors.

  16. Design study status of compact containment BWR

    SciTech Connect

    Heki, H.; Nakamaru, M.; Kuroki, M.; Kojima, Y.; Arai, K.; Tahara, M.; Hoshi, T.

    2006-07-01

    The reactor concept considered in this paper has a relatively mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Japan Atomic Power Company has been taking initiative in developing the concept of the Compact Containment Boiling Water Reactor (CCR). The CCR., which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's relatively mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, upper entry control rod drives (CRDs) and simplified safety system with high pressure resistible containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The upper entry CRDs enable a simplified safety system followed by in-vessel retention (IVR) capability with the compact primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration of RPV below the top of the core height, passive cooling system by isolation condenser (IC). The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. Further core design study has been carried out taking into account compact reactor size and reduction of fuel

  17. First neutron spectrometry measurements in the ASDEX Upgrade tokamak

    NASA Astrophysics Data System (ADS)

    Tardini, G.; Zimbal, A.; Esposito, B.; Gagnon-Moisan, F.; Marocco, D.; Neu, R.; Schuhmacher, H.; the ASDEX Upgrade Team

    2012-03-01

    A compact neutron spectrometer based on the liquid scintillator BC501A has been installed on the ASDEX Upgrade tokamak. The aim is to measure neutron energy distribution functions as footprints of fast ions distribution functions, generated mainly via Neutral Beam Injection (NBI) in present day tokamaks. A flexible and fast software has been developed to perform digital pulse shape separation and to evaluate pulse height spectra. First measurements of count rates and pulse height spectra show a good signal to noise ratio for integration times comparable to the NBI slowing down time and to the energy confinement time. Due to the perpendicular line of sight, D-d fusion with perpendicular NBI is detected more efficiently and the line broadening of the 2.45 MeV neutrons is higher. Ion Cyclotron Resonance Heating (ICRH) combined to NBI exhibits a synergy effect, with count rates higher than the sum of the counts due to NBI and ICRH separately. Although the collimator is designed to screen gammas as much as possible, some qualitative gamma analysis is also possible, providing information in case of runaway electrons during disruptions. The experimental campaign for the characterisation of the system (detector + acquisition system) is complete and the determination of the response function is in progress.

  18. Steady State Tokamak Equilibria without Current Drive

    SciTech Connect

    Shaing, K.C.; Aydemir, A.Y.; Lin-Liu, Y.R.; Miller, R.L.

    1997-11-01

    Steady state tokamak equilibria without current drive are found. This is made possible by including the potato bootstrap current close to the magnetic axis. Tokamaks with this class of equilibria do not need seed current or current drive, and are intrinsically steady state. {copyright} {ital 1997} {ital The American Physical Society}

  19. Numerical tokamak turbulence project (OFES grand challenge)

    SciTech Connect

    Beer, M; Cohen, B I; Crotinger, J; Dawson, J; Decyk, V; Dimits, A M; Dorland, W D; Hammett, G W; Kerbel, G D; Leboeuf, J N; Lee, W W; Lin, Z; Nevins, W M; Reynders, J; Shumaker, D E; Smith, S; Sydora, R; Waltz, R E; Williams, T

    1999-08-27

    The primary research objective of the Numerical Tokamak Turbulence Project (NTTP) is to develop a predictive ability in modeling turbulent transport due to drift-type instabilities in the core of tokamak fusion experiments, through the use of three-dimensional kinetic and fluid simulations and the derivation of reduced models.

  20. GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

    NASA Astrophysics Data System (ADS)

    Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

    2015-09-01

    An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

  1. Can tokamaks PFC survive a single event of any plasma instabilities?

    NASA Astrophysics Data System (ADS)

    Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

    2013-07-01

    Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

  2. ACOUSTIC COMPACTION LAYER DETECTION

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The depth and strength of compacted layers in fields have been determined traditionally using the ASAE standardized cone penetrometer method. However, an on-the-go method would be much faster and much less labor intensive. The soil measurement system described here attempts to locate the compacted...

  3. Dynamical compactness and sensitivity

    NASA Astrophysics Data System (ADS)

    Huang, Wen; Khilko, Danylo; Kolyada, Sergiĭ; Zhang, Guohua

    2016-05-01

    To link the Auslander point dynamics property with topological transitivity, in this paper we introduce dynamically compact systems as a new concept of a chaotic dynamical system (X , T) given by a compact metric space X and a continuous surjective self-map T : X → X. Observe that each weakly mixing system is transitive compact, and we show that any transitive compact M-system is weakly mixing. Then we discuss the relationships between it and other several stronger forms of sensitivity. We prove that any transitive compact system is Li-Yorke sensitive and furthermore multi-sensitive if it is not proximal, and that any multi-sensitive system has positive topological sequence entropy. Moreover, we show that multi-sensitivity is equivalent to both thick sensitivity and thickly syndetic sensitivity for M-systems. We also give a quantitative analysis for multi-sensitivity of a dynamical system.

  4. Compaction properties of isomalt.

    PubMed

    Bolhuis, Gerad K; Engelhart, Jeffrey J P; Eissens, Anko C

    2009-08-01

    Although other polyols have been described extensively as filler-binders in direct compaction of tablets, the polyol isomalt is rather unknown as pharmaceutical excipient, in spite of its description in all the main pharmacopoeias. In this paper the compaction properties of different types of ispomalt were studied. The types used were the standard product sieved isomalt, milled isomalt and two types of agglomerated isomalt with a different ratio between 6-O-alpha-d-glucopyranosyl-d-sorbitol (GPS) and 1-O-alpha-d-glucopyranosyl-d-mannitol dihydrate (GPM). Powder flow properties, specific surface area and densities of the different types were investigated. Compactibility was investigated by compression of the tablets on a compaction simulator, simulating the compression on high-speed tabletting machines. Lubricant sensitivity was measured by compressing unlubricated tablets and tablets lubricated with 1% magnesium stearate on an instrumented hydraulic press. Sieved isomalt had excellent flow properties but the compactibility was found to be poor whereas the lubricant sensitivity was high. Milling resulted in both a strong increase in compactibility as an effect of the higher surface area for bonding and a decrease in lubricant sensitivity as an effect of the higher surface area to be coated with magnesium stearate. However, the flow properties of milled isomalt were too bad for use as filler-binder in direct compaction. Just as could be expected, agglomeration of milled isomalt by fluid bed agglomeration improved flowability. The good compaction properties and the low lubricant sensitivity were maintained. This effect is caused by an early fragmentation of the agglomerated material during the compaction process, producing clean, lubricant-free particles and a high surface for bonding. The different GPS/GPM ratios of the agglomerated isomalt types studied had no significant effect on the compaction properties. PMID:19327398

  5. Stabilization of compactible waste

    SciTech Connect

    Franz, E.M.; Heiser, J.H. III; Colombo, P.

    1990-09-01

    This report summarizes the results of series of experiments performed to determine the feasibility of stabilizing compacted or compactible waste with polymers. The need for this work arose from problems encountered at disposal sites attributed to the instability of this waste in disposal. These studies are part of an experimental program conducted at Brookhaven National Laboratory (BNL) investigating methods for the improved solidification/stabilization of DOE low-level wastes. The approach taken in this study was to perform a series of survey type experiments using various polymerization systems to find the most economical and practical method for further in-depth studies. Compactible dry bulk waste was stabilized with two different monomer systems: styrene-trimethylolpropane trimethacrylate (TMPTMA) and polyester-styrene, in laboratory-scale experiments. Stabilization was accomplished by wetting or soaking compactible waste (before or after compaction) with monomers, which were subsequently polymerized. Three stabilization methods are described. One involves the in-situ treatment of compacted waste with monomers in which a vacuum technique is used to introduce the binder into the waste. The second method involves the alternate placement and compaction of waste and binder into a disposal container. In the third method, the waste is treated before compaction by wetting the waste with the binder using a spraying technique. A series of samples stabilized at various binder-to-waste ratios were evaluated through water immersion and compression testing. Full-scale studies were conducted by stabilizing two 55-gallon drums of real compacted waste. The results of this preliminary study indicate that the integrity of compacted waste forms can be readily improved to ensure their long-term durability in disposal environments. 9 refs., 10 figs., 2 tabs.

  6. Transport equations in tokamak plasmas

    SciTech Connect

    Callen, J. D.; Hegna, C. C.; Cole, A. J.

    2010-05-15

    Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfven waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The 'mean field' effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The

  7. The Microwave Tokamak Experiment (MTX)

    SciTech Connect

    Thomassen, K.I.; Cohen, B.I.; Hooper, E.B.; Lang, D.D.; Nevins, W.M.

    1987-10-02

    A new experimental facility is being assembled at the Lawrence Livermore National Laboratory (LLNL) for studying microwave propagation and absorption in high density plasmas. A unique feature of the facility is the free electron laser (FEL) used to generate high peak power microwaves at 250 GHz, at a repetition rate so as to produce up to 2 MW of average power for up to 30 s. Called the Microwave Tokamak Experiment (MTX), the facility will be used for studies of plasma heating, current drive, and confinement.

  8. Breakdown in the pretext tokamak

    SciTech Connect

    Benesch, J.F.

    1981-06-01

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges.

  9. Development of a tokamak plasma optimized for stability and confinement

    SciTech Connect

    Politzer, P.A.

    1995-02-01

    Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density ({beta}) and high energy confinement ({tau}{sub E}); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H{equivalent_to} {tau}{sub E}/{tau}{sub ITER-89P} = 4) and high normalized {beta} ({beta}{sub N}{equivalent_to} {beta}/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q{sub min} = 2.6, and q{sub 95} = 6]. This model plasma uses profiles which the authors expect to be realizable. At {beta}{sub N} {>=} 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H {>=} 3 with VH-mode-like confinement.

  10. Disruption avoidance in the Frascati Tokamak Upgrade by means of magnetohydrodynamic mode stabilization using electron-cyclotron-resonance heating.

    PubMed

    Esposito, B; Granucci, G; Smeulders, P; Nowak, S; Martín-Solís, J R; Gabellieri, L

    2008-02-01

    Disruption avoidance by stabilization of MHD modes through injection of ECRH at different radial locations is reported. Disruptions have been induced in the FTU (Frascati Tokamak Upgrade) deuterium plasmas by Mo injection or by exceeding the density limit (D gas puffing). ECRH is triggered when the V(loop) exceeds a preset threshold value. Coupling between MHD modes (m/n=3/2, 2/1, 3/1) occurs before disruption. Direct heating of one coupled mode is sufficient to avoid disruptions, while heating close to the mode leads to disruption delay. These results could be relevant for the International Thermonuclear Experimental Reactor tokamak operation. PMID:18352292

  11. Global Hybrid Simulations of Energetic Particle Effects on the n=1 Mode in Tokamaks: Internal Kink and Fishbone Instability

    SciTech Connect

    G.Y. Fu; W. Park; H.R. Strauss; J. Breslau; J. Chen; S. Jardin; L.E. Sugiyama

    2005-08-09

    Global hybrid simulations of energetic particle effects on the n=1 internal kink mode have been carried out for tokamaks. For the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis Editors et al., Nucl. Fusion 39:2137 (1999)], it is shown that alpha particle effects are stabilizing for the internal kink mode. However, the elongation of ITER reduces the stabilization effects significantly. Nonlinear simulations of the precessional drift fishbone instability for circular tokamak plasmas show that the mode saturates due to flattening of the particle distribution function near the resonance region. The mode frequency chirps down rapidly as the flattening region expands radially outward. Fluid nonlinearity reduces the saturation level.

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

    PubMed

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

    2008-10-01

    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. PMID:19044649

  13. Predictive Modeling of Tokamak Configurations*

    NASA Astrophysics Data System (ADS)

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  14. Theoretical design of a compact energy recovering divertor

    NASA Astrophysics Data System (ADS)

    Baver, D. A.

    2015-11-01

    An energy recovering divertor (ERD) is a type of plasma direct converter (PDC) designed to fit in the divertor channel of a tokamak. Such a device reduces the heat load to the divertor plate by converting a portion of it into electrical energy. This recovered energy can then be used for auxiliary heating and current drive, fundamentally altering the relationship between scientific and engineering breakeven and reducing dependence on bootstrap current. Previous work on the ERD concept focused on amplification of Alfven waves in a manner similar to a free-electron laser. While conceptually straightforward, this concept was also bulky, thus limiting its applicability to existing tokamak experiments. A design is presented for an ERD based on sheath-localized waves. This makes possible a device sufficiently compact to fit in the divertor channel of many existing tokamak experiments, and moreover requires no new shaping coils to achieve the desired magnetic geometry or topology. In addition, incidental advantages of this concept will be discussed.

  15. Feasibility study of a fission-suppressed tokamak fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  16. Kinetic theory of plasma adiabatic major radius compression in tokamaks

    NASA Astrophysics Data System (ADS)

    Gorelenkova, M. V.; Gorelenkov, N. N.; Azizov, E. A.; Romannikov, A. N.; Herrmann, H. W.

    1998-05-01

    In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm's law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation.

  17. Modeling of ICRF Internal Transport Barrier Control for Advanced Tokamaks

    NASA Astrophysics Data System (ADS)

    Sund, R. S.; Scharer, J. E.

    1998-11-01

    We present an analysis of TFTR ICRF current drive experiments carried out by Majeski et al.(R. Majeski, J. Rodgers, G. Schilling, C. Phillips, J. Hosea and the TFTR Group, private communication.) The influence of deuterium, tritium, minority specie, electron and alpha concentrations, temperatures and beam fractions are considered for the two-ion mode conversion current drive experiments. Direct comparison with experimental data is carried out by means of a nonlocal large gyroradius ICRF code(O. Sauter, Ph.D. thesis, Ecole Polytechnique de Lausanne, Switzerland (1992).) which incorporates 1-D plasma profiles. It is found that substantial beam and alpha particle absorption can occur for some cases. Application of ion cyclotron range of frequencies internal transport barrier control requires further examination of fast wave mode conversion and the interaction of ion Bernstein waves with plasmas in advanced tokamaks. The effects of perpendicular and parallel magnetic gradients on the ion, electron, and alpha particle absorption are examined. A viable internal transport barrier control scheme for a reactor grade advanced tokamak will be discussed.

  18. Negative hydrogen ion source for TOKAMAK neutral beam injector (invited)

    NASA Astrophysics Data System (ADS)

    Okumura, Y.; Fujiwara, Y.; Kashiwagi, M.; Kitagawa, T.; Miyamoto, K.; Morishita, T.; Hanada, M.; Takayanagi, T.; Taniguchi, M.; Watanabe, K.

    2000-02-01

    Intense negative ion source producing multimegawatt hydrogen/deuterium negative ion beams has been developed for the neutral beam injector (NBI) in TOKAMAK thermonuclear fusion machines. Negative ions are produced in a cesium seeded multi-cusp plasma generator via volume and surface processes, and accelerated with a multistage electrostatic accelerator. The negative ion source for JT-60U has produced 18.5 A/360 keV (6.7 MW) H- and 14.3 A/380 keV (5.4 MW) D- ion beams at average current densities of 11 mA/cm2 (H-) and 8.5 mA/cm2 (D-). A high energy negative ion source has been developed for the next generation TOKAMAK such as the International Thermonuclear Experimental Reactor (ITER). The source has demonstrated to accelerate negative ions up to 1 MeV, the energy required for ITER. Higher negative ion current density of more than 20 mA/cm2 was obtained in the ITER concept sources. It was confirmed that the consumption rate of cesium is small enough to operate the source for a half year in ITER-NBI without maintenance.

  19. Dynamic simulations of the cryogenic system of a tokamak

    NASA Astrophysics Data System (ADS)

    Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

    2015-12-01

    Power generation in the next decades could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to 4.4K. The plasma works cyclically and the coil system is subjected to pulsed heat load which has to be handled by the refrigerator. By smoothing the variable loads, the refrigerator capacity can be set close to the average power; optimizing investment and operational costs. Within the “Broader Approach agreement” related to ITER project, CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) is in charge of providing the cryogenic system for the Japanese tokamak (JT-60SA), that is currently under construction in Naka. The system has been designed to handle the pulsed heat loads. To prepare the acceptance tests of the cryogenic system foreseen in 2016, both dynamic modelling and experimental tests on a scaled down mock-up are of high interest for assessing pulsed load smoothing control. After explaining HELIOS (HElium Loop for hIgh lOad Smoothing) operating modes, a dynamic model is presented, with results on the pulsed heat load scenarios. All the simulations have been performed with EcosimPro® and the associated cryogenic library CRYOLIB.

  20. Intrinsic momentum transport in tokamaks with tilted elliptical flux surfaces

    NASA Astrophysics Data System (ADS)

    Ball, Justin; Parra, Felix; Barnes, Michael; Dorland, William; Hammett, Gregory; Rodrigues, Paulo; Loureiro, Nuno

    2014-10-01

    Recent work demonstrated that breaking the up-down symmetry of tokamaks removes a constraint limiting intrinsic momentum transport, and hence toroidal rotation, to be small. We show, through MHD analysis, that ellipticity is most effective at introducing up-down asymmetry throughout the plasma. Using GS2, a local δf gyrokinetic code that self-consistently calculates momentum transport, we simulate tokamaks with tilted elliptical poloidal cross-sections and a Shafranov shift. These simulations show both the magnitude and poloidal dependence of nonlinear momentum transport. The results are consistent with TCV experimental measurements and suggest that this mechanism can generate rotation with an Alfven Mach number of several percent in a tilted elliptical ITER-like machine. It appears that rotation generated with up-down asymmetry may be sufficient to stabilize the resistive wall mode in reactor-sized devices. J.R.B. and F.I.P. were partially supported by the RCUK Energy Programme (grant number EP/I501045) and the European Unions Horizon 2020 research and innovation programme.

  1. Finite pressure effects on the tokamak sawtooth crash

    SciTech Connect

    Nishimura, Yasutaro

    1998-07-01

    The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed.

  2. Nuclear design of a very-low-activation fusion reactor

    SciTech Connect

    Cheng, E.T.; Hopkins, G.R.

    1983-06-01

    An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design.

  3. Waste management for JAERI fusion reactors

    NASA Astrophysics Data System (ADS)

    Tobita, K.; Nishio, S.; Konishi, S.; Jitsukawa, S.

    2004-08-01

    In the fusion reactor design study at Japan Atomic Energy Institute (JAERI), several waste management strategies were assessed. The assessed strategies are: (1) reinforced neutron shield to clear the massive ex-shielding components from regulatory control; (2) low aspect ratio tokamak to reduce the total waste; (3) reuse of liquid metal breeding material and neutron shield. Combining these strategies, the weight of disposal waste from a low aspect ratio reactor VECTOR is expected to be comparable with the metal radwaste from a light water reactor (˜4000 t).

  4. Compact microchannel system

    DOEpatents

    Griffiths, Stewart

    2003-09-30

    The present invention provides compact geometries for the layout of microchannel columns through the use of turns and straight channel segments. These compact geometries permit the use of long separation or reaction columns on a small microchannel substrate or, equivalently, permit columns of a fixed length to occupy a smaller substrate area. The new geometries are based in part on mathematical analyses that provide the minimum turn radius for which column performance in not degraded. In particular, we find that straight channel segments of sufficient length reduce the required minimum turn radius, enabling compact channel layout when turns and straight segments are combined. The compact geometries are obtained by using turns and straight segments in overlapped or nested arrangements to form pleated or coiled columns.

  5. Dark compact planets

    NASA Astrophysics Data System (ADS)

    Tolos, Laura; Schaffner-Bielich, Jürgen

    2015-12-01

    We investigate compact objects formed by dark matter admixed with ordinary matter made of neutron-star matter and white-dwarf material. We consider non-self annihilating dark matter with an equation of state given by an interacting Fermi gas. We find new stable solutions, dark compact planets, with Earth-like masses and radii from a few Km to few hundred Km for weakly interacting dark matter which are stabilized by the mutual presence of dark matter and compact star matter. For the strongly interacting dark matter case, we obtain dark compact planets with Jupiter-like masses and radii of few hundred Km. These objects could be detected by observing exoplanets with unusually small radii. Moreover, we find that the recently observed 2 M⊙ pulsars set limits on the amount of dark matter inside neutron stars which is, at most, 1 0-6 M⊙ .

  6. Compact baby Skyrmions

    SciTech Connect

    Adam, C.; Klimas, P.; Sanchez-Guillen, J.; Wereszczynski, A.

    2009-11-15

    For the baby Skyrme model with a specific potential, compacton solutions, i.e., configurations with a compact support and parabolic approach to the vacuum, are derived. Specifically, in the nontopological sector, we find spinning Q-balls and Q-shells, as well as peakons. Moreover, we obtain compact baby skyrmions with nontrivial topological charge. All these solutions may form stable multisoliton configurations provided they are sufficiently separated.

  7. Integrated modeling of temperature profiles in L-mode tokamak discharges

    SciTech Connect

    Rafiq, T.; Kritz, A. H.; Tangri, V.; Pankin, A. Y.; Voitsekhovitch, I.; Budny, R. V.

    2014-12-15

    Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

  8. Design and installation of a ferromagnetic wall in tokamak geometry.

    PubMed

    Hughes, P E; Levesque, J P; Rivera, N; Mauel, M E; Navratil, G A

    2015-10-01

    Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak-Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability. PMID:26520952

  9. Composition And Electrical Properties Of Dust From Tokamak Compass

    SciTech Connect

    Vaverka, J.; Beranek, M.; Pavlu, J.; Richterova, I.; Vysinka, M.; Safrankova, J.; Nemecek, Z.

    2011-11-29

    In spite of the fact that fusion is a subject of the study for many years, there are still a lot of open questions. One of the interesting topics in fusion research is a presence of dust grains in reactors. In the paper, dust grains born in tokamak Compass are studied and compared with samples of a spherical geometry and well known composition. A unique experimental setup was used for investigations of charging properties of such grains and the SEM and EDX spectroscopy was applied for a study of grain composition. We focus on the secondary emission because this process plays a prominent role when a portion of energetic electrons is present in surroundings of a particular grain. It was shown that depending on the grain size and material, energetic electrons charge the grains to positive potentials comparable with the energy of impinging electrons.

  10. Results from deuterium-tritium tokamak confinement experiments

    SciTech Connect

    Hawryluk, R.J.

    1997-02-01

    Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues.

  11. Power and particle exhaust in tokamaks

    SciTech Connect

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER`s nominal design positions; important directions for further research are identified.

  12. Dust measurements in tokamaks (invited)

    SciTech Connect

    Rudakov, D. L.; Yu, J. H.; Boedo, J. A.; Hollmann, E. M.; Krasheninnikov, S. I.; Moyer, R. A.; Muller, S. H.; Pigarov, A. Yu.; Rosenberg, M.; Smirnov, R. D.; West, W. P.; Boivin, R. L.; Bray, B. D.; Brooks, N. H.; Hyatt, A. W.; Wong, C. P. C.; Roquemore, A. L.; Skinner, C. H.; Solomon, W. M.; Ratynskaia, S.

    2008-10-15

    Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 {mu}m in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C{sub 2} dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

  13. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    SciTech Connect

    Gentile, C.A.; LaMarche, P.H.; Anderson, J.L.

    1995-07-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During `` time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems.

  14. First-wall/blanket materials selection for STARFIRE tokamak reactor

    SciTech Connect

    Smith, D.L.; Mattas, R.F.; Clemmer, R.G.; Davis, J.W.

    1980-01-01

    The development of the reference STARFIRE first-wall/blanket design involved numerous trade-offs in the materials selection process for the breeding material, coolant structure, neutron multiplier, and reflector. The major parameters and properties that impact materials selection and design criteria are reviewed.

  15. Neutron spectroscopy on TFTR (Tokamak Fusion Test Reactor)

    SciTech Connect

    Nishitani, T.; Strachan, J.D.

    1988-05-01

    This paper describes the use of an /sup 3/He ionization chamber for neutron spectroscopy on TFTR during 1987. The ion temperature was measured using neutron spectroscopy for one set of ohmically heated plasmas. The deduced ion temperatures agreed to within 20% with those measured by other diagnostics. 11 refs., 11 figs., 1 tab.

  16. Fusion energy calorimeter for the tokamak fusion test reactor

    SciTech Connect

    Jassby, D.L.; Imel, G.R.

    1981-04-01

    One and two-dimensional neutronic analyses treating the transport and scattering of neutrons and the production and transport of gamma rays in the TFTR demonstrate that the fusion energy production in a D-T pulse in the TFTR can be determined with an uncertainty of +- 15% or less, simply by integrating the measured profile of temperature increase along the central radial axis of a large hydrocarbon moderator that fills the bay between adjacent toroidal-field coils, just outside the vacuum vessel. Limitations in thermopile temperature measurements dictate a minimum fusion-neutron fluence at the vacuum vessel of the order of 10/sup 12/ n/cm/sup 2/ per pulse (a source strength of 10/sup 18/ n/pulse in TFTR), in order that this simple calorimeter can provide useful accuracy.

  17. TFTR (Tokamak Fusion Test Reactor) neutral beam injected power measurement

    SciTech Connect

    Kamperschroer, J.H.; Grisham, L.R.; Dudek, L.E.; Gammel, G.M.; Johnson, G.A.; Kugel, H.W.; Lagin, L.; O'Connor, T.E.; Shah, P.A.; Sichta, P.

    1989-05-01

    Energy flow within TFTR neutral beamlines is measured with a waterfall calorimetry system capable of simultaneously measuring the energy deposited within four heating beamlines (three ion sources each), or of measuring the energy deposited in a separate neutral beam test stand. Of the energy extracted from the ion source in the well instrumented test stand, 99.5 +- 3.5% can be accounted for. When the ion deflection magnet is energized, however, 6.5% of the extracted energy is lost. This loss is attributed to a spray of devious particles onto unmonitored surfaces. A 30% discrepancy is also observed between energy measurements on the internal beamline calorimeter and energy measurements on a calorimeter located in the test stand target chamber. Particle reflection from the flat plate calorimeter in the target chamber, which the incident beam strikes at a near-grazing angle of 12/degree/, is the primary loss of this energy. A slight improvement in energy accountability is observed as the beam pulse length is increased. This improvement is attributed to systematic error in the sensitivity of the energy measurement to small fluctuations on the supply water temperature. An overall accuracy of 15% is estimated for the total power injected into TFTR. Contributions to this error are uncertainties in the beam neutralization efficiency, reionization and beam scrape-off in the drift duct, and fluctuations in the temperature of the supply water. 28 refs., 9 figs., 1 tab.

  18. Quasi-Isotropic Approximation of Geometrical Optics Method as Adequate Electrodynamical Basis for Tokamak Plasma Polarimetry

    NASA Astrophysics Data System (ADS)

    Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco

    Basic principles and recent findings of quasi-isotropic approximation (QIA) of a geometrical optics method are presented in a compact manner. QIA was developed in 1969 to describe electromagnetic waves in weakly anisotropic media. QIA represents the wave field as a power series in two small parameters, one of which is a traditional geometrical optics parameter, equal to wavelength ratio to plasma characteristic scale, and the other one is the largest component of anisotropy tensor. As a result, "" QIA ideally suits to tokamak polarimetry/interferometry systems in submillimeter range, where plasma manifests properties of weakly anisotropic medium.

  19. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle.

    PubMed

    Chektybayev, B; Shapovalov, G; Kolodeshnikov, A

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors. PMID:26026523

  20. The radial electric field in a tokamak with reversed magnetic shear

    NASA Astrophysics Data System (ADS)

    Zhu, P.; Horton, W.; Sugama, H.

    1999-06-01

    Neoclassical theory with the impurity rotational velocity is used to evaluate the radial electric field, Er, in a tokamak. The result of using the complete matrix method for the deuterium-carbon plasma is compared with a reduced analytic formula for determining Er [Ernst et al., Phys. Plasmas 5, 665 (1998)]. The analytic formula is shown to overestimate the Er magnitude and its gradient. Two transport measures of the effect of the Er shear are compared for the reversed shear and enhanced reversed shear discharges in Tokamak Fusion Test Reactor [Mazzucato et al., Phys. Rev. Lett. 77, 3145 (1996)]. It is shown that the combined Er and magnetic shear measure, Υs, from linear stability theory gives a higher correlation with the observed transition between the two discharges than the vorticity measure ωs from Er shear alone.

  1. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    SciTech Connect

    Chektybayev, B. Shapovalov, G.; Kolodeshnikov, A.

    2015-05-15

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  2. Visible wide angle view imaging system of KTM tokamak based on multielement image fiber bundle

    NASA Astrophysics Data System (ADS)

    Chektybayev, B.; Shapovalov, G.; Kolodeshnikov, A.

    2015-05-01

    In the paper, new visible wide angle view imaging system of KTM tokamak is described. The system has been designed to observe processes inside of plasma and the processes occurring due to plasma-wall interactions through the long equatorial port. Imaging system is designed based on special image fiber bundle and entrance wide angle lens, which provide image of large section of the vacuum chamber, both poloidal half-section and divertor through the sufficiently long equatorial port. The system also consists of two video cameras: slow and fast with image intensifier. Commercial equipment had been used in design of the system that allowed reducing the cost and time for research and development. The paper also discusses advantages and disadvantages of the system in comparison with conventional endoscopes based on a lens system and considers its promising utilization in future tokamaks and future steady state fusion reactors.

  3. Numerical verification of bounce-harmonic resonances in neoclassical toroidal viscosity for tokamaks.

    PubMed

    Kim, Kimin; Park, Jong-Kyu; Boozer, Allen H

    2013-05-01

    This Letter presents the first numerical verification for the bounce-harmonic (BH) resonance phenomena of the neoclassical transport in a tokamak perturbed by nonaxisymmetric magnetic fields. The BH resonances were predicted by analytic theories of neoclassical toroidal viscosity (NTV), as the parallel and perpendicular drift motions can be resonant and result in a great enhancement of the radial momentum transport. A new drift-kinetic δf guiding-center particle code, POCA, clearly verified that the perpendicular drift motions can reduce the transport by phase-mixing, but in the BH resonances the motions can form closed orbits and particles radially drift out fast. The POCA calculations on resulting NTV torque are largely consistent with analytic calculations, and show that the BH resonances can easily dominate the NTV torque when a plasma rotates in the perturbed tokamak and therefore, is a critical physics for predicting the rotation and stability in the International Thermonuclear Experimental Reactor. PMID:23683209

  4. Influence of collisions on parametric instabilities induced by lower hybrid waves in tokamak plasmas

    NASA Astrophysics Data System (ADS)

    Castaldo, C.; Di Siena, A.; Fedele, R.; Napoli, F.; Amicucci, L.; Cesario, R.; Schettini, G.

    2016-01-01

    Parametric instabilities induced at the plasma edge by lower hybrid wave power externally coupled to tokamak plasmas have, via broadening of the antenna spectrum, strong influence on the power deposition and current drive in the core. For modeling the parametric instabilities at the tokamak plasma edge in lower hybrid current drive experiments, the effect of the collisions has been neglected so far. In the present work, a specific collisional parametric dispersion relation, useful to analyze these nonlinear phenomena near the lower hybrid antenna mouth, is derived for the first time, based on a kinetic model. Numerical solutions show that in such cold plasma regions the collisions prevent the onset of the parametric instabilities. This result is important for present lower hybrid current drive experiments, as well as in fusion reactor scenarios.

  5. The simulation of L-H transition in tokamak plasma using MMM95 transport model

    NASA Astrophysics Data System (ADS)

    Intharat, P.; Chatthong, B.; Onjun, T.; Poolyarat, N.; Picha, R.

    2015-05-01

    BALDUR integrative predictive modelling code together with a Multimode (MMM95) anomalous transport model is used to simulate the evolution profiles, including plasma current, temperature, density and energy in a tokamak reactor. It is found that a self - transition from low confinement mode (L-mode) to high confinement mode (H-mode) regimes can be achieved once a sufficient auxiliary heating applied to the plasma is reached. The result agrees with experimental observations from various tokamaks. A strong reduction of turbulent transport near the edge of plasma is also observed, which is related to the formation of steep radial electric field near the edge regime. From transport analysis, it appears that the resistive ballooning mode is the dominant term near the plasma edge regime, which is significantly reduced during the transition.

  6. Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

    PubMed

    Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

    2012-01-01

    Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas. PMID:22233634

  7. Observation of Energetic Particle Driven Modes Relevant to Advanced Tokamak Regimes

    SciTech Connect

    R. Nazikian; B. Alper; H.L. Berk; D. Borba; C. Boswell; R.V. Budny; K.H. Burrell; C.Z. Cheng; E.J. Doyle; E. Edlund; R.J. Fonck; A. Fukuyama; N.N. Gorelenkov; C.M. Greenfield; D.J. Gupta; M. Ishikawa; R.J. Jayakumar; G.J. Kramer; Y. Kusama; R.J. La Haye; G.R. McKee; W.A. Peebles; S.D. Pinches; M. Porkolab; J. Rapp; T.L. Rhodes; S.E. Sharapov; K. Shinohara; J.A. Snipes; W.M. Solomon; E.J. Strait; M. Takechi; M.A. Van Zeeland; W.P. West; K.L. Wong; S. Wukitch; L. Zeng

    2004-10-21

    Measurements of high-frequency oscillations in JET [Joint European Torus], JT-60U, Alcator C-Mod, DIII-D, and TFTR [Tokamak Fusion Test Reactor] plasmas are contributing to a new understanding of fast ion-driven instabilities relevant to Advanced Tokamak (AT) regimes. A model based on the transition from a cylindrical-like frequency-chirping mode to the Toroidal Alfven Eigenmode (TAE) has successfully encompassed many of the characteristics seen in experiments. In a surprising development, the use of internal density fluctuation diagnostics has revealed many more modes than has been detected on edge magnetic probes. A corollary discovery is the observation of modes excited by fast particles traveling well below the Alfven velocity. These observations open up new opportunities for investigating a ''sea of Alfven Eigenmodes'' in present-scale experiments, and highlight the need for core fluctuation and fast ion measurements in a future burning-plasma experiment.

  8. Limestone compaction: an enigma

    USGS Publications Warehouse

    Shinn, Eugene A.; Halley, Robert B.; Hudson, J. Harold; Lidz, Barbara H.

    1977-01-01

    Compression of an undisturbed carbonate sediment core under a pressure of 556 kg/cm2 produced a “rock” with sedimentary structures similar to typical ancient fine-grained limestones. Surprisingly, shells, foraminifera, and other fossils were not noticeably crushed, which indicates that absence of crushed fossils in ancient limestones can no longer be considered evidence that limestones do not compact.

  9. Compact optical transconductance varistor

    SciTech Connect

    Sampayan, Stephen

    2015-09-22

    A compact radiation-modulated transconductance varistor device having both a radiation source and a photoconductive wide bandgap semiconductor material (PWBSM) integrally formed on a substrate so that a single interface is formed between the radiation source and PWBSM for transmitting PWBSM activation radiation directly from the radiation source to the PWBSM.

  10. COMPACT SCHOOL AND $$ SAVINGS.

    ERIC Educational Resources Information Center

    BAIR, W.G.

    A REVIEW OF THE CRITERIA FOR CONSIDERING THE USE OF A TOTAL ENERGY SYSTEM WITHIN A SCHOOL BUILDING STATES THE WINDOWLESS, COMPACT SCHOOL OFFERS MORE EFFICIENT SPACE UTILIZATION WITH LESS AREA REQUIRED FOR GIVEN STUDENT POPULATION AND LOWER OPERATION COSTS. THE AUTHOR RECOMMENDS THAT THESE BUILDINGS BE WINDOWLESS TO REDUCE HEAT COSTS, HOWEVER, AT…

  11. Summary of US compact torus experiments

    SciTech Connect

    Hartman, C.W.

    1981-03-11

    During the past several years a rapid increase has occurred in compact torus (CT) research in the United States, reflecting renewed interest in this simplified reactor consequences of this configuration. This paper reviews early approaches to CT formation and results and summarizes present experimental studies. Recent experiments have demonstrated a number of macroscopic aspects of the CT, including the conditions under which a macroscopically stable CT can be formed and maintained. Scaling experiments and more detailed studies of plasma transport in progress are discussed along with experiments under construction.

  12. Panel sees limited interest in compact nukes

    SciTech Connect

    Not Available

    1983-11-01

    Participants in the Joint Power Generation conference thought compact (200- to 300-MW) nuclear reactors would be useful to developing countries, but only the Canadians showed interest in becoming suppliers. Others said they would simply downsize existing designs. A 300-MW mini-Candu that can be built in 48 months will use proven components and have the same price tag as a full-sized unit. A market may develop in the future in the US and other industrialized countries for low-temperature heat sources. Another 5 to 10 developing countries would likely join the 7 now using nuclear power. (DCK)

  13. Bifurcated helical core equilibrium states in tokamaks

    NASA Astrophysics Data System (ADS)

    Cooper, W. A.; Chapman, I. T.; Schmitz, O.; Turnbull, A. D.; Tobias, B. J.; Lazarus, E. A.; Turco, F.; Lanctot, M. J.; Evans, T. E.; Graves, J. P.; Brunetti, D.; Pfefferlé, D.; Reimerdes, H.; Sauter, O.; Halpern, F. D.; Tran, T. M.; Coda, S.; Duval, B. P.; Labit, B.; Pochelon, A.; Turnyanskiy, M. R.; Lao, L.; Luce, T. C.; Buttery, R.; Ferron, J. R.; Hollmann, E. M.; Petty, C. C.; van Zeeland, M.; Fenstermacher, M. E.; Hanson, J. M.; Lütjens, H.

    2013-07-01

    Tokamaks with weak to moderate reversed central shear in which the minimum inverse rotational transform (safety factor) qmin is in the neighbourhood of unity can trigger bifurcated magnetohydrodynamic equilibrium states, one of which is similar to a saturated ideal internal kink mode. Peaked prescribed pressure profiles reproduce the ‘snake’ structures observed in many tokamaks which has led to a novel explanation of the snake as a bifurcated equilibrium state. Snake equilibrium structures are computed in simulations of the tokamak à configuration variable (TCV), DIII-D and mega amp spherical torus (MAST) tokamaks. The internal helical deformations only weakly modulate the plasma-vacuum interface which is more sensitive to ripple and resonant magnetic perturbations. On the other hand, the external perturbations do not alter the helical core deformation in a significant manner. The confinement of fast particles in MAST simulations deteriorate with the amplitude of the helical core distortion. These three-dimensional bifurcated solutions constitute a paradigm shift that motivates the applications of tools developed for stellarator research in tokamak physics investigations.

  14. Spheromak reactor-design study

    SciTech Connect

    Les, J.M.

    1981-06-30

    A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

  15. Microtearing modes in tokamak discharges

    NASA Astrophysics Data System (ADS)

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

  16. Advanced Construction of Compact Containment BWR

    SciTech Connect

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-07-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  17. Compact Nanowire Sensors Probe Microdroplets.

    PubMed

    Schütt, Julian; Ibarlucea, Bergoi; Illing, Rico; Zörgiebel, Felix; Pregl, Sebastian; Nozaki, Daijiro; Weber, Walter M; Mikolajick, Thomas; Baraban, Larysa; Cuniberti, Gianaurelio

    2016-08-10

    The conjunction of miniature nanosensors and droplet-based microfluidic systems conceptually opens a new route toward sensitive, optics-less analysis of biochemical processes with high throughput, where a single device can be employed for probing of thousands of independent reactors. Here we combine droplet microfluidics with the compact silicon nanowire based field effect transistor (SiNW FET) for in-flow electrical detection of aqueous droplets one by one. We chemically probe the content of numerous (∼10(4)) droplets as independent events and resolve the pH values and ionic strengths of the encapsulated solution, resulting in a change of the source-drain current ISD through the nanowires. Further, we discuss the specificities of emulsion sensing using ion sensitive FETs and study the effect of droplet sizes with respect to the sensor area, as well as its role on the ability to sense the interior of the aqueous reservoir. Finally, we demonstrate the capability of the novel droplets based nanowire platform for bioassay applications and carry out a glucose oxidase (GOx) enzymatic test for glucose detection, providing also the reference readout with an integrated parallel optical detector. PMID:27417510

  18. Sawtooth Instability in the Compact Toroidal Hybrid

    NASA Astrophysics Data System (ADS)

    Herfindal, J. L.; Maurer, D. A.; Hartwell, G. J.; Ennis, D. A.; Knowlton, S. F.

    2015-11-01

    Sawtooth instabilities have been observed in the Compact Toroidal Hybrid (CTH), a current-carrying stellarator/tokamak hybrid device. The sawtooth instability is driven by ohmic heating of the core plasma until the safety factor drops below unity resulting in the growth of an m = 1 kink-tearing mode. Experiments varying the vacuum rotational transform from 0.02 to 0.13 are being conducted to study sawtooth property dependance on vacuum flux surface structure. The frequency of the sawtooth oscillations increase from 2 kHz to 2.8 kHz solely due the decrease in rise time of the oscillation, the crash time is unchanged. CTH has three two-color SXR cameras, a three-channel 1mm interferometer, and a new bolometer system capable of detecting the signatures of sawtooth instabilities. The new bolometer system consists of two cameras, each containing a pair of diode arrays viewing the plasma directly or through a beryllium filter. Electron temperature measurements are found with the two-color SXR cameras through a ratio of the SXR intensities. Impurity radiation can drastically affect the electron temperature measurement, therefore new filters consisting of aluminum and carbon were selected to avoid problematic line radiation while maximizing the signal for a 100 eV plasma. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

  19. Global gyrokinetic simulation of tokamak transport

    SciTech Connect

    Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T. |

    1998-10-01

    A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or {eta}{sub i}({eta}{sub i} {equivalent_to} {partial_derivative}{ell}nT{sub i}/{partial_derivative}{ell}n n{sub i}) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling.

  20. Modeling plasma/material interactions during a tokamak disruption

    SciTech Connect

    Hassanein, A.; Konkashbaev, I.

    1994-10-01

    Disruptions in tokamak reactors are still of serious concern and present a potential obstacle for successful operation and reliable design. Erosion of plasma-facing materials due to thermal energy dump during a disruption can severely limit the lifetime of these components, therefore diminishing the economic feasibility of the reactor. A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The initial burst of energy delivered to facing-material surfaces from direct impact of plasma particles causes sudden ablation of these materials. As a result, a vapor cloud is formed in front of the incident plasma particles. Shortly thereafter, the plasma particles are stopped in the vapor cloud, heating and ionizing it. The energy transmitted to the material surfaces is then dominated by photon radiation. It is the dynamics and the evolution of this vapor cloud that finally determines the net erosion rate and, consequently, the component lifetime. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics, and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed.

  1. Merchant Marine Ship Reactor

    DOEpatents

    Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

    1961-05-01

    A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

  2. MERCHANT MARINE SHIP REACTOR

    DOEpatents

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  3. REACTOR CONTROL DEVICE

    DOEpatents

    Graham, R.H.

    1962-09-01

    A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

  4. Parametric studies of next-step spherical tokamaks using high-temperature superconductors

    NASA Astrophysics Data System (ADS)

    Menard, Jonathan; Brown, Tom; El-Guebaly, Laila; Zhai, Yuhu

    2015-11-01

    High-temperature superconducting (HTS) magnets are potentially attractive for compact spherical tokamak (ST) applications due to higher operating temperature which could reduce thermal shielding requirements and reduce device size relative to configurations that utilize low-temperature superconductors (LTS). HTS conductors can also operate with very high current densities and high magnetic fields. Recent studies have shown that for elongation and kink stability dependencies on aspect ratio consistent with NSTX data, accounting for the engineering limits of HTS magnets, and having only a modest central solenoid, the optimal aspect ratio for an HTS tokamak pilot plant is between A = 1.7 and 2.3 depending on inboard shielding thickness. These results point to the interesting finding that the optimal aspect ratio for a compact HTS pilot plant may be near A = 2 which is an unexplored configuration in the present fusion program. The potential implications for required energy confinement, plasma formation and sustainment, power exhaust, and other performance parameters will be discussed. Work Supported by U.S.D.O.E. Contract No. DE-AC02-09CH11466.

  5. Compact Spreader Schemes

    SciTech Connect

    Placidi, M.; Jung, J. -Y.; Ratti, A.; Sun, C.

    2014-07-25

    This paper describes beam distribution schemes adopting a novel implementation based on low amplitude vertical deflections combined with horizontal ones generated by Lambertson-type septum magnets. This scheme offers substantial compactness in the longitudinal layouts of the beam lines and increased flexibility for beam delivery of multiple beam lines on a shot-to-shot basis. Fast kickers (FK) or transverse electric field RF Deflectors (RFD) provide the low amplitude deflections. Initially proposed at the Stanford Linear Accelerator Center (SLAC) as tools for beam diagnostics and more recently adopted for multiline beam pattern schemes, RFDs offer repetition capabilities and a likely better amplitude reproducibility when compared to FKs, which, in turn, offer more modest financial involvements both in construction and operation. Both solutions represent an ideal approach for the design of compact beam distribution systems resulting in space and cost savings while preserving flexibility and beam quality.

  6. Compact waveguide splitter networks.

    PubMed

    Qian, Yusheng; Song, Jiguo; Kim, Seunghyun; Hu, Weisheng; Nordin, Gregory P

    2008-03-31

    We demonstrate compact waveguide splitter networks in siliconon- insulator (SOI) rib waveguides using trench-based splitters (TBSs) and bends (TBBs). Rather than a 90 degrees geometry, we use 105 degrees TBSs to facilitate reliable fabrication of high aspect ratio trenches suitable for 50/50 splitting when filled with SU8. Three dimensional (3D) finite difference time domain (FDTD) simulation is used for splitter and bend design. Measured TBB and TBS optical efficiencies are 84% and 68%, respectively. Compact 105 degrees 1 x 4, 1 x 8, and 1 x 32 trench-based splitter networks (TBSNs) are demonstrated. The measured total optical loss of the 1 x 32 TBSN is 9.15 dB. Its size is only 700 microm x 1600 microm for an output waveguide spacing of 50 microm. PMID:18542598

  7. Tokamak Spectroscopy for X-Ray Astronomy

    NASA Technical Reports Server (NTRS)

    Fournier, Kevin B.; Finkenthal, M.; Pacella, D.; May, M. J.; Soukhanovskii, V.; Mattioli, M.; Leigheb, M.; Rice, J. E.

    2000-01-01

    This paper presents the measured x-ray and Extreme Ultraviolet (XUV) spectra of three astrophysically abundant elements (Fe, Ca and Ne) from three different tokamak plasmas. In every case, each spectrum touches on an issue of atomic physics that is important for simulation codes to be used in the analysis of high spectral resolution data from current and future x-ray telescopes. The utility of the tokamak as a laboratory test bed for astrophysical data is demonstrated. Simple models generated with the HULLAC suite of codes demonstrate how the atomic physics issues studied can affect the interpretation of astrophysical data.

  8. Electron cyclotron emission diagnostics on KSTAR tokamak.

    PubMed

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration. PMID:21033954

  9. Electron cyclotron emission diagnostics on KSTAR tokamak

    SciTech Connect

    Jeong, S. H.; Lee, K. D.; Kwon, M.; Kogi, Y.; Kawahata, K.; Nagayama, Y.; Mase, A.

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  10. Compact heat exchangers

    SciTech Connect

    Kays, W.M.; London, A.L.

    1984-01-01

    This third edition is an update of the second edition published in 1964. New data and more modern theoretical solutions for flow in the simple geometries are included, although this edition does not differ radically from the second edition. It contains basic test data for eleven new surface configurations, including some of the very compact ceramic matrices. Al dimensions are given in both the English and the Systeme International (SI) system of units.

  11. Compact infrared detector

    NASA Technical Reports Server (NTRS)

    Gupta, A.; Hong, S.; Moacanin, J.

    1981-01-01

    Broadband IR detector integrated into compact package for pollution monitoring and weather prediction is small, highly responsive, and immune to high noise. Sensing material is transparent sheet metalized with reflecting coating and overcoated with black material on same side. Pulse produced by chopping of infrared source beam creates transient "thermal lens" that temporarily defocuses laser beam probe. Detector monitoring beam measures defocusing which parallels infrared intensity.

  12. Electron cyclotron resonance heating in the microwave tokamak experiment

    SciTech Connect

    Allen, S.L.; Casper, T.A.; Fenstermacher, M.E.

    1992-09-01

    This paper presents the results from a series of Electron Cyclotron Resonance Heating (ECRH) experiments on the Microwave Tokamak Experiment (MTX). On-axis heating at B{sub T} = 5T (f{sub ce} = 140 GHz) has been performed at electron densities up to cutoff. We have used both a long-pulse gryotron ({approximately}200 kW, {approximately}0.1s) and a pulsed Free Electron Laser (FEL) as microwave sources. Gyrotron experiments with power densities corresponding to 4 MW m{sup {minus}3}. A far infrared (FIR) polarimeter measured peaking of plasma current profiles in some discharges during the ECRH pulse. During high-power single-pulse FEL experiments, single-pass microwave !transmission measurements show nonlinear effects; i.e., higher transmission than predicted by linear theory. A corrugated-wall duct was used in the tokamak port to increase the gradient of the parallel refractive index n{sub parallel} of the incident wave, and increased absorption was observed. Evidence of electron tail heating during FEL pulses was observed on soft x-ray and ECE diagnostics. These results are in agreement with predictions of nonlinear theory; extrapolation of this theory to reactor-like conditions indicates efficient absorption and heating. A Laser Assisted Particle Probe Spectroscopy (LAPPS) diagnostic provided estimates of the vacuum electric field of the FEL which were consistent with the measured power. Multiple pulse operation of the ETA-II accelerator for the FEL has also been demonstrated, indicating the feasibility of high-average power FEL operation.

  13. Granule consolidation during compaction.

    PubMed

    Rubinstein, M H

    1976-03-01

    The deformation of small cylindrical aggregates of dibasic calcium phosphate was measured during compaction. An analogy between these aggregates and cylindrical granules was proposed. No change in the original shape of the aggregates occurred; the cylindrical shape was maintained even at high compaction pressures. Relaxation of the aggregates occurred at pressures higher than 420 MNm-2 (60.9 x 10(3) lb in.-2) when removed from the compacts, but no relaxation took place at pressures below this value. In addition, the aggregates relaxed by an increase in thickness only; there was no corresponding change in diameter. Up to a pressure of 200 MNm-2 (29.0 x 10(3) lb in.-2), an increase in aggregate diameter occurred, which was accompanied by a reduction in thickness. This change produced only a small reduction in volume, which was attributable to interparticulate slippage resulting in a closer packed arrangement. At a pressure of 200 MNm-2, the aggregate diameter no longer increased because solid bridges were formed between the particles and the die wall, preventing further spreading. From 200 to 420 MNm-2, failure of the material occurred by plastic deformation, which produced only a decrease in aggregate thickness. From 420 to 800 MNm-2 (116.0 x 10(3) lb in.-2), a structure was formed that could support the applied load without further reduction of thickness, and this structure was shown to behave elastically. PMID:1263085

  14. Rotation of tokamak halo currents

    SciTech Connect

    Boozer, Allen H.

    2012-05-15

    During tokamak disruptions, halo currents, which can be tenths of the total plasma current, can flow at the plasma edge along the magnetic field lines that intercept the chamber walls. Non-axisymmetric halo currents are required to maintain force balance as the plasma kinks when the edge safety factor drops to about two in a vertical displacement event. The plasma quickly assumes a definite toroidal velocity v{sub a}(r) with respect to that of the magnetic kink, v{sub k}, where v{sub a}(r) is set by the radial electric field required for ambipolarity. The plasma velocity, v{sub pl}=v{sub a}+v{sub k}, near the edge is influenced by the interaction with neutrals and with the potential in the halo required for quasi-neutrality on open magnetic field lines, and the plasma velocity in the core is influenced by external error fields. When plasma effects dominate magnetic locking, the magnetic kink should rotate at a diamagnetic speed of either the edge or the core. If the magnetic field lines of the halo plasma intercept the wall at locations of very different electrical conductivity, the toroidal rotation of the halo currents can intermittently stall at wall locations of high conductivity. Such stalling is seen in experiments. The toroidal phase difference between the stalled halo currents and the kink, which is expected to rotate smoothly, must satisfy {delta}{phi}<{+-}{pi}/2. A concern cited by ITER engineers is that the time varying force of the rotating halo could substantially increase the disruption loads on in-vessel components.

  15. Rotation of tokamak halo currents

    NASA Astrophysics Data System (ADS)

    Boozer, Allen H.

    2012-05-01

    During tokamak disruptions, halo currents, which can be tenths of the total plasma current, can flow at the plasma edge along the magnetic field lines that intercept the chamber walls. Non-axisymmetric halo currents are required to maintain force balance as the plasma kinks when the edge safety factor drops to about two in a vertical displacement event. The plasma quickly assumes a definite toroidal velocity va(r) with respect to that of the magnetic kink, vk, where va(r) is set by the radial electric field required for ambipolarity. The plasma velocity, vpl=va+vk, near the edge is influenced by the interaction with neutrals and with the potential in the halo required for quasi-neutrality on open magnetic field lines, and the plasma velocity in the core is influenced by external error fields. When plasma effects dominate magnetic locking, the magnetic kink should rotate at a diamagnetic speed of either the edge or the core. If the magnetic field lines of the halo plasma intercept the wall at locations of very different electrical conductivity, the toroidal rotation of the halo currents can intermittently stall at wall locations of high conductivity. Such stalling is seen in experiments. The toroidal phase difference between the stalled halo currents and the kink, which is expected to rotate smoothly, must satisfy δϕ <±π/2. A concern cited by ITER engineers is that the time varying force of the rotating halo could substantially increase the disruption loads on in-vessel components.

  16. Increase in compact toroid mass by accelerator-region ionization of high-Z noble gas on CTIX

    NASA Astrophysics Data System (ADS)

    Horton, Robert D.; Hwang, David Q.; Liu, Fei; Hong, Sean; Klauser, Ruth; Evans, Russell W.; Buchenauer, Dean A.

    2012-10-01

    A promising technique for runaway electron (RE) mitigation in large-tokamak disruptions is the injection of compact toroid (CT) plasmas of high atomic number. With sufficient kinetic energy density, high-Z CTs can reach the tokamak magnetic axis where RE effects are strongest. At CT velocities of 100 km/s or more, penetration to the axis occurs on a sub-millisecond time scale. In addition to reducing avalanche RE production by collisions, high-Z CTs can cool RE by bremsstrahung effects. From theoretical calculations, using Xe ions, bremsstrahlung cooling exceeds the effect of collisions at RE energy above about 10 MeV, a value expected to be well exceeded in large tokamaks. Past experiments on the CTIX compact-toroid injector have demonstrated increased CT mass using snowplow accretion of puffed noble gas by an initial hydrogenic CT. These experiments will be continued using a higher ratio of accreted high-Z plasma to H plasma, to maximize CT kinetic energy density. Results will be compared with a 1D model using external circuit effects, coaxial railgun kinetics, and ionization. The model will be used to predict performance of CT injectors of greater energy, suitable for RE suppression on mid-sized tokamaks.

  17. Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

    SciTech Connect

    Majeski, R.

    2010-05-20

    At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700 deg. C. However, at a sufficiently high operating temperature (700 - 1000 deg. C), tungsten is self-annealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment.The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium.Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands and fusion experiments to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 deg. C), it has been now been used as a PFC in several confinement experiments (TFTR, T11-M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

  18. Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

    NASA Astrophysics Data System (ADS)

    Majeski, R.

    2010-05-01

    At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700°C. However, at a sufficiently high operating temperature (700 - 1000 °C), tungsten is self-annealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment. The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium. Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands and fusion experiments to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 °C), it has been now been used as a PFC in several confinement experiments (TFTR, T11-M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

  19. Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

    SciTech Connect

    R. Majeski

    2010-01-15

    At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700oC. However, at a sufficiently high operating temperature (700 - 1000 oC), tungsten is selfannealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment. The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium. Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands1 and fusion experiments2,3 to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 oC), it has been now been used as a PFC in several confinement experiments (TFTR, T11- M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

  20. Component Manufacturing Development for the National Compact Stellarator Experiment (NCSX)

    SciTech Connect

    P.J. Heitzenroeder; T.G. Brown; J.H. Chrzanowski; M.J. Cole; P.L. Goranson; G.H. Neilson; B.E. Nelson; W.T. Reiersen; L.L Sutton; D.E. Williamson; M.E. Viola

    2004-10-28

    NCSX [National Compact Stellarator Experiment] is the first of a new class of stellarators called compact stellarators which hold the promise of retaining the steady state feature of the stellarator but at a much lower aspect ratio and using a quasi-axisymmetric magnetic field to obtain tokamak-like performance. Although much of NCSX is conventional in design and construction, the vacuum vessel and modular coils provide significant engineering challenges due to their complex shapes, need for high dimensional accuracy, and the high current density required in the modular coils due space constraints. Consequently, a three-phase development program has been undertaken. In the first phase, laboratory/industrial studies were performed during the development of the conceptual design to permit advances in manufacturing technology to be incorporated into NCSX's plans. In the second phase, full-scale prototype modular coil winding forms, compacted cable conductors, and 20 degree sectors of the vacuum vessel were fabricated in industry. In parallel, the NCSX project team undertook R&D studies that focused on the windings. The third (production) phase began in September 2004. First plasma is scheduled for January 2008.

  1. Scale-up of spherical tokamak solenoid-free startup by coaxial helicity injection

    SciTech Connect

    Tang, X. Z.; Boozer, A. H.

    2007-10-15

    Current multiplication and flux amplification are two critical measures in assessing the usefulness of magnetic helicity injection for forming the plasma confining magnetic field in laboratory spheromak and spherical tokamak (ST) experiments. While the two concepts are closely related for spheromaks, they are independent for a Taylor-relaxed ST plasma, and negatively correlated for a more realistic, partially relaxed ST plasma. An important application of this understanding leads to the so-called relaxed transient coaxial helicity injection scheme for solenoid-free ST startup, which can deliver reactor-grade high current multiplication and flux amplification.

  2. Physics Studies of a Proposed Small Aspect Ratio Torsatron-Tokamak Hybrid

    NASA Astrophysics Data System (ADS)

    Valanju, P. M.; Gentle, K. W.; McCool, S.; Miner, W. H., Jr.; Ross, D. W.; Wiley, J. C.; Wootton, A. J.

    1996-11-01

    Physics studies of a proposed small aspect ratio torsatron-tokamak hybrid (SMARTH) are presented. This small aspect ratio configuration attempts to remedy a major drawback of conventional stellarator reactors, viz. the large volume resulting from the large aspect ratio. Considering a small aspect ratio machine with 8 field periods (m=8) and with 8 or 16 modular coils, we investigate vacuum configurations and MHD equilibria to determine the optimum configuration which produces large rotational transform, large plasma volume, large beta, and small field ripple. Implications for particle confinement will also be discussed. Effects of radial electric fields on orbit losses and confinement will also be explored.

  3. Electron cyclotron emission in a rf current-driven tokamak plasma

    SciTech Connect

    Giruzzi, G.; Fidone, I.; Granata, G.; Meyer, R.L.

    1984-07-01

    The general properties of electron cyclotron radiation emitted by the current-carrying superthermal tail in tokamak plasmas are investigated. Two situations are considered, namely, the case of an extended tail as it occurs in present-day experiments and that in which the velocity range of the tail is much less than its mean speed. For the former case, we show that most of the emitted radiation obeys Kirchhoff's law. In the latter case, which is suited for reactor-like parameters, a simple relation is obtained between the tail momentum distribution and the radiation temperature.

  4. Design of multipulse Thomson scattering diagnostic for SST-1 tokamak

    NASA Astrophysics Data System (ADS)

    Kumar, Ajai; Chavda, Chhaya; Saxena, Y. C.; Singh, Ranjeet; Thakar, Aruna; Thomas, Jinto; Patel, Kiran; Pandya, Kaushal; Bedakihale, Vijay

    2007-04-01

    A multipulse Nd:YAG (Yttrium aluminum garnet) Thomson scattering (TS) system is designed and developed for measuring electron temperature (Te) and density (ne) profiles of SST-1 tokamak. The system operates at vertical, divertor, and horizontal (midplane) regions of plasma and measures the electron temperature of 20eVto1.5keV and density of 1018-1019m-3. Six Nd:YAG lasers synchronized with external control is used to get three different temporal resolutions (30Hz, 180Hz, and 1kHz). The entire system is laboratory tested for the stability of alignment and performance over a distance of 30m. Different imaging lens assemblies are designed to image the scattered photons from each of the scattering region to an array of optical fibers. A low cost and compact five-channel interference filter polychromator is designed, fabricated, and tested for its image quality and the filter transmission characteristics. Detection system with an avalanche photodiode and required signal conditioning electronics is developed for detecting the scattered photons. A data acquisition and control module operating on PXI bus is developed for the real time data acquisition and system control. A detailed description of design and testing of TS subsystems is presented in this article.

  5. Thomson scattering diagnostic for the Microwave Tokamak Experiment

    SciTech Connect

    Foote, J.H.; Barter, J.D.; Sewall, N.R.; Jolly, J.J.; Schlander, L.F.

    1990-05-04

    The Thomson-scattering diagnostic system (TSS) on the Microwave Tokamak Experiment (MTX) at LLNL routinely monitors electron temperature (T{sub e}) and density. Typical measured values at the plasma center under clean conditions are 900 {plus minus} 70 eV and 1 to 2 {times} 10{sup 14} ({plus minus}30%) cm{sup {minus}3}. The TSS apparatus is compact, with all elements mounted on one sturdy, two-level optics table. Because of this, we maintain with minimum effort the alignment of both the ruby-laser input optics and the scattered-light collecting optics. Undesired background signals, e.g., plasma light as well as ruby-laser light scattered off obstacles and walls, are generally small compared with the Thomson-scattered signals we normally detect. In the MTX T{sub e} region, the TSS data are definitely fitted better when relativistic effects are included in the equations. Besides determining the temperature of the Maxwellian electron distribution, the system is designed to detect electron heating from GW-level free-electron laser (FEL) pulses by measuring large wavelength shifts of the scattered laser photons. TSS data suggest that we may indeed by able to detect these electrons, which can have energies up to 10 keV, according to computer simulation. 7 refs., 4 figs.

  6. Neutral particle analyzer diagnostics on the TCV tokamak

    SciTech Connect

    Karpushov, Alexander N.; Duval, Basil P.; Schlatter, Christian; Afanasyev, Valery I.; Chernyshev, Fedor V.

    2006-03-15

    Experimental apparatus and data analysis techniques used in neutral particle analyzer (NPA) diagnostics on the Tokamak a Configuration Variable (TCV) are described. Two NPAs are used on TCV to measure the energy spectrum of neutral particle fluxes from the plasma. The 'five-channel energy analyzer of atomic particles' used in double electrical analysis mode with fast voltage sweeping detect particles without atomic mass discrimination in the energy range of 0.6-8.0 keV with a time resolution of 0.5-2.0 ms and an energy resolution of 7%-20%. The 28-channel ''compact neutral particle analyzer'' (CNPA) is an EIIB spectrometer with mass and energy separations designed for medium sized fusion machines featuring a carbon neutral stripping foil, a permanent magnet for dispersion, and channel-electron multiplier detectors. The CNPA simultaneously detects two mass species [hydrogen (H) and deuterium (D) or D and helium (He)] in the 0.5-50 keV energy range with a resolution of 60%-10% and a time resolution of 0.5-4.0 ms. The CNPA views the plasma across the path of the diagnostic neutral beam and can perform active charge-exchange NPA measurement. Data analysis procedures and numerical algorithms developed for NPA measurement are routinely used on TCV to obtain information on the plasma ion temperature, ion energy distribution function, plasma isotope ratios, and other plasma characteristics.

  7. Lower hybrid system design for the Tokamak physics experiment

    SciTech Connect

    Goranson, P.L.; Conner, D.L.; Swain, D.W.; Yugo, J.J.; Bernabei, S.; Greenough, N.

    1995-12-31

    The lower hybrid (LH) launcher configuration has been redesigned to integrate the functions of the vertical four-way power splitter and the front waveguide array (front array). This permits 256 waveguide channels to be fed by only 64 waveguides at the vacuum window interface. The resulting configuration is a more compact coupler, which incorporates the simplicity of a multijunction coupler while preserving the spectral flexibility of a conventional lower hybrid launcher. Other spin-offs of the redesign are reduction in thermal incompatibility between the front array and vacuum windows, improved maintainability, in situ vacuum window replacement, a reduced number of radio frequency (rf) connections, and a weight reduction of 7300 kg. There should be a significant cost reduction as well. Issues associated with the launcher design and fabrication have been addressed by a research and development program that includes brazing of the front array and testing of the power splitter configuration to confirm that phase errors due to reflections in the shorted splitter legs will not significantly impact the rf spectrum. The Conceptual Design Review requires that radiation levels at the torus radial port mounting flange and outer surface of the toroidal field coils should be sufficiently low to permit hands-on maintenance. Low activation materials and neutron shielding are incorporated in the launcher design to meet these requirements. The launcher is configured to couple 3 MW of steady state LH heating/LH current drive power at 3.7 GHz to the Tokamak Physics Experiment plasma.

  8. Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

    NASA Astrophysics Data System (ADS)

    Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; Holcomb, C. T.; Lao, L. L.; McKee, G. R.; Meneghini, O.; Staebler, G. M.; Grierson, B. A.; Qian, J. P.; Solomon, W. M.; Turnbull, A. D.; Holland, C.; Guo, W. F.; Ding, S. Y.; Pan, C. K.; Xu, G. S.; Wan, B. N.

    2016-06-01

    Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of βp and βN , despite strong internal transport barriers. Good confinement has been achieved with reduced toroidal rotation. These high βp plasmas challenge the energy transport understanding, especially in the electron energy channel. A new turbulent transport model, named TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. More investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.

  9. Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

    SciTech Connect

    Malik, M.A.

    1988-01-01

    There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such as INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism.

  10. Electron cyclotron heating at down-shifted frequencies in existing tokamak devices

    SciTech Connect

    Mazzucato, E.; Fidone, I.; Giruzzi, G.; Krivenski, V.

    1985-06-01

    Plasma heating in existing tokamak devices by electron cyclotron waves with frequency (f) significantly smaller than the electron gyrofrequency (f/sub c/) is investigated for the case of Maxwellian plasmas. It is shown that for central electron temperatures larger than 3 keV, strong absorption of extraordinary waves can occur at values of toroidal field for which the condition f = f/sub c/ is not satisfied in the plasma region. The cases of f = 60 GHz and f = 100 GHz are discussed for the PLT and TFTR devices as representative of medium (approx. =30 kG) and high (approx. =50 kG) magnetic field tokamaks, respectively. Numerical calculations with a ray tracing code indicate that most of the rf energy is absorbed in a central plasma region. These results are of practical interest because they significantly simplify the main technical problem of ECH in a tokamak reactor, i.e., the development of high frequency and high power microwave sources.

  11. Advanced Tokamak Regimes in Alcator C-Mod with Lower Hybrid Current Drive

    NASA Astrophysics Data System (ADS)

    Parker, R.; Bonoli, P.; Gwinn, D.; Hutchinson, I.; Porkolab, M.; Ramos, J.; Bernabei, S.; Hosea, J.; Wilson, R.

    1999-11-01

    Alcator C-Mod has been proposed as a test-bed for developing advanced tokamak scenarios owing to its strong shaping, relatively long pulse length capability at moderate field, e.g. t ~ L/R at B = 5T and T_eo ~ 7keV, and the availability of strong ICRF heating. We plan to exploit this capability by installing up to 4 MW RF power at 4.6 GHz for efficient off-axis current drive by lower hybrid waves. By launching LH waves with a grill whose n_xx spectrum can be dynamically controlled over the range 2 < n_xx < 3.5, the driven current profile can be modified so that, when combined with bootstrap current in high ɛβ_pol regimes, q_min > 2. Such reversed or nearly zero shear regimes have already been proposed as the basis of an advanced tokamak burning-plasma experiment-ATBX (M. Porkolab et al, IAEA-CN-69/FTP/13, IAEA,Yokohama 1998.), and could provide the basis for a demonstration power reactor. Theoretical and experimental basis for this advanced tokamak research program on C-Mod, including design of the lower hybrid coupler, its spectrum and current drive capabilities will be presented.

  12. Tokamak operation with safety factor q95 < 2 via control of MHD stability.

    PubMed

    Piovesan, P; Hanson, J M; Martin, P; Navratil, G A; Turco, F; Bialek, J; Ferraro, N M; La Haye, R J; Lanctot, M J; Okabayashi, M; Paz-Soldan, C; Strait, E J; Turnbull, A D; Zanca, P; Baruzzo, M; Bolzonella, T; Hyatt, A W; Jackson, G L; Marrelli, L; Piron, L; Shiraki, D

    2014-07-25

    Magnetic feedback control of the resistive-wall mode has enabled the DIII-D tokamak to access stable operation at safety factor q(95) = 1.9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at a given toroidal magnetic field. In tokamaks with a divertor, the limit occurs at q(95) = 2, as confirmed in DIII-D. Since the energy confinement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a whole new high-current regime not accessible before. This result brings significant possible benefits in terms of fusion performance, but it also extends resistive-wall mode physics and its control to conditions never explored before. In present experiments, the q(95) < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime. PMID:25105626

  13. Elementary Processes Underlying Alpha Channeling in Tokamaks

    SciTech Connect

    NM.J. Fisch

    2012-06-15

    Alpha channeling in tokamaks is speculative, but also extraordinarily attractive. Waves that can accomplish this effect have been identified. Key aspects of the theory now enjoy experimental confirmation. This paper will review the elementary processes of wave-particle interactions in plasma that underlie the alpha channeling effect

  14. Toroidal Alfven wave stability in ignited tokamaks

    SciTech Connect

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

  15. Microinstabilities in weak density gradient tokamak systems

    SciTech Connect

    Tang, W.M.; Rewoldt, G.; Chen, L.

    1986-04-01

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient.

  16. Analysis of sawtooth relaxation oscillations in tokamaks

    SciTech Connect

    Yamazaki, K.; McGuire, K.; Okabayashi, M.

    1982-07-01

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated.

  17. Stabilization of tokamak plasma by lithium streams

    SciTech Connect

    L.E. Zakharov

    2000-08-07

    The stabilization theory of free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. While the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.

  18. Transformer Recharging with Alpha Channeling in Tokamaks

    SciTech Connect

    N.J. Fisch

    2009-12-21

    Transformer recharging with lower hybrid waves in tokamaks can give low average auxiliary power if the resistivity is kept high enough during the radio frequency (rf) recharging stage. At the same time, operation in the hot ion mode via alpha channeling increases the effective fusion reactivity. This paper will address the extent to which these two large cost saving steps are compatible. __________________________________________________

  19. Banana drift transport in tokamaks with ripple

    SciTech Connect

    Linsker, R.; Boozer, A.H.

    1982-01-01

    Ripple transport in tokamaks is discussed for the ''banana drift'' collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and consequently the resulting transport coefficients can differ by several orders of magnitude.

  20. Banana drift transport in tokamaks with ripple

    SciTech Connect

    Linsker, R.; Boozer, A.H.

    1981-04-01

    Ripple transport in tokamaks is discussed for the banana drift collisionality regime, which lies below the ripple plateau regime treated earlier. The physical mechanisms that dominate banana drift transport are found to differ from those considered in previous work on this regime, and the resulting transport coefficients can consequently differ by several orders of magnitude.

  1. UCLA Tokamak Program Close Out Report.

    SciTech Connect

    Taylor, Robert John

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  2. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    SciTech Connect

    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

    2003-10-01

    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.

  3. Spontaneous generation of rotation in tokamak plasmas

    SciTech Connect

    Parra Diaz, Felix

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  4. Tokamak startup with electron cyclotron heating

    SciTech Connect

    Holly, D J; Prager, S C; Shepard, D A; Sprott, J C

    1980-04-01

    Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.

  5. Photometry of compact galaxies.

    NASA Technical Reports Server (NTRS)

    Shen, B. S. P.; Usher, P. D.; Barrett, J. W.

    1972-01-01

    Photometric histories of the N galaxies 3C 390.3 and PKS 0521-36. Four other compact galaxies, Markarian 9, I Zw 92, 2 Zw 136, and III Zw 77 showed no evidence of variability. The photometric histories were obtained from an exhaustive study of those plates of the Harvard collection taken with large aperture cameras. The images of all galaxies reported were indistinguishable from stars due to the camera f-ratios and low surface brightness of the outlying nebulosities of the galaxies. Standard techniques for the study of variable stars are therefore applicable.

  6. Compact Q-balls

    NASA Astrophysics Data System (ADS)

    Bazeia, D.; Losano, L.; Marques, M. A.; Menezes, R.; da Rocha, R.

    2016-07-01

    In this work we deal with non-topological solutions of the Q-ball type in two space-time dimensions, in models described by a single complex scalar field that engenders global symmetry. The main novelty is the presence of stable Q-balls solutions that live in a compact interval of the real line and appear from a family of models controlled by two distinct parameters. We find analytical solutions and study their charge and energy, and show how to control the parameters to make the Q-balls classically and quantum mechanically stable.

  7. Compact LINAC for deuterons

    SciTech Connect

    Kurennoy, S S; O' Hara, J F; Rybarcyk, L J

    2008-01-01

    We are developing a compact deuteron-beam accelerator up to the deuteron energy of a few MeV based on room-temperature inter-digital H-mode (IH) accelerating structures with the transverse beam focusing using permanent-magnet quadrupoles (PMQ). Combining electromagnetic 3-D modeling with beam dynamics simulations and thermal-stress analysis, we show that IHPMQ structures provide very efficient and practical accelerators for light-ion beams of considerable currents at the beam velocities around a few percent of the speed of light. IH-structures with PMQ focusing following a short RFQ can also be beneficial in the front end of ion linacs.

  8. Compact multiframe blind deconvolution.

    PubMed

    Hope, Douglas A; Jefferies, Stuart M

    2011-03-15

    We describe a multiframe blind deconvolution (MFBD) algorithm that uses spectral ratios (the ratio of the Fourier spectra of two data frames) to model the inherent temporal signatures encoded by the observed images. In addition, by focusing on the separation of the object spectrum and system transfer functions only at spatial frequencies where the measured signal is above the noise level, we significantly reduce the number of unknowns to be determined. This "compact" MFBD yields high-quality restorations in a much shorter time than is achieved with MFBD algorithms that do not model the temporal signatures; it may also provide higher-fidelity solutions. PMID:21403711

  9. Compact Plasma Accelerator

    NASA Technical Reports Server (NTRS)

    Foster, John E.

    2004-01-01

    A plasma accelerator has been conceived for both material-processing and spacecraft-propulsion applications. This accelerator generates and accelerates ions within a very small volume. Because of its compactness, this accelerator could be nearly ideal for primary or station-keeping propulsion for spacecraft having masses between 1 and 20 kg. Because this accelerator is designed to generate beams of ions having energies between 50 and 200 eV, it could also be used for surface modification or activation of thin films.

  10. Compact laser amplifier system

    DOEpatents

    Carr, R.B.

    1974-02-26

    A compact laser amplifier system is described in which a plurality of face-pumped annular disks, aligned along a common axis, independently radially amplify a stimulating light pulse. Partially reflective or lasing means, coaxially positioned at the center of each annualar disk, radially deflects a stimulating light directed down the common axis uniformly into each disk for amplification, such that the light is amplified by the disks in a parallel manner. Circumferential reflecting means coaxially disposed around each disk directs amplified light emission, either toward a common point or in a common direction. (Official Gazette)

  11. Compact gate valve

    DOEpatents

    Bobo, Gerald E.

    1977-01-01

    This invention relates to a double-disc gate valve which is compact, comparatively simple to construct, and capable of maintaining high closing pressures on the valve discs with low frictional forces. The valve casing includes axially aligned ports. Mounted in the casing is a sealed chamber which is pivotable transversely of the axis of the ports. The chamber contains the levers for moving the valve discs axially, and an actuator for the levers. When an external drive means pivots the chamber to a position where the discs are between the ports and axially aligned therewith, the actuator for the levers is energized to move the discs into sealing engagement with the ports.

  12. Experimental test of the system of vertical and longitudinal lithium limiters on T-11M tokamak as a prototype of plasma facing components of a steady-state fusion neutron source

    NASA Astrophysics Data System (ADS)

    Mirnov, S. V.; Belov, A. M.; Djigailo, N. T.; Dzhurik, A. S.; Kravchuk, S. I.; Lazarev, V. B.; Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu.; Shcherbak, A. N.

    2015-11-01

    A new functional model of the prototype of closed Li circuit for protection of the chamber wall was tested in T-11M tokamak by simultaneous use of the vertical Li limiter as an emitter of Li and a new longitudinal Li limiter as its collector. Such technological scheme can be suggested for the steady-state fusion neutron source on the tokamak basis. During plasma shots the cryogenic target of T-11M collected Li flow emitted by the vertical capillary Li limiter almost completely (up to 80%). These Li and hydrogen isotopes were captured and extracted outside the tokamak vacuum chamber without venting of the vessel which is a key requirement for the use of Li in the steady-state tokamak reactor.

  13. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  14. Experimental studies of compact toroids. Progress report, 1990--1991

    SciTech Connect

    Not Available

    1991-12-01

    The Berkeley Compact Toroid Experiment (BCTX) device is a plasma device with a Marshall-gun generated, low aspect ratio toroidal plasma. The device is capable of producing spheromak-type discharges and may, with some modification, produce low-aspect ratio tokamak configurations. A unique aspect of this experimenal devie is its large lower hybrid (LH) heating system, which consists of two 450MHz klystron tubes generating 20 megawatts each into a brambilla-type launching structure. Successful operation with one klystron at virtually full power (18 MW) has been accomplished with 110 {mu}s pulse length. A second klystron is currently installed in its socket and magnet but has not been added to the RF drive system. This report describes current activities and accomplishments and describes the anticipated results of next year`s activity.

  15. Materials issues in fusion reactors

    NASA Astrophysics Data System (ADS)

    Suri, A. K.; Krishnamurthy, N.; Batra, I. S.

    2010-02-01

    The world scientific community is presently engaged in one of the toughest technological tasks of the current century, namely, exploitation of nuclear fusion in a controlled manner for the benefit of mankind. Scientific feasibility of controlled fusion of the light elements in plasma under magnetic confinement has already been proven. International efforts in a coordinated and co-operative manner are presently being made to build ITER - the International Thermonuclear Experimental Reactor - to test, in this first step, the concept of 'Tokamak' for net fusion energy production. To exploit this new developing option of making energy available through the route of fusion, India too embarked on a robust fusion programme under which we now have a working tokamak - the Aditya and a steady state tokamak (SST-1), which is on the verge of functioning. The programme envisages further development in terms of making SST-2 followed by a DEMO and finally the fusion power reactor. Further, with the participation of India in the ITER program in 2005, and recent allocation of half - a - port in ITER for placing our Lead - Lithium Ceramic Breeder (LLCB) based Test Blanket Module (TBM), meant basically for breeding tritium and extracting high grade heat, the need to understand and address issues related to materials for these complex systems has become all the more necessary. Also, it is obvious that with increasing power from the SST stages to DEMO and further to PROTOTYPE, the increasing demands on performance of materials would necessitate discovery and development of new materials. Because of the 14.1 MeV neutrons that are generated in the D+T reaction exploited in a tokamak, the materials, especially those employed for the construction of the first wall, the diverter and the blanket segments, suffer crippling damage due to the high He/dpa ratios that result due to the high energy of the neutrons. To meet this challenge, the materials that need to be developed for the tokamaks

  16. The negative triangularity tokamak: stability limits and prospects as a fusion energy system

    NASA Astrophysics Data System (ADS)

    Medvedev, S. Yu.; Kikuchi, M.; Villard, L.; Takizuka, T.; Diamond, P.; Zushi, H.; Nagasaki, K.; Duan, X.; Wu, Y.; Ivanov, A. A.; Martynov, A. A.; Poshekhonov, Yu. Yu.; Fasoli, A.; Sauter, O.

    2015-06-01

    The paper discusses edge stability, beta limits and power handling issues for negative triangularity tokamaks. The edge magnetohydrodynamic stability is the most crucial item for power handling. For the case of negative triangularity the edge stability picture is quite different from that for conventional positive triangularity tokamaks: the second stability access is closed for localized Mercier/ballooning modes due to the absence of a magnetic well, and nearly internal kink modes set the pedestal height limit to be weakly sensitive to diamagnetic stabilization just above the margin of the localized mode Mercier criterion violation. While a negative triangularity tokamak is thought to have a low beta limit with its magnetic hill property, it is found that plasmas with reactor-relevant values of normalized beta βN > 3 can be stable to global kink modes without wall stabilization with appropriate core pressure profile optimization against localized mode stability, and also with increased magnetic shear in the outer half-radius. The beta limit is set by the n = 1 mode for the resulting flat pressure profile. The wall stabilization is very inefficient due to strong coupling between external and internal modes. The n > 1 modes are increasingly internal when approaching the localized mode limit, and set a lower beta in the case of the peaked pressure profile leading to a Mercier unstable core. With the theoretical predictions supported by experiments, a negative triangularity tokamak would become a prospective fusion energy system with other advantages including a larger separatrix wetted area, more flexible divertor configuration design, wider trapped particle-free scrape-off layer, lower background magnetic field for internal poloidal field coils, and larger pumping conductance from the divertor room.

  17. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor.

    PubMed

    Leipold, F; Furtula, V; Salewski, M; Bindslev, H; Korsholm, S B; Meo, F; Michelsen, P K; Moseev, D; Nielsen, S K; Stejner, M

    2009-09-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic using gyrotrons operated at 60 GHz will meet the requirements for spatially and temporally resolved measurements of the velocity distributions of confined fast alphas in ITER by evaluating the scattered radiation (CTS signal). While a receiver antenna on the low field side of the tokamak, resolving near perpendicular (to the magnetic field) velocity components, has been enabled, an additional antenna on the high field side (HFS) would enable measurements of near parallel (to the magnetic field) velocity components. A compact design solution for the proposed mirror system on the HFS is presented. The HFS CTS antenna is located behind the blankets and views the plasma through the gap between two blanket modules. The viewing gap has been modified to dimensions 30x500 mm(2) to optimize the CTS signal. A 1:1 mock-up of the HFS mirror system was built. Measurements of the beam characteristics for millimeter-waves at 60 GHz used in the mock-up agree well with the modeling. PMID:19791936

  18. Compaction of Titanium Powders

    SciTech Connect

    Stephen J. Gerdemann; Paul D. Jablonski

    2010-11-01

    Accurate modeling of powder densification has been an area of active research for more than 60 years. The earliest efforts were focused on linearization of the data because computers were not readily available to assist with curve-fitting methods. In this work, eight different titanium powders (three different sizes of sponge fines <150 μm, <75 μm, and < 45 μm; two different sizes of a hydride-dehydride [HDH] <75 μm and < 45 μm; an atomized powder; a commercially pure [CP] Ti powder from International Titanium Powder [ITP]; and a Ti 6 4 alloy powder) were cold pressed in a single-acting die instrumented to collect stress and deformation data during compaction. From these data, the density of each compact was calculated and then plotted as a function of pressure. The results show that densification of all the powders, regardless of particle size, shape, or chemistry, can be modeled accurately as the sum of an initial density plus the sum of a rearrangement term and a work-hardening term. These last two terms are found to be a function of applied pressure and take the form of an exponential rise.

  19. Compact electrostatic comb actuator

    DOEpatents

    Rodgers, M. Steven; Burg, Michael S.; Jensen, Brian D.; Miller, Samuel L.; Barnes, Stephen M.

    2000-01-01

    A compact electrostatic comb actuator is disclosed for microelectromechanical (MEM) applications. The actuator is based upon a plurality of meshed electrostatic combs, some of which are stationary and others of which are moveable. One or more restoring springs are fabricated within an outline of the electrostatic combs (i.e. superposed with the moveable electrostatic combs) to considerably reduce the space required for the actuator. Additionally, a truss structure is provided to support the moveable electrostatic combs and prevent bending or distortion of these combs due to unbalanced electrostatic forces or external loading. The truss structure formed about the moveable electrostatic combs allows the spacing between the interdigitated fingers of the combs to be reduced to about one micron or less, thereby substantially increasing the number of active fingers which can be provided in a given area. Finally, electrostatic shields can be used in the actuator to substantially reduce unwanted electrostatic fields to further improve performance of the device. As a result, the compact electrostatic comb actuator of the present invention occupies only a fraction of the space required for conventional electrostatic comb actuators, while providing a substantial increase in the available drive force (up to one-hundred times).

  20. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.