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Sample records for tokamak fusion core

  1. Tokamak Fusion Core Experiment maintenance study

    International Nuclear Information System (INIS)

    Snyder, A.M.; Watts, K.D.

    1985-01-01

    The recently completed Tokamak Fusion Core Experiment (TFCX) design project was carried out to investigate potential next generation tokamak concepts. An important aspect of this project was the early development and incorporation of remote maintainability throughout the design process. This early coordination and incorporation of maintenance aspects to the design of the device and facilities would assure that the machine could ultimately be maintained and repaired in an efficient and cost effective manner. To meet this end, a rigorously formatted engineering trade study was performed to determine the preferred configuration for the TFCX reactor based primarily on maintenance requirements. The study indicated that the preferred design was one with an external vacuum vessel and torrodial field coils that could be removed via a simple radial motion. The trade study is presented and the preferred TFCX configuration is described

  2. The Tokamak Fusion Core Experiment studies

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Tokamak Fusion Core Experiment (TFCX) special-purpose remote maintenance systems

    International Nuclear Information System (INIS)

    Masson, L.S.; Welland, H.J.

    1985-01-01

    A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX

  4. Parametric system studies of candidate TF coil system options for the Tokamak Fusion Core Experiment (TFCX)

    International Nuclear Information System (INIS)

    Reiersen, W.T.; Flanagan, C.A.; Miller, J.B.

    1983-01-01

    System studies were performed to determine the sensitivity of hybrid and superconducting toroidal field (TF) coil system options to maximum field at the TF coil and to field enhancement due to resistive insert coils. The studies were performed using Tokamak Fusion Core Experiment (TFCX) design assumptions, guidelines, and criteria and involved iterative execution of the Fusion Engineering Design Center (FEDC) systems code, magnetohydrodynamics (MHD) equilibrium code, and EFFI (a code to evaluate magnetic field strength). The results indicate that for TFCX with no minimum wall loading specified, a design point chosen solely on the basis of cost would likely be in the low-field region of design space where the cost advantage of hybrids is least apparent. However, as the desired neutron wall loading increases, the hybrid option suggests an increasing cost advantage over the all-superconducting option; this cost advantage is countered by increased complexity in design -- particularly in assembly and maintenance

  5. Parametric system studies of candidate TF coil system options for the Tokamak Fusion Core Experiment (TFCX)

    International Nuclear Information System (INIS)

    Reiersen, W.T.; Flanagan, C.A.; Miller, J.B.

    1983-01-01

    System studies were performed to determine the sensitivity of hybrid and superconducting toroidal field (TF) coil system options to maximum field at the TF coil and to field enhancement due to resistive insert coils. The studies were performed using Tokamak Fusion Core Experiment (TFCX) design assumptions, guidelines, and criteria and involved iterative execution of the Fusion Engineering Design Center (FEDC) systems code, magnetohydrodynamics (MHD) equilibrium code, and EFFI (a code to evaluate magnetic field strength). The results indicate that for TFCX with no minimum wall loading specified, a design point chosen solely on the basis of cost would likely be in the low-field region of design space where the cost advantage of hybrids is least apparent. However, as the desired neutron wall loading increases, the hybrid option suggests an increasing cost advantage over the all-superconducting option; this cost advantage is countered by increased complexity in design - particularly in assembly and maintenance

  6. Design assumptions and bases for small D-T-fueled Sperical Tokamak (ST) fusion core

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Galambos, J.D.; Fogarty, P.J.

    1996-01-01

    Recent progress in defining the assumptions and clarifying the bases for a small D-T-fueled ST fusion core are presented. The paper covers several issues in the physics of ST plasmas, the technology of neutral beam injection, the engineering design configuration, and the center leg material under intense neutron irradiation. This progress was driven by the exciting data from pioneering ST experiments, a heightened interest in proof-of-principle experiments at the MA level in plasma current, and the initiation of the first conceptual design study of the small ST fusion core. The needs recently identified for a restructured fusion energy sciences program have provided a timely impetus for examining the subject of this paper. Our results, though preliminary in nature, strengthen the case for the potential realism and attractiveness of the ST approach

  7. Impact of an integrated core/SOL description on the R and B T optimization of tokamak fusion reactors

    Science.gov (United States)

    Siccinio, M.; Fable, E.; Angioni, C.; Saarelma, S.; Scarabosio, A.; Zohm, H.

    2018-01-01

    An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion 58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion 55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field B T—for different safety factors q, allowable loop voltages V loop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion 53 073019 and Wenninger et al 2017 Nucl. Fusion 57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and B T once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L-H transition) is present.

  8. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

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

  9. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Robinson, D.C.; Akers, R.; Allfrey, S.J.

    1999-01-01

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  10. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Robinson, D.C.; Akers, R.; Allfrey, S.J.

    2001-01-01

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  11. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  12. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Hively, L.M.; Sigmar, D.J.

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  13. Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Nohara, Kiyohiko

    2009-01-01

    The structural material is one of key issues for the development of reliable superconducting magnets and peripheral equipments of fusion reactors. Standard stainless steels like SUS 304 and 316 steels available at present do not meet requirements. We are developing a new austenitic steel that has proposed target properties named 'JAERI BOX'. Additions of N and V at different amounts were tested to improve strength and fracture toughness of a base alloy SUS316LN at 4.2 K. Mechanical properties of the developed steel were examined. It is found that the charpy absorbed energy and the fracture toughness of the developed steel at 4.2 K are within JAERI BOX. (T.I.)

  14. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

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

  15. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

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

  16. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  17. The ARIES tokamak fusion reactor study

    International Nuclear Information System (INIS)

    Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

    1989-01-01

    The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features 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 in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

  18. Realizing steady-state tokamak operation for fusion energy

    International Nuclear Information System (INIS)

    Luce, T. C.

    2011-01-01

    Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

  19. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

    Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

    2017-01-01

    Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

  20. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  1. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1998-01-01

    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

  2. The spherical tokamak fusion power plant

    International Nuclear Information System (INIS)

    Wilson, H.R.; Voss, G.; Ahn, J.W.

    2003-01-01

    The design of a 1GW(e) steady state fusion power plant, based on the spherical tokamak concept, has been further iterated towards a fully self-consistent solution taking account of plasma physics, engineering and neutronics constraints. In particular a plausible solution to exhaust handling is proposed and the steam cycle refined to further improve efficiency. The physics design takes full account of confinement, MHD stability and steady state current drive. It is proposed that such a design may offer a fusion power plant which is easy to maintain: an attractive feature for the power plants following ITER. (author)

  3. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  4. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

  5. Minerals resource implications of a tokamak fusion reactor economy

    Energy Technology Data Exchange (ETDEWEB)

    Cameron, E; Conn, R W; Kulcinski, G L; Sviatoslavsky, I

    1979-09-01

    The mineral resource implications of an economy of tokamak-type fusion reactors are assessed based upon the recent conceptual reactor design study, NUWMAK, developed at the University of Wisconsin. For comparative purposes, various structural alloys of vanadium and steel are assumed to be usable in the NUWMAK design in place of the titanium alloy originally selected. In addition, the inner blanket core and magnet system of the conceptual reactor, HFCTR, developed at the Massachusetts Institute of Technology, are assumed to be interchangeable with the comparable components in NUWMAK. These variations permit a range of likely requirements to be assessed.

  6. Minerals resource implications of a tokamak fusion reactor economy

    International Nuclear Information System (INIS)

    Cameron, E.; Conn, R.W.; Kulcinski, G.L.; Sviatoslavsky, I.

    1979-09-01

    The mineral resource implications of an economy of tokamak-type fusion reactors are assessed based upon the recent conceptual reactor design study, NUWMAK, developed at the University of Wisconsin. For comparative purposes, various structural alloys of vanadium and steel are assumed to be usable in the NUWMAK design in place of the titanium alloy originally selected. In addition, the inner blanket core and magnet system of the conceptual reactor, HFCTR, developed at the Massachusetts Institute of Technology, are assumed to be interchangeable with the comparable components in NUWMAK. These variations permit a range of likely requirements to be assessed

  7. Compact fusion energy based on the spherical tokamak

    Science.gov (United States)

    Sykes, A.; Costley, A. E.; Windsor, C. G.; Asunta, O.; Brittles, G.; Buxton, P.; Chuyanov, V.; Connor, J. W.; Gryaznevich, M. P.; Huang, B.; Hugill, J.; Kukushkin, A.; Kingham, D.; Langtry, A. V.; McNamara, S.; Morgan, J. G.; Noonan, P.; Ross, J. S. H.; Shevchenko, V.; Slade, R.; Smith, G.

    2018-01-01

    Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

  8. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    Mantsinen, M.J.

    1995-03-01

    The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

  9. Advantages of iron core in a tokamak

    International Nuclear Information System (INIS)

    Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

    1977-01-01

    A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

  10. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    2003-01-01

    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 and 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 and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and 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

  11. Tokamak fusion test reactor FELIX plate experiment

    International Nuclear Information System (INIS)

    Hua, T.O.; Nygren, R.E.; Turner, L.R.

    1986-01-01

    For a conducting material exposed to both a time-varying and a static magnetic field, such as a limiter blade in a tokamak, the induced eddy currents and the deflection arising from those eddy currents can be strongly coupled. The coupling effects reduce the currents and deflections markedly, sometimes an order of magnitude, from the values predicted if coupling is neglected. A series of experiments to study current-deflection coupling were performed using the Fusion Electromagnetic Inductance Experiment (FELIX) facility at Argonne National Laboratory. Magnetic damping and magnetic stiffness resulting from the coupling are discussed, and analytical expressions for induced eddy current and rigid body rotation in the FELIX plate experiment are compared with the experimental results. Predictions for the degree of coupling based on various parameters are made using the analytical model

  12. Commercial feasibility of fusion power based on the tokamak concept

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    The impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants is determined. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  13. Application of mineral insulated cable (MIC) in Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Luo Tianyong; Jiang Jiaming; Cen Yishun

    2014-01-01

    To avoid the instability of plasma and achieve some experimental tasks in Tokamak fusion reactor, many in-vessel coils are designed such as the coils to mitigate the effect of Edge Localized Modes (ELMs coils) and the coils to provide vertical stabilization (VS coils). The in-vessel location presents special challenges in terms of nuclear radiation and temperature, and requires the use of mineral-insulated conductors. The in-vessel coils in ITER are designed to be Mineral-insulated Cable (MIC) with three-layer structures. The inner is hollow-core tube made by OFHC or CuCrZr, the middle is the insulation layer made by Mgo and the outer is the jacket by SS316L or Inconel 718. To control the effect of Edge Localized Modes and vertical instability of plasma, the MIC in-vessel coils shall be used in HL-2M. More details about the application of MIC in Tokamak fusion reactor will be shown in this report. (authors)

  14. Application of internally cooled superconductors to tokamak fusion reactors

    International Nuclear Information System (INIS)

    Materna, P.A.

    1986-01-01

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

  15. Neutronic analysis of fusion tokamak devices by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

    2011-01-01

    A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

  16. Tokamak Fusion Test Reactor D-T results

    International Nuclear Information System (INIS)

    Meade, D.M.

    1995-01-01

    Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, α confinement, α heating and possible α-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20MW of tritium and 14MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2MW. The fusion power density in the core of the plasma was about 1.8MWm -3 , approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of τ E ∝A 0.6 . The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6keV can be attributed to electron heating by the α particles. The approximately 5% loss of α particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy α particles and the resultant α ash density. At fusion power levels of 7.5MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional α loss due to the fluctuations was observed. (orig.)

  17. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  18. On the economic prospects of nuclear fusion with tokamaks

    International Nuclear Information System (INIS)

    Pfirsch, D.; Schmitter, K.H.

    1987-12-01

    This paper describes a method of cost and construction energy estimation for tokamak fusion power stations conforming to the present, early stage of fusion development. The method is based on first-wall heat load constraints rather than β limitations, which, however, might eventually be the more critical of the two. It is used to discuss the economic efficiency of pure fusion, with particular reference to the European study entitled 'Environmental Impact and Economic Prospects of Nuclear Fusion'. It is shown that the claims made therein for the economic prospects of pure fusion with tokamaks, when discussed on the basis of the present-day technology, do not stand up to critical examination. A fusion-fission hybrid, however, could afford more positive prospects. Support for the stated method is even derived when it is properly applied for cost estimation of advanced gascooled and Magnox reactors, the two very examples presented by the European study to 'disprove' it. (orig.)

  19. Fusion performance analysis of plasmas with reversed magnetic shear in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Ruskov, E.; Bell, M.; Budny, R.V.; McCune, D.C.; Medley, S.S.; Nazikian, R.; Synakowski, E.J.; Goeler, S. von; White, R.B.; Zweben, S.J.

    1999-01-01

    A case for substantial loss of fast ions degrading the performance of tokamak fusion test reactor plasmas [Phys. Plasmas 2, 2176 (1995)] with reversed magnetic shear (RS) is presented. The principal evidence is obtained from an experiment with short (40 - 70 ms) tritium beam pulses injected into deuterium beam heated RS plasmas [Phys. Rev. Lett. 82, 924 (1999)]. Modeling of this experiment indicates that up to 40% beam power is lost on a time scale much shorter than the beam - ion slowing down time. Critical parameters which connect modeling and experiment are: The total 14 MeV neutron emission, its radial profile, and the transverse stored energy. The fusion performance of some plasmas with internal transport barriers is further deteriorated by impurity accumulation in the plasma core. copyright 1999 American Institute of Physics

  20. A need for non-tokamak approaches to magnetic fusion energy

    International Nuclear Information System (INIS)

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

    1992-01-01

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

  1. Fusion-product transport in axisymmetric tokamaks: losses and thermalization

    International Nuclear Information System (INIS)

    Hively, L.M.

    1980-01-01

    High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-β, non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated

  2. Technology issues for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

    The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community

  3. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

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

  4. Demountable toroidal fusion core facility for physics optimization and fusion engineering

    International Nuclear Information System (INIS)

    Bogart, S.L.; Wagner, C.E.; Krall, N.A.; Dalessandro, J.A.; Weggel, C.F.; Lund, K.O.; Sedehi, S.

    1986-01-01

    Following a successful compact ignition tokamak (CIT) experiment, a fusion facility will be required for physics optimization (POF) and fusion engineering research (FERF). The POF will address issues such as high-beta operation, current drive, impurity control, and will test geometric and configurational variations such as the spherical torus or the reversed-field pinch (RFP). The FERF will be designed to accumulate rapidly a large neutron dose in prototypical fusion subsystems exposed to radiation. Both facilities will require low-cost replacement cores and rapid replacement times. The Demountable Toroidal Fusion Core (DTFC) facility is designed to fulfill these requirements. It would be a cost-effective stepping stone between the CIT and a demonstration fusion reactor

  5. Remote operation of the GOLEM tokamak for Fusion Education

    Czech Academy of Sciences Publication Activity Database

    Grover, O.; Kocman, J.; Odstrčil, M.; Odstrčil, T.; Matušů, M.; Stöckel, Jan; Svoboda, V.; Vondrášek, G.; Žára, J.

    2016-01-01

    Roč. 112, November (2016), s. 1038-1044 ISSN 0920-3796. [Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research IAEA /10./. Ahmedabad, 20.04.2015-24.04.2015] Institutional support: RVO:61389021 Keywords : Tokamak technology * Remote participation * Education * Nuclear fusion Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379616303441

  6. Tokamak Fusion Test Reactor neutral beam injection system vacuum chamber

    International Nuclear Information System (INIS)

    Pedrotti, L.R.

    1977-01-01

    Most of the components of the Neutral Beam Lines of the Tokamak Fusion Test Reactor (TFTR) will be enclosed in a 50 cubic meter box-shaped vacuum chamber. The chamber will have a number of unorthodox features to accomodate both neutral beam and TFTR requirements. The design constraints, and the resulting chamber design, are presented

  7. Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Goeler, S. von; Houlberg, W.A.

    2001-01-01

    Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  8. Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Von Goeler, S.; Houlberg, W.A.

    1999-01-01

    Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

  9. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    1978-08-01

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  10. Development of 3D ferromagnetic model of tokamak core withstrong toroidal asymmetry

    Czech Academy of Sciences Publication Activity Database

    Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Pánek, Radomír

    96-97, October (2015), s. 302-305 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * ferromagnetic core * model of ferromagnet * integral method * tokamak GOLEM Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615002100

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  13. Transients and burn dynamics in advanced tokamak fusion reactors

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  14. Evaluation of applicability of 2D iron core model for two-limb configuration of GOLEM tokamak

    Czech Academy of Sciences Publication Activity Database

    Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Vondrášek, G.

    2013-01-01

    Roč. 88, 6-8 (2013), s. 835-838 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Tokamak * Ferromagnetic core * Integral method * Tokamak GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613002573#

  15. In-vessel maintenance concepts for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Yount, J.A.

    1983-01-01

    Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

  16. Applicability of the PHITS code to a tokamak fusion device

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko; Okuno, Koichi; Kawasaki, Hiromitsu

    2011-01-01

    The three-dimensional Monte-Carlo code PHITS (particle and Heavy Ion Transport code System) has been developed to perform the radiation transport analysis, design of the radiation shields and neutronics calculations for tokamak-type D-D fusion reactors. A subroutine was included in PHITS to represent the toroidal neutron source of 2.45 MeV neutrons from the D-D reaction. Here, an example of preliminary tests using PHITS is given. (author)

  17. Net energy balance of tokamak fusion power plants

    International Nuclear Information System (INIS)

    Buende, R.

    1981-10-01

    The net energy balance for a tokamak fusion power plant was determined by using a PWR power plant as reference system, replacing the fission-specific components by fusion-specific components and adjusting the non-reactor-specific components to altered conditions. For determining the energy input to the fusion plant a method was developed that combines the advantages of the energetic input-output method with those of process chain analysis. A comparison with PWR, HTR, FBR, and coal-fired power plants is made. As a result the net energy balance of the fusion power plant turns out to be more advantageous than that of an LWR, HTR or coal-fired power plant and nearly in the same range as FBR power plants. (orig.)

  18. Net energy balance of tokamak fusion power plants

    International Nuclear Information System (INIS)

    Buende, R.

    1983-01-01

    The net energy balance for a tokamak fusion power plant of present day design is determined by using a PWR power plant as reference system, replacing the fission-specific components by fusion-specific components and adjusting the non-reactor-specific components to altered conditions. For determining the energy input to the fusion plant a method was developed that combines the advantages of the energetic input-output method with those of process chain analysis. A comparison with PWR, HTR, FBR, and coal-fired power plants is made. As a result the energy expenditures of the fusion power plant turn out to be lower than that of an LWR, HTR, or coal-fired power plant of equal net electric power output and nearly in the same range as FBR power plants. (orig.)

  19. Development of large insulator rings for the TOKAMAK Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1977-01-01

    Research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applictions, fabrication approach and testing activities are highlighted

  20. Simulation of fusion power in tokamak reactor

    International Nuclear Information System (INIS)

    Gaber, F.A.; Elsharif, R.N.; Sayed, Y.A.

    1993-01-01

    The paper deals with the transient response of the fusion power against perturbation in the injection rate of the fuel to ± 10% step change. The steady state results are in good agreement with the references results. The adequacy of these study was tested by assessing the physical plausibility of the obtained result, as well as, comparison with other validated model. 2 fig., 2 tab

  1. Thermonuclear Tokamak plasmas in the presence of fusion alpha particles

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1988-01-01

    In this overview, we have focused on several results of the thermonuclear plasma research pertaining to the alpha particle physics and diagnostics in a fusion tokamak plasma. As regards the discussion of alpha particle effects, two distinct classes of phenomena have been distinguished: the simpler class containing phenomena exhibited by individual alpha particles under the influence of bulk plasma properties and, the more complex class including collective effects which become important for increasing alpha particle density. We have also discussed several possibilities to investigate alpha particle effects by simulation experiments using an equivalent population of highly energetic ions in the plasma. Generally, we find that the present theoretical knowledge on the role of fusion alpha particles in a fusion tokamak plasma is incomplete. There are still uncertainties and partial lack of quantitative results in this area. Consequently, further theoretical work and, as far a possible, simulation experiments are needed to improve the situation. Concerning the alpha particle diagnostics, the various diagnostic techniques and the status of their development have been discussed in two different contexts: the escaping alpha particles and the confined alpha particles in the fusion plasma. A general conclusion is that many of the different diagnostic methods for alpha particle measurements require further major development. (authors)

  2. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1999-01-01

    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

  3. Steady-state operation requirements of tokamak fusion reactor concepts

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1991-06-01

    In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

  4. Analysis of Confinement Strategies for a Tokamak Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

    2001-01-15

    The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established.

  5. Analysis of Confinement Strategies for a Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

    2001-01-01

    The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established

  6. Remote operation of the GOLEM tokamak for Fusion Education

    Energy Technology Data Exchange (ETDEWEB)

    Grover, O.; Kocman, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Odstrcil, M. [University of Southampton, Southampton SO17 1BJ (United Kingdom); Odstrcil, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Matusu, M. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, Prague CZ-182 21 (Czech Republic); Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Vondrasek, G. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Zara, J. [Faculty of Electrical Engineering CTU Prague, CZ-166 27 (Czech Republic)

    2016-11-15

    Highlights: • The remote operation of the tokamak GOLEM for educational purposes. - Abstract: Practically oriented education in the field of thermonuclear fusion is highly requested. However, the high complexity of appropriate experiments makes it difficult to develop and maintain laboratories where students can take part in hands-on experiments in this field of study. One possible solution is to establish centres with specific high temperature plasma experiments where students can visit such a laboratory and perform their experiments in-situ. With the advancements of IT technologies it naturally follows to make a step forward and connect these with necessary plasma physics technologies and thus allow to access even sophisticated experiments remotely. Tokamak GOLEM is a small, modest device with its infrastructure linked to web technologies allowing students to set-up necessary discharge parameters, submit them into a queue and within minutes obtain the results in the form of a discharge homepage.

  7. Remote operation of the GOLEM tokamak for Fusion Education

    International Nuclear Information System (INIS)

    Grover, O.; Kocman, J.; Odstrcil, M.; Odstrcil, T.; Matusu, M.; Stöckel, J.; Svoboda, V.; Vondrasek, G.; Zara, J.

    2016-01-01

    Highlights: • The remote operation of the tokamak GOLEM for educational purposes. - Abstract: Practically oriented education in the field of thermonuclear fusion is highly requested. However, the high complexity of appropriate experiments makes it difficult to develop and maintain laboratories where students can take part in hands-on experiments in this field of study. One possible solution is to establish centres with specific high temperature plasma experiments where students can visit such a laboratory and perform their experiments in-situ. With the advancements of IT technologies it naturally follows to make a step forward and connect these with necessary plasma physics technologies and thus allow to access even sophisticated experiments remotely. Tokamak GOLEM is a small, modest device with its infrastructure linked to web technologies allowing students to set-up necessary discharge parameters, submit them into a queue and within minutes obtain the results in the form of a discharge homepage.

  8. A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

    Directory of Open Access Journals (Sweden)

    Mitsuru Kikuchi

    2010-11-01

    Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

  9. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-01

    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 DandD (Decontamination and Decommissioning) activity

  10. An investigation on detection and measurement of fusion neutron spectrum and radiation flux in large tokamak

    International Nuclear Information System (INIS)

    Yang Jinwei; Li Wenzhong; Zhang Wei

    2003-01-01

    The detection methods, detectors and spectrometers of D-D and D-T fusion neutron have been overviewed in large tokamaks. Some options are proposed for developing new detection systems of fusion neutrons suitable to the HL-2A tokamak. (authors)

  11. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  12. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

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

    International Nuclear Information System (INIS)

    Deng Meigen; Feng Kaiming; Yang Bangchao

    2002-01-01

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

  14. Plan for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D ampersand D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D ampersand D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D ampersand D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D ampersand D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates

  15. Feasibility study of a fission-suppressed tokamak fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Neef, W.S.

    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 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 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 3 O 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. Core fuelling to produce peaked density profiles in large tokamaks

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.

    1995-01-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. It is shown that with radially 'hollow' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles that are peaked off-axis. The fuelling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fuelling does not require Megavolt particle energies. Even with beam voltages of ∼ 200 keV, however, exceptionally good particle confinement is needed to achieve net electrical power generation. The required ratio of particle to thermal diffusivities is an order of magnitude outside the range reported for tokamaks. In a system with no power production requirement (e.g., neutron sources) neutral beam fuelling should be capable of producing peaked density profiles in devices as large as ITER. Fuelling systems with low energy cost per particle - such as cryogenic pellet injection - must be used in power producing tokamaks when τ P ∼ τ E . Simulations with pellet injection speeds of 7 km/s show that the peaking factor, n e0 / e >, approaches 2. (author). 65 refs, 8 figs

  17. Fusion reactivity, confinement, and stability of neutral-beam heated plasmas in TFTR and other tokamaks

    International Nuclear Information System (INIS)

    Park, Hyeon, K.

    1996-05-01

    The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  18. [Fusion research/tokamak]. Final report, 1 May 1988 - 30 April 1994

    International Nuclear Information System (INIS)

    1994-01-01

    The objectives of the Fusion Research Center Program are: (1) to advance /the transport studies of tokamaks, including the development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for the text-upgrade tokamak. Work is described on five basic categories: (1) magnetic fusion energy database; (2) computational support and numerical modeling; (3) support for TEXT-upgrade and diagnostics; (4) transport studies; and (5) Alfven waves

  19. Physics of the Tokamak Pedestal, and Implications for Magnetic Fusion Energy

    Science.gov (United States)

    Snyder, Philip

    2017-10-01

    High performance in tokamaks is achieved via the spontaneous formation of a transport barrier in the outer few percent of the confined plasma. This narrow insulating layer, referred to as a ``pedestal,'' typically results in a >30x increase in pressure across a 0.4-5cm layer. Predicted fusion power scales with the square of the pedestal top pressure (or ``pedestal height''), hence a fusion reactor strongly benefits from a high pedestal, provided this can be attained without large Edge Localized Modes (ELMs), which may erode plasma facing materials. The overlap of drift orbit, turbulence, and equilibrium scales across this narrow layer leads to rich and complex physics, and challenges traditional analytic and computational approaches. We review studies employing gyrokinetic, neoclassical, MHD, and other methods, which have explored how a range of instabilities, influenced by complex geometry, and strong ExB flows and bootstrap current, drive transport across the pedestal and guide its structure and dynamics. Development of high resolution diagnostics, and coordinated experiments on several tokamaks, have validated understanding of important aspects of the physics, while highlighting open issues. A predictive model (EPED) has proven capable of predicting the pedestal height and width to 20-25% accuracy in large statistical studies. This model was used to predict a new, high pedestal ``Super H-Mode'' regime, which was subsequently discovered on DIII-D, and motivated experiments on Alcator C-Mod which achieved world record, reactor relevant pedestal pressure. We review open issues including improved formalism, particle and momentum transport, the role of neutrals and impurities, ELM control, and pedestal formation. Finally we discuss coupling pedestal and core predictive models to enable more comprehensive optimization of the tokamak fusion concept. Supported by the US DOE under DE-FG02-95ER54309, FC02-06ER54873, DE-FC02-04ER54698, DE-FC02-99ER54512.

  20. Parasitic momentum flux in the tokamak core

    Science.gov (United States)

    Stoltzfus-Dueck, T.

    2017-10-01

    Tokamak plasmas rotate spontaneously without applied torque. This intrinsic rotation is important for future low-torque devices such as ITER, since rotation stabilizes certain instabilities. In the mid-radius `gradient region,' which reaches from the sawtooth inversion radius out to the pedestal top, intrinsic rotation profiles may be either flat or hollow, and can transition suddenly between these two states, an unexplained phenomenon referred to as rotation reversal. Theoretical efforts to explain the mid-radius rotation shear have largely focused on quasilinear models, in which the phase relationships of some selected instability result in a nondiffusive momentum flux (``residual stress''). In contrast, the present work demonstrates the existence of a robust, fully nonlinear symmetry-breaking momentum flux that follows from the free-energy flow in phase space and does not depend on any assumed linear eigenmode structure. The physical origin is an often-neglected portion of the radial ExB drift, which is shown to drive a symmetry-breaking outward flux of co-current momentum whenever free energy is transferred from the electrostatic potential to ion parallel flows. The fully nonlinear derivation relies only on conservation properties and symmetry, thus retaining the important contribution of damped modes. The resulting rotation peaking is counter-current and scales as temperature over plasma current. As first demonstrated by Landau, this free-energy transfer (thus also the corresponding residual stress) becomes inactive when frequencies are much higher than the ion transit frequency, which allows sudden transitions between hollow and flat profiles. Simple estimates suggest that this mechanism may be consistent with experimental observations. This work was funded in part by the Max-Planck/Princeton Center for Plasma Physics and in part by the U.S. Dept. of Energy, Office of Science, Contract No. DE-AC02-09CH11466.

  1. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  2. New tritium monitor for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Jalbert, R.A.

    1985-01-01

    At DT-fueled fusion reactors, there will be a need for tritium monitors that can simultaneously measure in real time the concentrations of HTO, HT and the activated air produced by fusion neutrons. Such a monitor has been developed, tested and delivered to the Princeton Plasma Physics Laboratory for use at the Tokamak Fusion Test Reactor (TFTR). It uses semipermeable membranes to achieve the removal of HTO from the sampled air for monitoring and a catalyst to convert the HT to HTO, also for removal and monitoring. The remaining air, devoid of tritium, is routed to a third detector for monitoring the activated air. The sensitivities are those that would be expected from tritium instruments employing conventional flow-through ionization chambers: 1 to 3 μCi/m 3 . Its discriminating ability is approximately 10 -3 for any of the three components (HTO, HT and activated air) in any of the other two channels. For instance, the concentration of HT in the HTO channel is 10 -3 times its original concentration in the sampled air. This will meet the needs of TFTR

  3. Decontamination and decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

    1994-01-01

    The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

  4. Parametric requirements for noncircular tokamak commercial fusion plants

    International Nuclear Information System (INIS)

    Bourque, R.F.

    1978-05-01

    Systems analyses have been performed in order to determine the parameter ranges for economical operation of equilibrium commercial fusion power plants using noncircular tokamak reactors. The fully-integrated systems code SCOPE was employed in which costs of electricity were used as the critical dependent variable by which the effects of changes in system parameters were observed. The results of these studies show that many of the operating parameters required for economical operation can be considerably relaxed from what has been thought necessary. For example: (1) Neutron wall loadings over about 2 megawatts per square meter are unnecessary for economical tokamak operation; (2) electrical power levels beyond 1000 MW(e) are not required. In fact, power levels as low as 100 MW(e) may be acceptable; (3) total beta values of 10 to 12 percent are adequate for economic operation; (4) duty cycles as low as 50 percent have only a small effect on plant economics; (5) an acceptable first wall lifetime is 15 MW-yr per square meter

  5. Tokamak fusion reactors with less than full tritium breeding

    International Nuclear Information System (INIS)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

  6. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  8. Tritium pellet injector design for tokamak fusion test reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  9. STARFIRE: a commercial tokamak fusion power plant study

    International Nuclear Information System (INIS)

    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

  10. Maintenance features of the Compact Ignition Tokamak fusion reactor

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  11. Neutral-beam-injected tokamak fusion reactors: a review

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1976-08-01

    The theories of energetic-ion velocity distributions, stability, injection, and orbits were summarized. The many-faceted role of the energetic ions in plasma heating, fueling, and current maintenance, as well as in the direct enhancement of fusion power multiplication and power density, is discussed in detail for three reactor types. The relevant implications of recent experimental results on several beam-injected tokamaks are examined. The behavior of energetic ions is found to be in accordance with classical theory, large total ion energy densities are readily achieved, and plasma equilibrium and stability are maintained. The status of neutral-beam injectors and of conceptual design studies of beam-driven reactors are briefly reviewed. The principal plasma-engineering problems are those associated directly with achieving quasi-stationary operation

  12. Core fueling to produce peaked density profiles in large tokamaks

    International Nuclear Information System (INIS)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.

    1994-06-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. We show that with radially ''hollow'' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles which are peaked off-axis. The fueling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fueling does not require MeV particle energy. Even with beam voltages of ∼200 keV, however, exceptionally good particle confinement, τ p much-gt τ E is required to achieve net electrical power generation. In system with no power production requirement (e.g., neutron sources) neutral beam fueling should be capable of producing peaked density profiles in devices as large as ITER. Fueling systems with low energy cost per particle (such as cryogenic pellet injection) must be used in power producing tokamaks when τ p ∼ τ E . Simulations with pellet injection speeds of 7 km/sec show the peaking factor, n eo /left-angle n e right-angle, approaching 2

  13. The ARIES-AT advanced tokamak, Advanced technology fusion power plant

    International Nuclear Information System (INIS)

    Najmabadi, Farrokh; Abdou, A.; Bromberg, L.

    2006-01-01

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N . ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

  14. Helical-type device and laser fusion. Rivals for tokamak-type device at n-fusion development in Japan

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Under the current policy on the research and development of nuclear fusion in Japan, as enunciated by the Atomic Energy Commission of Japan, the type of a prototype fusion reactor will be chosen after 2020 from tokamak, helical or some other type including the inertial confinement fusion using lasers. A prototype fusion reactor is the next step following the tokamak type International Thermonuclear Experimental Reactor (ITER). With the prototype reactor, the feasibility as a power plant will be examined. At present the main research and development of nuclear fusion in Japan are on tokamak type, which have been promoted by Japan Atomic Energy Research Institute (JAERI). As for the other types of nuclear fusion, researches have been carried out on the helical type in Kyoto University and National Institute for Fusion Science (NIFS), the mirror type in Tsukuba University, the tokamak type using superconductive coils in Kyushu University, and the laser fusion in Osaka University. The features and the present state of research and development of the Large Helical Device and the laser fusion which is one step away from the break-even condition are reported. (K.I.)

  15. Energy and particle core transport in tokamaks and stellarators compared

    Energy Technology Data Exchange (ETDEWEB)

    Beurskens, Marc; Angioni, Clemente; Beidler, Craig; Dinklage, Andreas; Fuchert, Golo; Hirsch, Matthias; Puetterich, Thomas; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald/Garching (Germany)

    2016-07-01

    The paper discusses expectations for core transport in the Wendelstein 7-X stellarator (W7-X) and presents a comparison to tokamaks. In tokamaks, the neoclassical trapped-particle-driven losses are small and turbulence dominates the energy and particle transport. At reactor relevant low collisionality, the heat transport is limited by ion temperature gradient limited turbulence, clamping the temperature gradient. The particle transport is set by an anomalous inward pinch, yielding peaked profiles. A strong edge pedestal adds to the good confinement properties. In traditional stellarators the 3D geometry cause increased trapped orbit losses. At reactor relevant low collisionality and high temperatures, these neoclassical losses would be well above the turbulent transport losses. The W7-X design minimizes neoclassical losses and turbulent transport can become dominant. Moreover, the separation of regions of bad curvature and that of trapped particle orbits in W7-X may have favourable implications on the turbulent electron heat transport. The neoclassical particle thermodiffusion is outward. Without core particle sources the density profile is flat or even hollow. The presence of a turbulence driven inward anomalous particle pinch in W7-X (like in tokamaks) is an open topic of research.

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

    International Nuclear Information System (INIS)

    Gillen, T.J.

    1980-11-01

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

  17. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S K; Lee, K W; Hwang, C K; Hong, B G; Hong, G W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new.

  18. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1978-01-01

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

  19. Conceptual radiation shielding design of superconducting tokamak fusion device by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Kawasaki, Hiromitsu; Okuno, Koichi

    2010-01-01

    A complete 3D neutron and photon transport analysis by Monte Carlo transport code system PHITS (Particle and Heavy Ion Transport code System) have been performed for superconducting tokamak fusion device such as JT-60 Super Advanced (JT-60SA). It is possible to make use of PHITS in the port streaming analysis around the devices for the tokamak fusion device, the duct streaming analysis in the building where the device is installed, and the sky shine analysis for the site boundary. The neutron transport analysis by PHITS makes it clear that the shielding performance of the superconducting tokamak fusion device with the cryostat is improved by the graphical results. From the standpoint of the port streaming and the duct streaming, it is necessary to calculate by 3D Monte Carlo code such as PHITS for the neutronics analysis of superconducting tokamak fusion device. (author)

  20. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    International Nuclear Information System (INIS)

    Kim, S. K.; Lee, K. W.; Hwang, C. K.; Hong, B. G.; Hong, G. W.

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new

  1. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  2. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    Energy Technology Data Exchange (ETDEWEB)

    Fu, P. E-mail: fupeng@mail.ipp.ac.cn; Liu, Z.Z.; Zou, J.H

    2002-05-01

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    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

  4. Issues for the electric utilities posed by DT tokamak fusion powerplants

    International Nuclear Information System (INIS)

    Roth, J.R.

    1990-01-01

    The DT tokamak is the mainline approach to magnetic fusion energy in all industrialized countries with a major commitment to fusion research. It achieved this status largely through historical accident and not as the result of considered choice among alternatives. After twenty-five years of intensive tokamak research, it is appropriate to ask whether the path down which the tokamak concept is leading the fusion community is the way to an acceptable powerplant for the electric utilities, or an aberration which should be replaced with an approach more promising in the long term. Issues surrounding the DT tokamak can be grouped in three broad areas: physics; safety/environmental; and engineering/economic. In addition to these problems, detailed engineering design studies of DT tokamak fusion powerplants over a twenty year period have revealed a number of additional problems. Most of thee are related to the presence of tritium and energetic neutron fluxes, which tend to make the cost of electricity of DT tokamaks higher than that of fossil or fission powerplants. These safety and economic issues of the DT tokamak powerplant also appear to be intractable, and have not been made to go away by twenty years of progressively more detailed and extensive engineering design studies

  5. TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

    Hatch, A.J.

    1975-01-01

    1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

  6. Forces on zonal flows in tokamak core turbulence

    International Nuclear Information System (INIS)

    Hallatschek, K.; Itoh, K.

    2005-01-01

    The saturation of stationary zonal flows (ZF) in the core of a tokamak has been analyzed in numerical fluid turbulence computer studies. The model was chosen to properly represent the kinetic global plasma flows, i.e., undamped stationary toroidal or poloidal flows and Landau damped geodesic acoustic modes. Reasonable agreement with kinetic simulations in terms of magnitude of transport and occurrence of the Dimits shift was verified. Contrary to common perception, in the final saturated state of turbulence and ZFs, the customary perpendicular Reynolds stress continues to drive the ZFs. The force balance is established by the essentially quasilinear parallel Reynolds stress acting on the parallel return flows required by incompressibility. (author)

  7. Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

    International Nuclear Information System (INIS)

    Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

    1979-01-01

    In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

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

    International Nuclear Information System (INIS)

    Bell, M.G.; Beer, M.

    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 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 a ∼ 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 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 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

  9. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    International Nuclear Information System (INIS)

    1993-12-01

    The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support

  10. Advanced fusion technologies developed for JT-60 superconducting tokamak

    International Nuclear Information System (INIS)

    Sakasai, Akira; Ishida, S.; Matsukawa, M.

    2003-01-01

    The modification of JT-60U is planned as a full superconducting tokamak (JT-60SC). The objectives of the JT-60SC program are to establish scientific and technological bases for the steady-state operation of high performance plasmas and utilization of reduced-activation materials in economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to DEMO reactor have been developed in the superconducting magnet technology and plasma facing components for the design of JT-60SC. To achieve a high current density in a superconducting strand, Nb 3 Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFC) of JT-60SC. The R and D to demonstrate applicability of Nb 3 Al conductor to the TFC by a react-and-wind technique have been carried out using a full-size Nb 3 Al conductor. A full-size NbTi conductor with low AC loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the CFC target was successfully demonstrated on the electron beam irradiation stand. (author)

  11. Understanding L-H transition in tokamak fusion plasmas

    Science.gov (United States)

    Xu, Guosheng; Wu, Xingquan

    2017-03-01

    This paper reviews the current state of understanding of the L-H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L-H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L-H transition. We uncover a comprehensive physical picture of the L-H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L-H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.

  12. Applicability of LBB concept to tokamak-type fusion machine

    International Nuclear Information System (INIS)

    Nakahira, Masataka

    2003-12-01

    A tokamak-type fusion machine has been characterized as having inherent plasma shutdown safety. An extremely small leakage of impurities such as primary cooling water, i.e., less than 0.1 g/s, will cause a plasma disruption. This plasma disruption will induce electromagnetic forces (EM forces) acting in the Vacuum Vessel (VV) and plasma-facing components. The VV forms the physical barrier that encloses tritium and activated dust. If the VV has the possibility of sustaining an unstable fracture from a through crack caused by EM forces, the structural safety will be assured and the inherent safety will be demonstrated. This paper analytically assures the Leak-Before-Break (LBB) concept as applied to the VV and is based on experimental leak rate data of a through crack having a very small opening. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2 mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is therefore concluded that EM forces induced by small leak to terminate plasma will not cause the unstable fracture of VV, and then the inherent safety is demonstrated. (author)

  13. Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

    1979-04-01

    The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

  14. Design of the TFTR [Tokamak Fusion Test Reactor] maintenance manipulator

    International Nuclear Information System (INIS)

    Loesser, G. D.; Heitzenroeder, P.; Bohme, G.; Selig, M.

    1987-01-01

    The Tokamak Fusion Test Reactor (TFTR) plans to generate a total of 3 x 10 21 neutrons during its deuterium-tritium run period in 1900. This will result in high levels of radiation, especially within the TFTR vacuum vessel. The maintenance manipulator's mission is to assist TFTR in meeting Princeton Plasma Physics Laboratory's personnel radiation exposure criteria and in maintaining as-low-as-reasonably-achievable principals by limiting the radiation exposure received by operating and maintenance personnel. The manipulator, which is currently being fabricated and tested by Kernforschungszentrum Karlsruhe, is designed to perform limited, but routine and necessary, functions within the TFTR vacuum torus after activation levels within the torus preclude such functions being performed by personnel. These functions include visual inspection, tile replacement, housekeeping tasks, diagnostic calibrations, and leak detection. To meet its functional objectives, the TFTR maintenance manipulator is required to be operable in TFTR's very high vacuum environment (typically 2 x 10 -8 Torr). It must also be bakeable at 150 degree C and able to withstand the radiation environment

  15. Controlled thermonuclear fusion and the latest progress on China's HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Li Jiangang; Yang Yu

    2003-01-01

    After 50 years of research on controlled thermonuclear fusion, a new stage will be reached in 2003, when a site for the International Thermonuclear Experimental Reactor project will be chosen to start the construction. Scientists hope that this project could herald a new era in which the energy problem will be solved completely. The great progress made on the HT-7 superconducting tokamak in China has provided positive and powerful support for fusion research. The HT-7 is one of the only two superconducting tokamaks in the world that can carry out minute-scale high temperature plasma research, and has achieved a duration of 63.95s for the hot plasma discharge. This is a major step towards real steady-state operation of the tokamak configuration. We present an overview of the latest progress on the tokamak experiments in the Institute of Plasma Physics, Chinese Academy of Sciences

  16. Source-to-incident flux relation for a tokamak fusion test reactor blanket module

    International Nuclear Information System (INIS)

    Imel, G.R.

    1982-01-01

    The source-to-incident 14-MeV flux relation for a blanket module on the Tokamak Fusion Test Reactor is derived. It is shown that assumptions can be made that allow an analytical expression to be derived, using point kernel methods. In addition, the effect of a nonuniform source distribution is derived, again by relatively simple point kernel methods. It is thought that the methodology developed is valid for a variety of blanket modules on tokamak reactors

  17. Fusion reaction spectra produced by anisotropic fast ions in the PLT tokamak

    International Nuclear Information System (INIS)

    Heidbrink, W.W.

    1984-02-01

    For beam-target fusion reactions, collimated measurements of the energy spectrum of one of the reaction products can provide information on the degree of anisotropy of the reacting beam ions. Measurements of the spectrum of 15 MeV protons produced by reactions between energetic 3 He ions and relatively cold deuterons during fast wave minority heating in the PLT tokamak indicate that the velocity distribution of fast 3 He ions is peaked perpendicular to the tokamak magnetic field

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

    International Nuclear Information System (INIS)

    Stacey, Weston M.

    2002-01-01

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

  19. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  20. Boundary perturbations coupled to core 3/2 tearing modes on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Tobias, B; Yu, L; Domier, C W; Luhmann, N C Jr; Austin, M E; Paz-Soldan, C; Turnbull, A D; Classen, I G J

    2013-01-01

    High confinement (H-mode) discharges on the DIII-D tokamak are routinely subject to the formation of long-lived, non-disruptive magnetic islands that degrade confinement and limit fusion performance. Simultaneous, 2D measurement of electron temperature fluctuations in the core and edge regions allows for reconstruction of the radially resolved poloidal mode number spectrum and phase of the global plasma response associated with these modes. Coherent, n = 2 excursions of the plasma boundary are found to be the result of coupling to an n = 2, kink-like mode which arises locked in phase to the 3/2 island chain. This coupling dictates the relative phase of the displacement at the boundary with respect to the tearing mode. This unambiguous phase relationship, for which no counter-examples are observed, is presented as a test for modeling of the perturbed fields to be expected outside the confined plasma. (paper)

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

    International Nuclear Information System (INIS)

    Rose, R.P.

    1979-05-01

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

  2. Important aspects of radiation shielding for fusion reactor tokamaks

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1977-01-01

    Radiation shielding is a key subsystem in tokamak reactors. Design of this shield must evolve from economic and technological trade-off studies that account for the strong interrelations among the various components of the reactor system. These trade-offs are examined for the bulk shield on the inner side of the torus and for the special shields of major penetrations. Results derived are applicable for a large class of tokamak-type reactors

  3. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  4. Recent results and near-term expectations in Tokamak fusion research in the U.S., Europe, and Japan

    International Nuclear Information System (INIS)

    Meade, D.

    1993-01-01

    The development of fusion is often thought about in terms of three different activities: scientific feasibility, engineering feasibility, and economic feasibility. This paper discusses the scientific feasibility of fusion. Reactor temperatures, reactor densities and confinement, particle control, plasma power handling, and self-heating are some of the issues examined. Collaboration and results from research at the Tokamak Fusion Test Reactor (TFTR) at Princeton, the JT-60U in Japan, and JET, the Joint European Torus Tokamak in Oxford are presented

  5. Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Merola, M.; Biggio, M.

    1989-01-01

    Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

  6. Development of DEMO-FNS tokamak for fusion and hybrid technologies

    Science.gov (United States)

    Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-07-01

    The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

  7. Fusion energy research, the tokamak of CRPP-EPFL, electrotechnical equipment

    International Nuclear Information System (INIS)

    1993-01-01

    The topics of this information meeting were: fusion energy research at CRPP, the TCV tokamak, an alternating current generator which does not stress the grid, AC/DC multi-megawatt converters, stabilisation of the plasma, a fast and modular power AC/DC converter. figs., tabs., refs

  8. Evaluation of a nonevaporable getter pump for tritium handling in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Singleton, M.F.; Griffith, C.M.

    1978-01-01

    Lawrence Livermore Laboratory has tested and evaluated a commercially available getter pump for use with tritium in the Tokamak Fusion Test Reactor (TFTR). The pump contains Zr(84%)--Al in cartridge form with a concentric heating unit. It performed well in all tests, except for frequent heater failures

  9. Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Jian Hongbing

    1992-01-01

    The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

  10. Control oriented modeling and simulation of the sawtooth instability in nuclear fusion tokamak plasmas

    NARCIS (Netherlands)

    Witvoet, G.; Westerhof, E.; Steinbuch, M.; Doelman, N.J.; Baar, de M.R.

    2009-01-01

    Tokamak plasmas in nuclear fusion are subject to various instabilities. A clear example is the sawtooth instability, which has both positive and negative effects on the plasma. To optimize between these effects control of the sawtooth period is necessary. This paper presents a simple control

  11. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  12. Edge and coupled core/edge transport modelling in tokamaks

    International Nuclear Information System (INIS)

    Lodestro, L.L.; Casper, T.A.; Cohen, R.H.

    1999-01-01

    Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal E x B drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB , the model predicts transitions to supersonic flow at the inboard midplane. 2D simulations show the importance of E x B flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)

  13. Edge and coupled core-edge transport modelling in tokamaks

    International Nuclear Information System (INIS)

    Lodestro, L.L.; Casper, T.A.; Cohen, R.H.

    2001-01-01

    Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal ExB drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB, the model predicts transitions to supersonic SOL flow at the inboard midplane. 2D simulations show the importance of ExB flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

  15. Fusion-product ash buildup in tokamak with radial electric field

    International Nuclear Information System (INIS)

    Downum, W.B.; Choi, C.K.; Miley, G.H.

    1979-01-01

    The buildup of thermalized fusion products (ash) in a tokamak can seriously limit burn times. Prior studies have concentrated on deposition profile effects on alpha particle transport in tokamaks but have not considered the effect on ash of radial electric fields (either created internally, e.g. due to high-energy alpha leakage, or generated externally). The present study focuses on this issue since it appears that electric fields might offer one approach to control of the ash. Approximate field and source profiles are used, based on prior calculations

  16. Fusion power core engineering for the ARIES-ST power plant

    International Nuclear Information System (INIS)

    Tillack, M.S.; Wang, X.R.; Pulsifer, J.; Malang, S.; Sze, D.K.; Billone, M.; Sviatoslavsky, I.

    2003-01-01

    ARIES-ST is a 1000 MWe fusion power plant based on a low aspect ratio 'spherical torus' (ST) plasma. The ARIES-ST power core was designed to accommodate the unique features of an ST power plant, to meet the top-level requirements of an attractive fusion energy source, and to minimize extrapolation from the fusion technology database under development throughout the world. The result is an advanced helium-cooled ferritic steel blanket with flowing PbLi breeder and tungsten plasma-interactive components. Design improvements, such as the use of SiC inserts in the blanket to extend the outlet coolant temperature range were explored and the results are reported here. In the final design point, the power and particle loads found in ARIES-ST are relatively similar to other advanced tokamak power plants (e.g. ARIES-RS [Fusion Eng. Des. 38 (1997) 3; Fusion Eng. Des. 38 (1997) 87]) such that exotic technologies were not required in order to satisfy all of the design criteria. Najmabadi and the ARIES Team [Fusion Eng. Des. (this issue)] provide an overview of ARIES-ST design. In this article, the details of the power core design are presented together with analysis of the thermal-hydraulic, thermomechanical and materials behavior of in-vessel components. Detailed engineering analysis of ARIES-ST TF and PF systems, nuclear analysis, and safety are given in the companion papers

  17. Comparative study of cost models for tokamak DEMO fusion reactors

    International Nuclear Information System (INIS)

    Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

    2012-01-01

    Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

  18. Testing of low Z coated limiters in tokamak fusion devices

    International Nuclear Information System (INIS)

    Whitely, J.B.; Mullendore, A.W.; Langley, R.A.

    1980-01-01

    Extensive testing on a laboratory scale has been used to select those coatings most suitable for this environment. From this testing which included pulsed electron beam heating, low energy ion bombardment and arcing, chemical vapor deposited coating of TiB 2 and TiC on Poco graphite substrates have been selected and tested as limiters in ISX. Both limiter materials gave clean, stable, reproducible tokamak discharges the first day of operation. After one weeks exposure, the TiC limiter showed only superficial damage with no coating failure. The TiB 2 limiter had some small areas of coating failure. TiC coated graphite limiters have also been briefly tested in the tokamaks Alcator and PDX with favorable results

  19. Burn stability of tokamak fusion plasmas with synergetic current drive

    International Nuclear Information System (INIS)

    Anderson, D.; Lisak, M.; Kolesnichenko, Ya.

    1991-01-01

    The stability of thermonuclear burn in Tokamak-reactors with non-inductive current generated with the simultaneous application of various methods is investigated. Particular emphasis is given to the ITER synergetic current drive scenario involving LH waves, neoclassical effects and NB injection. For ITER-like confinement laws, it is shown that this scenario may be unstable on the plasma skin time scale. Figs

  20. Measurements of fusion product emission profiles in tokamaks

    International Nuclear Information System (INIS)

    Strachan, J.D.; Heidbrink, W.W.; Hendel, H.W.; Lovberg, J.; Murphy, T.J.; Nieschmidt, E.B.; Tait, G.D.; Zweben, S.J.

    1986-11-01

    The techniques and results of fusion product emission profile measurements are reviewed. While neutron source strength profile measurements have been attempted by several methods, neutron scattering is a limitation to the results. Profile measurements using charged fusion products have recently provided an alternative since collimation is much easier for the charged particles

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

    International Nuclear Information System (INIS)

    Ford, G.W.K.

    1987-01-01

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

  2. Code of a Tokamak Fusion Energy Facility ITER

    International Nuclear Information System (INIS)

    Yasuhide Asada; Kenzo Miya; Kazuhiko Hada; Eisuke Tada

    2002-01-01

    The technical structural code for ITER (International Thermonuclear Experimental Fusion Reactor) and, as more generic applications, for D-T burning fusion power facilities (hereafter, Fusion Code) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Introduction of such newly developed technologies as inspection-free automatic welding into the Fusion Code is rationalized by a pilot application of a new code concept of s ystem-based code for integrity . The code concept means an integration of element technical items necessary for construction, operation and maintenance of mechanical components of fusion power facilities into a single system to attain an optimization of the total margin of these components. Unique and innovative items of the Fusion Code are typically as follows: - Use of non-metals; - Cryogenic application; - New design margins on allowable stresses, and other new design rules; - Use of inspection-free automatic welding, and other newly developed fabrication technologies; - Graded approach of quality assurance standard to cover radiological safety-system components as well as non-safety-system components; - Consideration on replacement components. (authors)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  5. Track-mounted remote handling system for the Tokamak Fusion Engineering Test

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Daubert, R.L.; Yount, J.A.

    1982-01-01

    Concepts for remote handling machines (IVM) designed to transverse the interior of toroidal vessels with guidance and support from track systems have been developed for the proposed Tokamak Fusion Engineering Test (TFET). TFET has been proposed as an upgrade for the Tokamak Fusion Test Reactor (TFTR), currently nearing completion. The track-mounted IVMs were conceived to perform in-vessel remote maintenance for TFET, including removal and replacement of pump limiter blades and protective tiles as well as other maintenance-related tasks such as vessel wall inspection leak testing and interior cleanup. The conceptual IVMs consist of three manipulator arms mounted on a common frame member: a single power manipulator arm with high load carrying capacity and two lower-capacity servomanipulator arms. Descriptions of the IVM concepts, in-vessel track systems, and ex-vessel handling systems are presented

  6. A study on the Fusion Reactor - Development of charge exchange recombination spectroscopy for tokamak diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tong Nyong; Kim, Dong Eon; Kim, Dae Sung; Kim, Seong Ho [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1996-09-01

    This project has been carried to train people and accumulate the knowledge and techniques related to the measurement of the profiles of ion temperature, toroidal rotation velocity, and fully-stripped ion density in a fusion tokamak plasma by the development of plasma diagnostics using charge exchange recombination (CER) spectroscopy. Daring the 1 st year, the basic study and review on the charge exchange process and the conceptual design and review of the diagnostics have been conducted. In addition, the various atomic data centers around the world have been surveyed and atomic data related to CER have been constructed. The results of this project can be used to the construction and tokamak machine installation of a CER plasma diagnostic to a new superconducting supported by National Fusion Program. 42 refs., 3 tabs., 16 figs. (author)

  7. Structural characterization of Mumps virus fusion protein core

    International Nuclear Information System (INIS)

    Liu Yueyong; Xu Yanhui; Lou Zhiyong; Zhu Jieqing; Hu Xuebo; Gao, George F.; Qiu Bingsheng; Rao Zihe; Tien, Po

    2006-01-01

    The fusion proteins of enveloped viruses mediating the fusion between the viral and cellular membranes comprise two discontinuous heptad repeat (HR) domains located at the ectodomain of the enveloped glycoproteins. The crystal structure of the fusion protein core of Mumps virus (MuV) was determined at 2.2 A resolution. The complex is a six-helix bundle in which three HR1 peptides form a central highly hydrophobic coiled-coil and three HR2 peptides pack against the hydrophobic grooves on the surface of central coiled-coil in an oblique antiparallel manner. Fusion core of MuV, like those of simian virus 5 and human respiratory syncytium virus, forms typical 3-4-4-4-3 spacing. The similar charecterization in HR1 regions, as well as the existence of O-X-O motif in extended regions of HR2 helix, suggests a basic rule for the formation of the fusion core of viral fusion proteins

  8. Fusion neutron yield and flux calculation on HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Fu Yanzhang; Zhu Yubao; Chen Juequan

    2006-01-01

    Neutron yield and flux have been numerically estimated on HT-7 tokamak. The total fusion neutron yield and neutron flux distribution on different positions and azimuth angles of the device are presented. Analyses on the errors induced by ion temperature and density distribution factors are given in detail. The results of the calculations provide a useful database for neutron diagnostics and neutron radiation protection. (authors)

  9. Safety aspects of activation products in a compact Tokamak Fusion Power Plant

    International Nuclear Information System (INIS)

    Willenberg, H.J.; Bickford, W.E.

    1978-10-01

    Neutron activation of materials in a compact tokamak fusion reactor has been investigated. Results of activation product inventory, dose rate, and decay heat calculations in the blanket and injectors are presented for a reactor design with stainless steel structures. Routine transport of activated materials into the plasma and vacuum systems is discussed. Accidental release of radioactive materials as a result of liquid lithium spills is also considered

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

    International Nuclear Information System (INIS)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.

    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 3 He ions, respectively. When the plasma was compressed, the d(d,n) 3 He fusion reaction rate increased a factor of five, and the 3 He(d,p) 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

  11. The effective cost of tritium for tokamak fusion power reactors with reduced tritium production systems

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Evans, K.

    1983-01-01

    If sufficient tritium cannot be produced and processed in tokamak blankets then at least two alternatives are possible. Tritium can be purchased; or reactors with reduced tritium (RT) content in the plasma can be designed. The latter choice may require development of magnet technology etc., but the authors show that the impact on the cost-of-electricity may be mild. Cost tradeoffs are compared to the market value of tritium. Adequate tritium production in fusion blankets is preferred, but the authors show there is some flexibility in the deployment of fusion if this is not possible

  12. Saturation of alpha particle driven instability in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gorelenkov, N.N.; Chen, Y.; White, R.B.; Berk, H.L.

    1999-01-01

    A nonlinear theory of kinetic instabilities near threshold [Berk et al., Plasma Phys. Rep. 23, 842 (1997)] is applied to calculate the saturation level of toroidicity-induced Alfven eigenmodes (TAE), and to be compared with the predictions of δf method calculations (Y. Chen, Ph.D. thesis, Princeton University, 1998). Good agreement is observed between the predictions of both methods and the predicted saturation levels are comparable to experimentally measured amplitudes of the TAE oscillations in Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)]. copyright 1999 American Institute of Physics

  13. Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically in influence its operation, safety and performance. Erosion will increase to a scale of several centimetres 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 stimulated an internationally co-ordinated effort in the part of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material inter actions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling 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 avenues for their resolution are presented. (author)

  14. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT 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 cm 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 co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding 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/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling 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 avenues for their resolution are presented. (orig.)

  15. Power supply requirements for a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.L.

    1979-01-01

    The power supply requirements for a 7-m major radius commerical tokamak reactor have been examined, using a system approach combining models of the reactor and poloidal coil set, plasma burn cycle and magnetohydrodynamics calculations, and power supply characteristics and cost data. A conventional system using a motor-generator flywheel set and solid-state rectifier-inverter power supplies was studied in addition to systems using a homopolar generator, superconducting energy storage inductor, and dump resistors. The requirements and cost of the power supplies depend on several factors but most critically on the ohmic heating ramp time used for startup. Long ramp times (greater than or equal to 8 s) seem to be feasible, from the standpoint of resistive volt-second losses, and would appear to make conventional systems quite competitive with nonconventional ones, which require further research and development

  16. Power supply requirements for a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.L.

    1979-02-01

    The power supply requirements for a 7-M major radius commercial tokamak reactor have been examined, using a system approach combining models of the reactor and poloidal coil set, plasma burn cycle and MHD calculations, and power supply characteristics and cost data. A conventional system using an MGF set and solid-state rectifier/inverter power supplies was studied in addition to systems using a homopolar generator, superconducting energy storage inductor, and dump resistors. The requirements and cost of the power supplies depend on several factors but most critically on the ohmic heating ramp time used for startup. Long ramp times (approx. > 8 s) seems to be feasible, from the standpoint of resistive volt-second losses, and would appear to make conventional systems quite competitive with nonconventional ones, which require further research and development

  17. Tokamak Fusion Test Reactor. Final conceptual design report

    International Nuclear Information System (INIS)

    1976-02-01

    The TFTR is the first U.S. magnetic confinement device planned to demonstrate the fusion of D-T at reactor power levels. This report addresses the physics objectives and the engineering goals of the TFTR project. Technical, cost, and schedule aspects of the project are included

  18. Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kessel, C.E.; Mau, T.K.; Miller, R.L.; Najmabadi, F.; Chan, V.S.; Chu, M.S.; LaHaye, R.; Lao, L.L.; Petrie, T.W.; Politzer, P.; John, St. H.E.; Snyder, P.; Staebler, G.M.; Turnbull, A.D.; West, W.P.

    2003-01-01

    The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented

  19. Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

    Energy Technology Data Exchange (ETDEWEB)

    S.C. Jardin; C.E. Kessel; T.K. Mau; R.L. Miller; F. Najmabadi; V.S. Chan; M.S. Chu; R. LaHaye; L.L. Lao; T.W. Petrie; P. Politzer; H.E. St. John; P. Snyder; G.M. Staebler; A.D. Turnbull; W.P. West

    2003-10-07

    The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.

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

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  1. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  2. Plasma current sustainment after iron core saturation in the STOR-M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Ding, Y.; Hubeny, M.; Lu, Y.; Onchi, T.; McColl, D.; Xiao, C.; Hirose, A. [Plasma Physics Laboratory, University of Saskatchewan, 116 Science Place, Saskatoon, SK S7N 5E2 (Canada)

    2014-10-15

    Highlights: • Plasma current can be started up by small iron core without central solenoid. • Iron core removes central solenoid. • Plasma current can be maintained after iron core saturation. • Hysteresis curve shows the partial core saturation. • Image field from iron core is estimated during discharge. • Spherical tokamak reactor without CS is proposed using the small iron core. - Abstract: We propose to use of a small iron core transformer to start up the plasma current in a spherical tokamak (ST) reactor without central solenoid (CS). Taking advantage of the high aspect ratio of the STOR-M iron core tokamak, we have demonstrated that the plasma current up to 10–15 kA can be started up using the outer Ohmic heating (OH) coils without CS, and that the plasma current can be maintained further by increasing the outer OH coil current during iron core saturation phase. When the magnetizing current reaches 1.2 kA and the iron core becomes saturated, the third capacitor bank connected to the outer OH coils is discharged to maintain the plasma current. The plasma current is slightly increased and maintained for additional 5 ms as expected from numerical calculations. Core saturation has been clearly observed on the hysteresis curve. This is the first experimental demonstration of the feasibility of slow transition from the iron core to air core transformer phase without CS. The results implies that a plasma current can be initiated by a small iron core and could be ramped up by additional heating and vertical field after iron core saturation in future STs without CS.

  3. Plasma driving system requirements for commercial tokamak fusion reactors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

    1978-01-01

    The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

  4. Plasma driving system requirements for commercial tokamak fusion reactors

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

    1977-01-01

    The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

  5. Spectra of heliumlike krypton from tokamak fusion test reactor plasmas

    International Nuclear Information System (INIS)

    Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Smith, A.; Fraenkel, B.

    1993-04-01

    Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 Angstrom including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport

  6. The ARIES-ST study: Assessment of the spherical tokamak concept as fusion power plants

    International Nuclear Information System (INIS)

    Najmabadi, F.; Tillack, M.; Miller, R.; Mau, T.K.; Jardin, S.; Stambaugh, R.; Steiner, D.; Waganer, L.

    2001-01-01

    Recent experimental achievements and theoretical studies have generated substantial interest in the spherical tokamak concept. The ARIES-ST study was undertaken as a national U.S. effort to investigate the potential of the spherical tokamak concept as a fusion power plant and as a vehicle for fusion development. The 1000-MWe ARIES-ST power plant has an aspect ratio of 1.6, a major radius of 3.2 m, a plasma elongation (at 95% flux surface) of 3.4 and triangularity of 0.64. This configuration attains a β of 54% (which is 90% of the maximum theoretical β). While the plasma current is 31 MA, the almost perfect alignment of bootstrap and equilibrium current density profiles results in a current-drive power of only 31 MW. The on-axis toroidal field is 2.1 T and the peak field at the TF coil is 7.6 T, which leads to 288 MW of Joule losses in the normal-conducting TF system. The ARIES-ST study has highlighted many areas where tradeoffs among physics and engineering systems are critical in determining the optimum regime of operation for spherical tokamaks. Many critical issues also have been identified which must be resolved in R and D programs. (author)

  7. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

    International Nuclear Information System (INIS)

    Nishio, Satoshi

    1998-01-01

    If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

  8. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

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

  9. Some stress-related issues in tokamak fusion reactor first walls

    International Nuclear Information System (INIS)

    Majumdar, S.; Pai, B.; Ryder, R.H.

    1987-01-01

    Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

  10. STAR Power, an Interactive Educational Fusion CD with a Dynamic, Shaped Tokamak Power Plant Simulator

    Science.gov (United States)

    Leuer, J. A.; Lee, R. L.; Kellman, A. G.; Chapman Nutt, G. C., Jr.; Holley, G.; Larsen, T. A.

    2000-10-01

    We describe an interactive, educational fusion adventure game developed within our fusion education program. The theme of the adventure is start-up of a state-of-the-art fusion power plant. To gain access to the power plant control room, the student must complete several education modules, including topics on building an atom, fusion reactions, charged particle motion in electric and magnetic fields, and building a power plant. Review questions, a fusion video, library material and glossary provide additional resources. In the control room the student must start-up a complex, dynamic fusion power plant. The simulation model contains primary elements of a tokamak based device, including a magnetic shaper capable of producing limited and diverted elongated plasmas. A zero dimensional plasma model based on ITER scaling and containing rate based conservation equations provides dynamic feedback through major control parameters such as toroidal field, fueling rate and heating. The game is available for use on PC and Mac. computers. Copies will be available at the conference.

  11. Ion cyclotron and spin-flip emissions from fusion products in tokamaks

    International Nuclear Information System (INIS)

    Arunasalam, V.; Greene, G.J.; Young, K.M.

    1993-02-01

    Power emission by fusion products of tokamak plasmas in their ion cyclotron range of frequencies (ICRF) and at their spin-flip resonance frequency is calculated for some specific model fusion product velocity-space distribution functions. The background plasma of say deuterium (D) is assumed to be in equilibrium with a Maxwellian distribution both for the electrons and ions. The fusion product velocity distributions analyzed here are: (1) A monoenergetic velocity space ring distribution. (2) A monoenergetic velocity space spherical shell distribution. (3) An anisotropic Maxwellian distribution with T perpendicular ≠ T parallel and with appreciable drift velocity along the confining magnetic field. Single ''dressed'' test particle spontaneous emission calculations are presented first and the radiation temperature for ion cyclotron emission (ICE) is analyzed both for black-body emission and nonequilibrium conditions. Thresholds for instability and overstability conditions are then examined and quasilinear and nonlinear theories of the electromagnetic ion cyclotron modes are discussed. Distinctions between ''kinetic or causal instabilities'' and ''hydrodynamic instabilities'' are drawn and some numerical estimates are presented for typical tokamak parameters. Semiquantitative remarks are offered on wave accessibility, mode conversion, and parametric decay instabilities as possible for spatially localized ICE. Calculations are carried out both for k parallel = 0 for k parallel ≠ 0. The effects of the temperature anisotropy and large drift velocities in the parallel direction are also examined. Finally, proton spin-flip resonance emission and absorption calculations are also presented both for thermal equilibrium conditions and for an ''inverted'' population of states

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  14. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

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

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Barry, K.

    1980-01-01

    The purpose of the STARFIRE study is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup, superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield

  16. A design study of superconducting energy storage system for a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Ueda, Kazuo

    1979-01-01

    A design study of a superconducting inductive energy storage system (SC-IES) has been carried out in commission with JAERI. The SC-IES is to be applied to the power supply system for a tokamak experimental fusion reactor. The study was initiated with the definition of the requirement for the SC-IES and selection of the coil shape. The design of the coil and the cryostat has been followed. The design parameters are: stored energy 10 GJ, B max 8 T, conductor Nb-Ti, overall size 18 m (diameter) x 10 m (height). Technical problems and usefullness of SC-IES are discussed also. (author)

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

    International Nuclear Information System (INIS)

    M.E. Lumia; C.A. Gentile

    2002-01-01

    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

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

    CERN Document Server

    Lumia, M E

    2002-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    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.

  20. Edge turbulence effect on ultra-fast swept reflectometry core measurements in tokamak plasmas

    Science.gov (United States)

    Zadvitskiy, G. V.; Heuraux, S.; Lechte, C.; Hacquin, S.; Sabot, R.

    2018-02-01

    Ultra-fast frequency-swept reflectometry (UFSR) enables one to provide information about the turbulence radial wave-number spectrum and perturbation amplitude with good spatial and temporal resolutions. However, a data interpretation of USFR is quiet tricky. An iterative algorithm to solve this inverse problem was used in past works, Gerbaud (2006 Rev. Sci. Instrum. 77 10E928). For a direct solution, a fast 1D Helmholtz solver was used. Two-dimensional effects are strong and should be taken into account during data interpretation. As 2D full-wave codes are still too time consuming for systematic application, fast 2D approaches based on the Born approximation are of prime interest. Such methods gives good results in the case of small turbulence levels. However in tokamak plasmas, edge turbulence is usually very strong and can distort and broaden the probing beam Sysoeva et al (2015 Nucl. Fusion 55 033016). It was shown that this can change reflectometer phase response from the plasma core. Comparison between 2D full wave computation and the simplified Born approximation was done. The approximated method can provide a right spectral shape, but it is unable to describe a change of the spectral amplitude with an edge turbulence level. Computation for the O-mode wave with the linear density profile in the slab geometry and for realistic Tore-Supra density profile, based on the experimental data turbulence amplitude and spectrum, were performed to investigate the role of strong edge turbulence. It is shown that the spectral peak in the signal amplitude variation spectrum which rises with edge turbulence can be a signature of strong edge turbulence. Moreover, computations for misaligned receiving and emitting antennas were performed. It was found that the signal amplitude variation peak changes its position with a receiving antenna poloidal displacement.

  1. Gyro-Landau fluid model of tokamak core fluctuations

    International Nuclear Information System (INIS)

    Leboeuf, J.N.; Carreras, B.A.; Dominguez, N.; Hedrick, C.L.; Sidikman, K.L.; Lynch, V.E.; Drake, J.B.; Walker, D.W.

    1992-01-01

    Dissipative trapped electron modes (DTEM) may be one of the causes of deterioration of confinement in tokamak and stellatator plasmas. We have implemented a fluid model to study DTEM turbulence in slab geometry. The electron dynamics include in addition to the adiabatic part, a non-adiabatic piece modeled with an i-delta-type response. The ion dynamics include Landau damping and FLR corrections through Landau fluid approximate techniques and Pade approximants for Γ 0 (b)=I 0 (b)e -b . The model follows from the gyrokinetic equation. Evolution equations, which closely resemble those used in standard reduced MHD, are presented since these are better suited to non-linear calculations. The numerical results of radially resolved calculations will be discussed. A recently developed hybrid model, which consists of a gyrokinetic implementation for the ions using particles and the same description for the electron dynamics as in the fluid model, will also be presented

  2. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

  3. Design study of electrical power supply system for tokamak fusion power reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

  4. Developing maintainability for tokamak fusion power systems. Phase II report. Volume I: executive summary

    International Nuclear Information System (INIS)

    Fuller, G.M.; Zahn, H.S.; Mantz, H.C.; Kaletta, G.R.; Waganer, L.M.; Carosella, L.A.; Conlee, J.L.

    1978-11-01

    The purpose of this report is to identify design features of fusion power reactors which contribute to the achievement of high levels of maintainability. Volume 1, the Executive Summary, presents the progress achieved toward this objective in this phase and includes a comparison with the results of the first phase study efforts. A series of maintainability design guidelines and an improved maintenance system are defined as initial steps in developing the requirements for a maintainable tokamak fusion power system. The principle comparative studies that are summarized include the determination of the benefits of various vacuum wall arrangements, the effect of unscheduled and scheduled maintenance of the first wall/blanket, some initial investigation of maintenance required for subsystems other than the first wall/blanket, and the impact of maintenance equipment failures

  5. Fusion plasma theory: Task 3, Auxiliary heating in tokamaks

    International Nuclear Information System (INIS)

    Scharer, J.E.

    1989-07-01

    The research that we have accomplished during the past year (1988--1989) includes the topics of ICRF fast wave waveguide coupling to H-mode profiles simulating CIT and full wave ICRF field solutions and a power conservation relation based on fundamental principles with JET and CIT heating applications. We have also published work on Fokker-Planck simulations of minority ion ICRF strong core electron sawteeth processes in JET, a publication on the effect of plasma edge density fluctuation and ponderomotive force effects on the coupling of ion Bernstein waves and a publication on the coupling of dielectric filled waveguides to plasmas in the ICRF. The analysis of ICRF H-mode coupling is crucial to the economic success of proposed ignition devices such as CIT and ITER. We have analyzed the coupling of ICRF waveguide launchers to H-mode density profiles modelled by a pedestal width and Gaussian edge variations with gradients comparable to current machines. We find that the launcher aperture spectrum, density gradients and width of the pedestal are important parameters in determining the coupling efficiency. The launcher-plasma admittance spectrum in k y -k z space is utilized to show that the H-mode launcher reflections increase when compared to the L-mode profile, but that they can be handled by launcher matching circuits and modest modifications of the H-mode profile. We plan to analyze the recent successful JET ICRF H-mode operation utilizing our formalism. We have also carried out a full wave ICRF field solution and the associated power conservation relation with expressions evaluated up to the third harmonic. We have implemented this in a computer code which utilizes invariant imbedding to solve the system of equations. 7 refs., 1 tab

  6. Task III: auxillary heating in tokamaks and tandem mirrors. Progress report on fusion plasma theory

    International Nuclear Information System (INIS)

    Scharer, J.E.

    1986-06-01

    The research we have accomplished with this grant has focused on ICRF coupling, wave propagation, heating and breakeven studies for tokamaks such as JET. The highlights include fundamental work on a differential equation for wave fields incorporating equilibrium gradients, strong absorption and mode conversion and a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas. We have also formulated and developed a code which solves differential equation for ICRF waveguide coupling in tokamak edge density regions. We are also examining the excitation of ion Bernstein waves from fast magnetosonic waves occurring in density gradients. Our current efforts involve the explanation of current JET ICRF results such as the large electron sawteeth in the core region in terms of hot, non-Maxwellian ICRF theory

  7. Equilibrium and stability studies for an iron-core tokamak with a poloidal divertor

    International Nuclear Information System (INIS)

    Solano, E.R.; Neilson, G.H.; Lao, L.L.

    1989-01-01

    A study of plasma equilibrium and stability in a tokamak with an unsaturated iron core is presented. A spool model is developed for the iron. Both, a simplified force balance code and a Grad-Shafranov solver are used to study the plasma equilibrium. It is observed that the iron can strongly modify the conditions for equilibrium and stability, and in some cases an infinite cylinder model for the iron core is not adequate. New criteria for plasma position stability in the presence of an iron core are introduced. 17 refs., 4 figs., 3 tabs

  8. ARIES-AT: An advanced tokamak, advanced technology fusion power plant

    International Nuclear Information System (INIS)

    Najmabadi, F.; Jardin, S.C.; Tillack, M.; Waganer, L.M.

    2001-01-01

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant. Several avenues were pursued in order to arrive at plasmas with a higher β and better bootstrap alignment compared to ARIES-RS that led to plasmas with higher β N and β. Advanced technologies that are examined in detail include: (1) Possible improvements to the overall system by using high-temperature superconductors, (2) Innovative SiC blankets that lead to a high thermal cycle efficiency of ∼60%; and (3) Advanced manufacturing techniques which aim at producing near-finished products directly from raw material, resulting in low-cost, and reliable components. The 1000-MWe ARIES-AT design has a major radius of 5.4 m, minor radius of 1.3 M, a toroidal β of 9.2% (β N =6.0) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current drive power is 24 MW. The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (5c/kWh), which is competitive with those projected for other sources of energy. (author)

  9. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

    1995-01-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

  10. Combined core/boundary layer plasma transport simulations in tokamaks

    International Nuclear Information System (INIS)

    Prinja, A.K.; Schafer, R.F. Jr.; Conn, R.W.; Howe, H.C.

    1987-01-01

    Significant new numerical results are presented from self-consistent core and boundary or scrape-off layer plasma simulations with 3-D neutral transport calculations. For a symmetric belt limiter it is shown that, for plasma conditions considered here, the pump limiter collection efficiency increases from 11% to 18% of the core efflux as a result of local reionization of blade deflected neutrals. This hitherto unobserved effect causes a significant amplification of upstream ion flux entering the pump limiter. Results from coupling of an earlier developed two-zone edge plasma model ODESSA to the PROCTR core plasma simulation code indicates that intense recycling divertor operation may not be possible because of stagnation of upstream flow velocity. This results in a self-consistent reduction of density gradient in an intermediate region between the central plasma and separatrix, and a concomitant reduction of core-efflux. There is also evidence of increased recycling at the first wall. (orig.)

  11. The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

    International Nuclear Information System (INIS)

    Shimada, Ryuichi; Tani, Keiji; Kishimoto, Hiroshi; Tamura, Sanae; Yanabu, Satoru.

    1979-01-01

    In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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).

  13. Collimator type monochromator as a possible impurities monitor for fusion plasmas. Preliminary tests on the Tokamak TM-1-MH

    International Nuclear Information System (INIS)

    Musa, G.; Lungu, C.P.; Badalec, J.; Jakubka, K.; Kopecky, V.; Stoeckel, J.; Zacek, F.

    1984-09-01

    A collimator type monochromator has been tested for the first time as the impurity monitor on Tokamak. The possibility to use this type of monochromator in fusion devices is analyzed and a monoslit device is proposed as a convenient monitor for impurities. (authors)

  14. Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

    Science.gov (United States)

    Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

    2000-12-01

    Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

  15. Ion cyclotron and spin-flip emissions from fusion products in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Arunasalam, V.; Greene, G.J.; Young, K.M.

    1993-02-01

    Power emission by fusion products of tokamak plasmas in their ion cyclotron range of frequencies (ICRF) and at their spin-flip resonance frequency is calculated for some specific model fusion product velocity-space distribution functions. The background plasma of say deuterium (D) is assumed to be in equilibrium with a Maxwellian distribution both for the electrons and ions. The fusion product velocity distributions analyzed here are: (1) A monoenergetic velocity space ring distribution. (2) A monoenergetic velocity space spherical shell distribution. (3) An anisotropic Maxwellian distribution with T [perpendicular] [ne] T[parallel]and with appreciable drift velocity along the confining magnetic field. Single dressed'' test particle spontaneous emission calculations are presented first and the radiation temperature for ion cyclotron emission (ICE) is analyzed both for black-body emission and nonequilibrium conditions. Thresholds for instability and overstability conditions are then examined and quasilinear and nonlinear theories of the electromagnetic ion cyclotron modes are discussed. Distinctions between kinetic or causal instabilities'' and hydrodynamic instabilities'' are drawn and some numerical estimates are presented for typical tokamak parameters. Semiquantitative remarks are offered on wave accessibility, mode conversion, and parametric decay instabilities as possible for spatially localized ICE. Calculations are carried out both for k[parallel] = 0 for k[parallel] [ne] 0. The effects of the temperature anisotropy and large drift velocities in the parallel direction are also examined. Finally, proton spin-flip resonance emission and absorption calculations are also presented both for thermal equilibrium conditions and for an inverted'' population of states.

  16. All Metal Iron Core For A Low Aspect Ratio Tokamak

    International Nuclear Information System (INIS)

    Gates, D.A.; Jun, C.; Zatz, I.; Zolfaghari, A.

    2010-01-01

    A novel concept for incorporating a iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance separators between the toroidal field windings. This design avoids the inherent problems of a multiturn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron-induced conductivity. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance separators. Useful loop voltage time histories have been obtained using achievable resistivities.

  17. Design of the air-core transformer in spherical tokamak

    International Nuclear Information System (INIS)

    Wang Zhongtian; Jian Guangde; Li Fangzhu; Mao Guoping

    2002-01-01

    An ideal current distribution in the air-core transformer coils is obtained using variation principle. Climbing mountain method is utilized for optimizing the dimension and position of the real coils. Not only can the requirement of minimizing the stray field in the plasma region be guaranteed, but also integer turns for the coil can be realized. The latter brings a significant convenience to engineering

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

    International Nuclear Information System (INIS)

    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, Y.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.

    1994-01-01

    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 Fluorinert TM 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

  19. Fast computational scheme for feedback control of high current fusion tokamaks

    International Nuclear Information System (INIS)

    Dong, J.Q.; Khayrutdinov, R.; Azizov, E.; Jardin, S.

    1992-01-01

    An accurate and fast numerical model of tokamak plasma evolution is presented. In this code (DINA) the equilibrium problem of plasmas with free boundaries in externally changing magnetic fields is solved simultaneously with the plasma transport equation. The circuit equations are solved for the vacuum vessel and passive and active coils. The code includes pellet injection, neutral beam heating, auxiliary heating, and alpha particle heating. Bootstrap and beam-driven plasma currents are accounted for. An inverse variable technique is utilized to obtain the coordinates of the equilibrium magnetic surfaces. This numerical algorithm permits to determine the flux coordinates very quickly and accurately. The authors show that using the fully resistive MHD analysis the region of stability (to vertical motions) is wider than using the rigid displacement model. Comparing plasma motions with the same gain, it is seen that the plasma oscillates more in the rigid analysis than in the MHD analysis. They study the influence of the pick up coil's location and the possibility of control of the plasma vertical position. They use a simple modification of the standard control law that enables the control of the plasma with pick up coils located at any position. This flexibility becomes critical in the design of future complex high current tokamak systems. The fully resistive MHD model permits to obtain accurate estimates of the plasma response. This approach yields computational time savings of one to two orders of magnitude with respect to other existing MHD models. In this sense, conventional numerical algorithms do not provide suitable models for application of modern control techniques into real time expert systems. The proposed inverse variable technique is rather suitable for incorporation in a comprehensive expert system for feedback control of fusion tokamaks in real time

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

    International Nuclear Information System (INIS)

    Ogawa, Yuichi

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Fitzpatrick, R.

    1997-01-01

    Stabilization of the 'resistive shell mode' is vital to the success of the 'advanced tokamak' concept. The most promising reactor relevant approach is to apply external feedback using, for instance, the previously proposed 'fake rotating shell' scheme [R. Fitzpatrick and T. H. Jensen, Phys. Plasmas 3, 2641 (1996)]. This scheme, like other simple feedback schemes, only works if the feedback controlled conductors are located inside the 'critical radius' 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 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 1997 American Institute of Physics

  2. Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

  3. Design study of superconducting inductive energy storages for tokamak fusion reactor

    International Nuclear Information System (INIS)

    1977-08-01

    Design of the superconducting inductive energy storages (SC-IES) has been studied. One SC-IES is for the power supply system in a experimental tokamak fusion reactor, and the other in a future practical reactor. Study started with definition of the requirements of SC-IES, followed by optimization of the coil shape and determination of major parameters. Then, the coil and the vessel were designed, including the following: for SC-IES of the experimental reactor, stored energy 10 GJ, B max 8 T, conductor NbTi and size 18 m diameter x 10 m height; for SC-IES of the practical reactor, stored energy 56 GJ, B max 10.5 T, conductor Nb 3 Sn and size 26 m diameter x 15 m height. Design of the coil protection system and an outline of the auxiliary systems (for refrigeration and evacuation) are also given, and further, problems and usefullness of SC-IES. (auth.)

  4. Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

    International Nuclear Information System (INIS)

    Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

    1989-01-01

    The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

  5. Fusion Plasma Theory Grant: Task 3, Auxiliary Radiofrequency Heating of Tokamaks

    International Nuclear Information System (INIS)

    Scharer, J.E.

    1993-06-01

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues. We have made progress in developing a ''3-D'' cavity backed antenna array code to examine ICRF coupling to general plasma edge profiles. The effects of the finite antenna length and feeders as well as Faraday shield blade angle are being examined. We are also developing an analysis to examine large k perpendicular ρ gyroradius interaction between alpha or beam particles and ICRF waves. This topic has important applications in the areas of ICRF heating for deuterium-tritium fusion plasmas, TAE modes, ash removal and minority ion current drive. Research progress, publications, and conference and workshop presentations are summarized in this report

  6. Electron cyclotron measurements with the fast scanning heterdyne radiometer on the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Taylor, G.; Efthimion, P.C.; McCarthy, M.P.; Fredd, E.A.; Cutler, R.C.

    1986-01-01

    Three fast scanning heterodyne receivers, swept between 75-110 GHz, 110-170 GHz, and 170-210 GHz, have measured electron cyclotron emission on the horizontal midplane of the tokamak fusion test reactor (TFTR) plasma. A second harmonic microwave mixer in the 170-210 GHz receiver allows the use of a 75-110 GHz backward wave oscillator as a swept local oscillator. Electron temperature profile evolution data with a time resolution of 2 msec and a profile acquisition rate of 250 Hz are presented for gas-fuelled and pellet-fuelled ohmic and neutral beam heated plasmas with toroidal fields up to 5.2 tesla. Recent results from a swept mode absolute calibration technique which can improve the accuracy and data collection efficiency during in-situ calibration are also presented

  7. Using plasma waves to create in tokamaks the necessary quasi-stationary conditions for controlled fusion

    International Nuclear Information System (INIS)

    Moreau, D.

    1993-04-01

    It is studied, on the one hand, how using hybrid waves with frequency near from lower hybrid frequency in fusion plasma. Works about coupling waves in plasma (chap.I), their propagation and response of the plasma to the absorption of the waves (chap.II). This method is the most effective until today. Because of limits, it has been investigated, on the other hand, fast magnetosonic wave to control current density in the centre of the discharge in a reactor or a very hot plasma. Theoretical study (chap.III) and experimental results (chap.IV) are presented. Experiments are in progress or planned in following tokamaks: D3-D (USA), JET (Europe), TORE SUPRA (France), JT-60 (Japan). figs. refs. tabs

  8. Dielectronic satellite spectra of hydrogenlike iron from TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    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-α 1 and Ly-α 2 lines and the associated dielectronic satellites, which are due to transitions 1snl-2pnl' with n ≥ 2, as well as the heliumlike 1s 2 ( 1 S 0 )-1s4p( 1 P 1 )and both hydrogenlike Ly-β 1 and Ly-β 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

  9. ARIES-I Fusion-Power-Core Engineering

    International Nuclear Information System (INIS)

    Sharafat, S.; Najmabadi, F.; Wong, C.P.C.

    1991-01-01

    The ARIES research program is a multi-institutional project, the goal of which is to determine the economic, safety, and environmental potential of tokamak fusion reactors. The ARIES-I steady-state tokamak reactor is a conceptual, DT-burning, 1000 MWe reactor with a major radius of 6.75 m, a minor radius of 1.5 m, and an average neutron wall loading of 2.5 MW/m 2 . The ARIES-I plasma operates in the first MHD stability regime with a toroidal beta of 1.9%. The choice to operate in the first stability regime, with a high aspect ratio and with a low plasma current, leads to the need for high magnetic field to achieve adequate fusion power density (β 2 B 4 ). The toroidal field at the plasma center is 11 T and the maximum field at the coil is 21 T. Nonetheless, it is found that the maximum stress in the structural material of these magnets is ∝700 MPa and industrially available alloys can be used. The impurity-control and particle-exhaust system is based on a high recycling double-null divertor system. The low-activation silicon-carbide (SiC) composite is used as structural material. The breeder material, Li 2 ZrO 3 , and the multiplier material, Be, are both sphere-packed between poloidally nested SiC-composite shells. The divertor plates consist of SiC-composite tube shells protected with 2 mm-thick tungsten armor. The first wall, blanket, shield, and divertor are all helium cooled with an inlet coolant temperature of 350deg C at a pressure of 10 MPa. The high helium-outlet temperature of 650deg C ensures a relatively high gross thermal efficiency of 49%. The ARIES-I design has demonstrated that tokamak reactors have the potential to achieve a high level of safety coupled with a Class-C waste-disposal rating. (orig.)

  10. Steady state technologies for tokamak based fusion neutron sources and hybrids

    International Nuclear Information System (INIS)

    Azizov, E.A.; Kuteev, B.V.

    2015-01-01

    Full text of publication follows. The development of demonstration fusion neutron sources for fusion nuclear science activity and hybrid applications has reached the stage of conceptual design on the basis of tokamak device in Russia. The conceptual design of FNS-ST has been completed in details (plasma current 1.5 MA, magnetic field 1.5 T, major radius 0.5 m, aspect ratio 1.67 and auxiliary heating power up to 15 MW) [1, 2]. A comparison of physical plasma parameters and economics for FNS-ST and a conventional tokamak FNS-CT (plasma current 1.5 MA, magnetic field 6.7 T, major radius 2.25 m, aspect ratio 3 and auxiliary heating power up to 30 MW) has been fulfilled [3]. This study suggested the feasibility to reach 1-20 MW of fusion power using these magnetic configuration options. Nevertheless, the efficiency of neutron production Q remains comparable for both due to the beam fusion input. The total ST-economics for the full project including operation and utilization costs is by a factor of 2 better than of CT. Zero [4] and one-dimensional [5] models have been developed and used in this system analysis. The characteristics of plasma confinement, stability and current drive in operation have been confirmed by numerous benchmarking simulations of modern experiments. Scenarios allowing us to reach and maintain steady state operation have been considered and optimized. The results of these studies will be presented. Prospective technical solutions for SSO-technology systems have been evaluated, and the choice of enabling technologies and materials of the basic FNS options has been made. A conceptual design of a thin-wall water cooled vacuum chamber for heat loadings up to 1.5 MW/m 2 has been fulfilled. The chamber consists of 2 mm Be tiles, pre-shaped CuCrZr 1 mm shell and 1 mm of stainless steel shell as a structural material. A concept of double-null divertor for FNS-ST has been offered that is capable to withstand heat fluxes up to 6 MW/m 2 . Lithium dust

  11. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Loennroth, J.-S.

    2009-01-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

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

    International Nuclear Information System (INIS)

    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 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 ∼ 10 megawatts (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

  13. Base isolation technique for tokamak type fusion reactor using adaptive control

    International Nuclear Information System (INIS)

    Koizumi, T.; Tsujiuchi, N.; Kishimoto, F.; Iida, H.; Fujita, T.

    1991-01-01

    In this paper relating to the isolation device of heavy structure such as nuclear fusion reactor, a control rule for reducing the response acceleration and relative displacement simultaneously was formulated, and the aseismic performance was improved by employing the adaptive control method of changing the damping factors of the system adaptively every moment. The control rule was studied by computer simulation, and the aseismic effect was evaluated in an experiment employing a scale model. As a results, the following conclusions were obtained. (1) By employing the control rule presented in this paper, both absolute acceleration and relative displacement can be reduced simultaneously without making the system unstable. (2) By introducing this control rule in a scale model assuming the Tokamak type fusion reactor, the response acceleration can be suppressed down to 78 % and also the relative displacement to 79 % as compared with the conventional aseismic method. (3) The sensitivities of absolute acceleration and relative displacement with respect to the control gain are not equal. However, by employing the relative weighting factor between the absolute acceleration and relative displacement, it is possible to increase the control capability for any kind of objective structures and appliances. (author)

  14. Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

    International Nuclear Information System (INIS)

    Chen Yiping; Liu Songlin

    2010-01-01

    Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

  15. Demonstration tokamak fusion power plant for early realization of net electric power generation

    International Nuclear Information System (INIS)

    Hiwatari, R.; Okano, K.; Asaoka, Y.; Shinya, K.; Ogawa, Y.

    2005-01-01

    A demonstration tokamak fusion power plant Demo-CREST is proposed as the device for early realization of net electric power generation by fusion energy. The plasma configuration for Demo-CREST is optimized to satisfy the electric breakeven condition (the condition for net electric power, P e net = 0 MW) with the plasma performance of the ITER reference operation mode. This optimization method is considered to be suitable for the design of a demonstration power plant for early realization of net electric power generation, because the demonstration power plant has to ensure the net electric generation. Plasma performance should also be more reliably achieved than in past design studies. For the plasma performance planned in the present ITER programme, net electric power from 0 to 500 MW is possible with Demo-CREST under the following engineering conditions: maximum magnetic field 16 T, thermal efficiency 30%, NBI system efficiency 50% and NBI current drive power restricted to 200 MW. By replacing the blanket system with one of higher thermal efficiency, a net electric power of about 1000 MW is also possible so that the performance of the commercial plant with Demo-CREST can also be studied from the economic point of view. The development path from the experimental reactor 'ITER' to the commercial plant 'CREST' through the demonstration power plant 'Demo-CREST' is proposed as an example of the fast track concept. (author)

  16. Path to Market for Compact Modular Fusion Power Cores

    Science.gov (United States)

    Woodruff, Simon; Baerny, Jennifer K.; Mattor, Nathan; Stoulil, Don; Miller, Ronald; Marston, Theodore

    2012-08-01

    The benefits of an energy source whose reactants are plentiful and whose products are benign is hard to measure, but at no time in history has this energy source been more needed. Nuclear fusion continues to promise to be this energy source. However, the path to market for fusion systems is still regularly a matter for long-term (20 + year) plans. This white paper is intended to stimulate discussion of faster commercialization paths, distilling guidance from investors, utilities, and the wider energy research community (including from ARPA-E). There is great interest in a small modular fusion system that can be developed quickly and inexpensively. A simple model shows how compact modular fusion can produce a low cost development path by optimizing traditional systems that burn deuterium and tritium, operating not only at high magnetic field strength, but also by omitting some components that allow for the core to become more compact and easier to maintain. The dominant hurdles to the development of low cost, practical fusion systems are discussed, primarily in terms of the constraints placed on the cost of development stages in the private sector. The main finding presented here is that the bridge from DOE Office of Science to the energy market can come at the Proof of Principle development stage, providing the concept is sufficiently compact and inexpensive that its development allows for a normal technology commercialization path.

  17. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  18. Forthcoming Break-Even Conditions of Tokamak Plasma Performance for Fusion Energy Development

    Science.gov (United States)

    Hiwatari, Ryoji; Okano, Kunihiko; Asaoka, Yoshiyuki; Tokimatsu, Koji; Konishi, Satoshi; Ogawa, Yuichi

    The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value ΒN), confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fnGW), the electric break-even condition requires the simultaneous achievement of 1.2 market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of Btmax = 16 T, ηe = 40 %, plant availability 60 %, and a radial build with/without CS coil, the economic break-even condition requires ΒN ˜ 5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with ΒN ˜ 3.0 in the ITER project leads to the upper region of the break-even price in the present world energy scenario, which implies that it is necessary to improve the plasma performance beyond that of the ITER advanced plasma operation.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    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

  1. Improved core-edge tokamak transport simulations with the CORSICA 2 code

    International Nuclear Information System (INIS)

    Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

    1996-01-01

    The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

  2. Reaction-rate coefficients, high-energy ions slowing-down, and power balance in a tokamak fusion reactor plasma

    International Nuclear Information System (INIS)

    Tone, Tatsuzo

    1978-07-01

    Described are the reactivity coefficient of D-T fusion reaction, slowing-down processes of deuterons injected with high energy and 3.52 MeV alpha particles generated in D-T reaction, and the power balance in a Tokamak reactor plasma. Most of the results were obtained in the first preliminary design of JAERI Experimental Fusion Reactor (JXFR) driven with stationary neutral beam injection. A manual of numerical computation program ''BALTOK'' developed for the calculations is given in the appendix. (auth.)

  3. Liquid Metals as Plasma-facing Materials for Fusion Energy Systems: From Atoms to Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stone, Howard A. [Princeton Univ., NJ (United States); Koel, Bruce E. [Princeton Univ., NJ (United States); Bernasek, Steven L. [Princeton Univ., NJ (United States); Carter, Emily A. [Princeton Univ., NJ (United States); Debenedetti, Pablo G. [Princeton Univ., NJ (United States); Panagiotopoulos, Athanassios Z. [Princeton Univ., NJ (United States)

    2017-06-23

    The objective of our studies was to advance our fundamental understanding of liquid metals as plasma-facing materials for fusion energy systems, with a broad scope: from atoms to tokamaks. The flow of liquid metals offers solutions to significant problems of the plasma-facing materials for fusion energy systems. Candidate metals include lithium, tin, gallium, and their eutectic combinations. However, such liquid metal solutions can only be designed efficiently if a range of scientific and engineering issues are resolved that require advances in fundamental fluid dynamics, materials science and surface science. In our research we investigated a range of significant and timely problems relevant to current and proposed engineering designs for fusion reactors, including high-heat flux configurations that are being considered by leading fusion energy groups world-wide. Using experimental and theoretical tools spanning atomistic to continuum descriptions of liquid metals, and bridging surface chemistry, wetting/dewetting and flow, our research has advanced the science and engineering of fusion energy materials and systems. Specifically, we developed a combined experimental and theoretical program to investigate flows of liquid metals in fusion-relevant geometries, including equilibrium and stability of thin-film flows, e.g. wetting and dewetting, effects of electromagnetic and thermocapillary fields on liquid metal thin-film flows, and how chemical interactions and the properties of the surface are influenced by impurities and in turn affect the surface wetting characteristics, the surface tension, and its gradients. Because high-heat flux configurations produce evaporation and sputtering, which forces rearrangement of the liquid, and any dewetting exposes the substrate to damage from the plasma, our studies addressed such evaporatively driven liquid flows and measured and simulated properties of the different bulk phases and material interfaces. The range of our studies

  4. Interplay between edge and outer core fluctuations in the tokamak Tore Supra

    International Nuclear Information System (INIS)

    Fenzi, C.; Garbet, X.; Capes, H.; Devynck, P.; Laviron, C.; Truc, A.; Gervais, F.; Hennequin, P.; Quemeneur, A.

    2000-01-01

    In the tokamak Tore Supra, when a poloidally and toroidally localized limiter, called a modular limiter, is introduced into the lower part of the scrape-off layer, density fluctuations located in the vicinity of this limiter present a specific feature with the appearance of a new spectral pattern in the associated frequency spectrum. This leads to a strong up-down asymmetry observed in both the plasma edge and the plasma outer core, with a maximum level of turbulence at the bottom of the plasma. The observed asymmetry characteristics show that magnetic connection lengths play a critical role here and that the limiter configuration has some effect on the outer core turbulence. (author)

  5. Forthcoming break-even conditions of tokamak plasma performance for fusion energy development

    International Nuclear Information System (INIS)

    Hiwatari, Ryoji; Okano, Kunihiko; Asaoka, Yoshiyuki; Tokimatsu, Koji; Konishi, Satoshi; Ogawa, Yuichi

    2005-01-01

    The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value β N , confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fn GW ), the electric break-even condition requires the simultaneous achievement of 1.2 N GW tmax =16 T, thermal efficiency η e =30%, and current drive power P NBI N ∼1.8, HH≠1.0, and fn GW ∼0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric break-even condition. The second condition is the economic break-even condition, which is required for fusion energy to be selected as an alternative energy source in the energy market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO 2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of B tmax =16 T, η e =40%, plant availability 60%, and a radial build with/without CS coil, the economic break-even condition requires β N ∼5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with β N ∼3.0 in the ITER project leads to the upper region of the break

  6. Introduction condition of a tokamak fusion power plant as an advanced technology in world energy scenario

    International Nuclear Information System (INIS)

    Hiwatari, R.; Tokimatsu, K.; Asaoka, Y.; Okano, K.; Konishi, S.; Ogawa, Y.

    2005-01-01

    The present study reveals the following two introduction conditions of a tokamak fusion power plant in a long term world energy scenario. The first condition is the electric breakeven condition, which is required for the fusion energy to be recognized as a suitable candidate of an alternative energy source in the long term world energy scenario. As for the plasma performance (normalized beta value β N , confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density limit fn GW ), the electric breakeven condition requires the simultaneous achievement of 1.2 N GW tmax =16 T, thermal efficiency η e =30%, and current drive power P NBI N ∼1.8, HH∼1.0, and fn GW ∼0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric breakeven condition. The second condition is the economic breakeven condition, which is required to be selected as an alternative energy source. By using a long term world energy and environment model, the potential of the fusion energy in the long term world energy scenario is being investigated. Under the constraint of 550 ppm CO 2 concentration in the atmosphere, a breakeven price for introduction of the fusion energy in the year 2050 is estimated from 65mill/kWh to 135mill/kWh, which is considered as the economic breakeven condition in the present study. Under the conditions of B tmax =16T, η e =40%, plant availability 60%, and a radial build with/without CS coil, the economic breakeven condition requires β N ∼2.5 for 135mill/kWh of higher breakeven price case and β N ∼6.0 for 65mill/kWh of lower breakeven price case. Finally, the demonstration of steady state operation with β N ∼3.0 in the ITER project leads to the prospect to achieve the upper region of breakeven price in the world energy scenario. (author)

  7. Optimization of fusion power density in the two-energy-component tokamak reactor

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1974-10-01

    The optimal plasma conditions for maximizing fusion power density P/sub f/ in a beam-driven D--T tokamak reactor (TCT) are considered. Given T/sub e/ = T/sub i/ and fixed total plasma pressure, there is an optimal n/sub e/tau/sub E/ for maximizing P/sub f/, viz. n/sub e/tau/sub E/ = 4 x 10 12 to 2 x 10 13 cm -3 sec for T/sub e/ = 3--15 keV and 200-keV D beams. The corresponding anti GAMMA equals (beam pressure/bulk-plasma pressure) is 0.96 to 0.70. P/sub fmax/ increases as T/sub e/ is reduced and can be an order of magnitude larger than the maximum P/sub f/ of a thermal reactor of the same beta, at any temperature. A lower practical limit to T/sub e/ may be set by requiring a minimum beam power multiplication Q/sub b/. For the purpose of fissile breeding, the minimum Q/sub b/ approximately 0.6, requiring T/sub e/ greater than or equal to 3 keV if Z = 1. The optimal operating conditions of a TCT for obtaining P/sub fmax/ are considerably different from those for enhancing Q/sub b/. Maximizing P/sub f/ requires restricting both T/sub e/ and n/sub e/tau/sub E/, maintaining a bulk plasma markedly enriched in tritium, and spoiling confinement of fusion alphas. Considerable impurity content can be tolerated without seriously degrading P/sub fmax/, and high-Z impurity radiation may be useful for regulating tau/sub E/. (auth)

  8. Rationalization and utilization of double-wall vacuum vessel for tokamak fusion facility

    International Nuclear Information System (INIS)

    Nakahira, Masataka

    2005-09-01

    Vacuum Vessel (VV) of ITER is difficult to apply a non-destructive in-service inspection (ISI) and then new safety concept is needed. Present fabrication standards are not applicable to the VV, because the access is limited to the backside of closure weld of double wall. Fabrication tolerance of VV is ± 5mm even the structure is huge as high as 10m. This accuracy requires a rational method on the estimation of welding deformation. In this report, an inherent safety feature of the tokamak is proved closing up a special characteristic of termination of fusion reaction due to tiny water leak. A rational concept not to require ISI without sacrificing safety is shown based on this result. A partial penetration T-welded joint is proposed to establish a rational fabrication method of double wall. Strength and susceptibility to crevice corrosion is evaluated for this joint and feasibility is confirmed. A rational method of estimation of welding deformation for large and complex structure is proposed and the efficiency is shown by comparing analysis and experimental results of full-scale test. (author)

  9. Simulations of phenomena related to edge transport in tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Konzett, S.

    2011-01-01

    This thesis investigates turbulence in a tokamak fusion plasma using numerical simulations. The fluid turbulence code ATTEMPT, which computes the drift dynamics of ions and electrons in electromagnetic fields, is applied to investigate three physical effects which are motivated by recent experimental findings. The first part shows that the statistics of drift fluid turbulence are largely unaffected by the presence of rational magnetic surfaces for typical edge parameter regimes. The second part contains an analysis of the dependence of correlation lengths on various physical parameters. A systematic approach reveals the impact of plasma parameters - which change in the transition from L to H-mode - on parallel, radial and perpendicular correlation lengths. In the last part of the thesis a new flux surface geometry is implemented in the ATTEMPT code. The modified geometry models the onset of the change in magnetic topology near a magnetic X-point. Computations show that turbulent fluctuations are reduced in an X-point distorted flux surface geometry, and the spectral structure of turbulence is altered substantially. (author) [de

  10. A Michelson interferometer/polarimeter on the Tokamak Fusion Test Reactor (TFTR)

    International Nuclear Information System (INIS)

    Park, H.K.; Mansfield, D.K.; Johnson, L.C.; Ma, C.H.

    1987-01-01

    A multichannel interferometer/polarimeter for the Tokamak Fusion Test Reactor (TFTR) has been developed in order to study the time dependent plasma current density (J/sub p/) and electron density (n/sub e/) profile simultaneously. The goal of the TFTR is demonstration of breakeven via dueuterium and tritium (DT) plasma. In order to be operated and maintained during DT operation phase, the system is designed based on the Michelson geometry which possesses intrinsic standing wave problems. So far, there has been no observable signals due to these standing waves. However, a standing wave resulted from the beam path design to achieve a optimum use of the laser power was found. This standing wave has not prevented initial 10 channel interferometer operation. However, a single channel polarimeter test indicated this standing wave was fatal for Faraday notation measurements. Techniques employing 1/2 wave plates and polarizers have been applied to eliminate this standing wave problem. The completion of 10 channel Faraday rotation measurements may be feasible in the near future

  11. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    International Nuclear Information System (INIS)

    1996-01-01

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in vii DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  12. Developing maintainability for tokamak fusion power systems. Phase I report. Volume I. Study results

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-10-01

    The overall purpose of the study is to identify design features of tokamak fusion power reactors which contribute to the achievement of high levels of maintainability. In this first phase, the principal emphasis is on scheduled maintenance whose frequency is determined by the life of the reactor first wall/blanket. Remote operations are baselined. Five conceptual reactor designs have been analyzed. Each concept is characterized by the size of the replaceable first wall/blanket module--large, intermediate, small--and whether access to the module was from the outside of the reactor, the inside of the reactor or a combination of both. The study results are expressed in terms of availability (scheduled maintenance downtime), the costs of maintenance (capital and recurring) and the percent effect of maintenance on the cost of electricity. During this first phase, the study benefitted significantly by the critical review of the feasibility of maintenance functions and the time-to-perform estimates by numerous persons involved in nuclear maintenance and remote operations

  13. Safety analysis of water cooled components inside the JET thermonuclear fusion tokamak

    International Nuclear Information System (INIS)

    Ageladarakis, P.; O'Dowd, N.; Papastergiou, S.

    1998-04-01

    The transient thermal behaviour of a number of components, installed in the vessel of the world's largest Fusion Tokamak (JET) has been examined with a theoretical model, which simulated normal operational conditions and abnormal scenarios namely: Loss of Coolant Flow; Loss of Torus Vacuum; and combinations. A number of theoretical results related to water and cryogenically cooled devices have been validated by a comprehensive experimental campaign conducted both inside the JET plasma chamber and in a test rig. The performance of water cooled components which may be subjected to boiling or freeze-up risks in case of a Loss of Water Flow event has also been analysed. Time constants of transient temperature changes were determined by the model while protective actions were prescribed in order to safeguard the equipment against associated risks. A completely automatic safety protection system has been designed on the basis of these analyses and implemented in the routine JET operation. During operation of JET the safety code reacted several times within the specified time limits and protected the relevant components during real off-normal events. (author)

  14. Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

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

    International Nuclear Information System (INIS)

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

    1988-09-01

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

  16. Electron temperature profiles in high power neutral-beam-heated TFTR [Tokamak Fusion Test Reactor] plasmas

    International Nuclear Information System (INIS)

    Taylor, G.; Grek, B.; Stauffer, F.J.; Goldston, R.J.; Fredrickson, E.D.; Wieland, R.M.; Zarnstorff, M.C.

    1987-09-01

    In 1986, the maximum neutral beam injection (NBI) power in the Tokamak Fusion Test Reactor (TFTR) was increased to 20 MW, with three beams co-parallel and one counter-parallel to I/sub p/. TFTR was operated over a wide range of plasma parameters; 2.5 19 19 m -3 . Data bases have been constructed with over 600 measured electron temperature profiles from multipoint TV Thomson scattering which span much of this parameter space. We have also examined electron temperature profile shapes from electron cyclotron emission at the fundamental ordinary mode and second harmonic extraordinary mode for a subset of these discharges. In the light of recent work on ''profile consistency'' we have analyzed these temperature profiles in the range 0.3 < (r/a) < 0.9 to determine if a profile shape exists which is insensitive to q/sub cyl/ and beam-heating profile. Data from both sides of the temperature profile [T/sub e/(R)] were mapped to magnetic flux surfaces [T/sub e/(r/a)]. Although T/sub e/(r/a), in the region where 0.3 < r/a < 0.9 was found to be slightly broader at lower q/sub cyl/, it was found to be remarkably insensitive to β/sub p/, to the fraction of NBI power injected co-parallel to I/sub p/, and to the heating profile going from peaked on axis, to hollow. 10 refs., 8 figs

  17. Design of deuterium and tritium pellet injector systems for Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Wysor, R.B.; Baylor, L.R.; Bryan, W.E.

    1985-01-01

    Three pellet injector designs developed by the Oak Ridge National Laboratory (ORNL) are planned for the Tokamak Fusion Test Reactor (TFTR) to reach the goal of a tritium pellet injector by 1988. These are the Repeating Pneumatic Injector (RPI), the Deuterium Pellet Injector (DPI) and the Tritium Pellet Injector (TPI). Each of the pellet injector designs have similar performance characteristics in that they deliver up to 4-mm-dia pellets at velocities up to 1500 m/s with a dsign goal to 2000 m/s. Similar techniques are utilized to freeze and extrude the pellet material. The injector systems incorporate three gun concepts which differ in the number of gun barrels and the method of forming and chambering the pellets. The RPI, a single barrel repeating design, has been operational on TFTR since April 1985. Fabrication and assembly are essentially complete for DPI, and TPI is presently on hold after completing about 80% of the design. The TFTR pellet injector program is described, and each of the injector systems is described briefly. Design details are discussed in other papers at this symposium

  18. Cryosorption of helium on argon frost in Tokamak Fusion Test Reactor neutral beamlines

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Cropper, M.B.; Dylla, H.F.; Garzotto, V.; Dudek, L.E.; Grisham, L.R.; Martin, G.D.; O'Connor, T.E.; Stevenson, T.N.; von Halle, A.; Williams, M.D.; Kim, J.

    1990-01-01

    Helium pumping on argon frost has been investigated on Tokamak Fusion Test Reactor (TFTR) neutral beam injectors and shown to be viable for limited helium beam operation. Maximum pumping speeds are ∼25% less than those measured for pumping of deuterium. Helium pumping efficiency is low, >20 argon atoms are required to pump each helium atom. Adsorption isotherms are exponential and exhibit a twofold increase in adsorption capacity as the cryopanel temperature is reduced from 4.3 K to 3.7 K. Pumping speed was found to be independent of cryopanel temperature over the temperature range studied. After pumping a total of 2000 Torr l of helium, the beamline base pressure rose to 2x10 -5 Torr from an initial value of 10 -8 Torr. Accompanying this three order of magnitude increase in pressure was a modest 40% decrease in pumping speed. The introduction of 168 Torr l of deuterium prior to helium injection reduced the pumping speed by a factor of two with no decrease in adsorption capacity

  19. Final Report for the "Fusion Application for Core-Edge Transport Simulations (FACETS)"

    Energy Technology Data Exchange (ETDEWEB)

    Cary, John R; Kruger, Scott

    2014-10-02

    The FACETS project over its lifetime developed the first self-consistent core-edge coupled capabilities, a new transport solver for modeling core transport in tokamak cores, developed a new code for modeling wall physics over long time scales, and significantly improved the capabilities and performance of legacy components, UEDGE, NUBEAM, GLF23, GYRO, and BOUT++. These improved capabilities leveraged the team’s expertise in applied mathematics (solvers and algorithms) and computer science (performance improvements and language interoperability). The project pioneered new methods for tackling the complexity of simulating the concomitant complexity of tokamak experiments.

  20. Self-consistent kinetic simulations of lower hybrid drift instability resulting in electron current driven by fusion products in tokamak plasmas

    International Nuclear Information System (INIS)

    Cook, J W S; Chapman, S C; Dendy, R O; Brady, C S

    2011-01-01

    We present particle-in-cell (PIC) simulations of minority energetic protons in deuterium plasmas, which demonstrate a collective instability responsible for emission near the lower hybrid frequency and its harmonics. The simulations capture the lower hybrid drift instability in a parameter regime motivated by tokamak fusion plasma conditions, and show further that the excited electromagnetic fields collectively and collisionlessly couple free energy from the protons to directed electron motion. This results in an asymmetric tail antiparallel to the magnetic field. We focus on obliquely propagating modes excited by energetic ions, whose ring-beam distribution is motivated by population inversions related to ion cyclotron emission, in a background plasma with a temperature similar to that of the core of a large tokamak plasma. A fully self-consistent electromagnetic relativistic PIC code representing all vector field quantities and particle velocities in three dimensions as functions of a single spatial dimension is used to model this situation, by evolving the initial antiparallel travelling ring-beam distribution of 3 MeV protons in a background 10 keV Maxwellian deuterium plasma with realistic ion-electron mass ratio. These simulations provide a proof-of-principle for a key plasma physics process that may be exploited in future alpha channelling scenarios for magnetically confined burning plasmas.

  1. Edge-core interaction of ITG turbulence in Tokamaks: Is the Tail Wagging the Dog?

    Science.gov (United States)

    Ku, S.; Chang, C. S.; Dif-Pradalier, G.; Diamond, P. H.

    2010-11-01

    A full-f XGC1 gyrokinetic simulation of ITG turbulence, together with the neoclassical dynamics without scale separation, has been performed for the whole-volume plasma in realistic diverted DIII-D geometry. The simulation revealed that the global structure of the turbulence and transport in tokamak plasmas results from a synergy between edge-driven inward propagation of turbulence intensity and the core-driven outward heat transport. The global ion confinement and the ion temperature gradient then self-organize quickly at turbulence propagation time scale. This synergy results in inward-outward pulse scattering leading to spontaneous production of strong internal shear layers in which the turbulent transport is almost suppressed over several radial correlation lengths. Co-existence of the edge turbulence source and the strong internal shear layer leads to radially increasing turbulence intensity and ion thermal transport profiles.

  2. The role of the neutral beam fueling profile in the performance of the Tokamak Fusion Test Reactor and other tokamak plasmas

    International Nuclear Information System (INIS)

    Park, H.K.; Batha, S.

    1997-02-01

    Scalings for the stored energy and neutron yield, determined from experimental data are applied to both deuterium-only and deuterium-tritium plasmas in different neutral beam heated operational domains in Tokamak Fusion Test Reactor. The domain of the data considered includes the Supershot, High poloidal beta, Low-mode, and limiter High-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and Ion Temperature Gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

  3. Analysis of the plasma magnetohydrodynamic equilibrium in iron core transformer Tokamak HL-1M

    International Nuclear Information System (INIS)

    Chen Xiaoguang; Yuan Baoshan

    1992-01-01

    The physical and mathematical model are presented on the problem of MHD equilibrium with the self consistent in iron core transformer HL-1M. Calculation and analysis for the plasma equilibrium of the stable boundary and free boundary are shown respectively, in an axisymmetric equilibrium model of two dimensions. First, a variation formulation of the problem is written and the equations of the poloided flux ψ are solved by a finite element method; the Picard and Newton algorithms are tested for the non-linearities. The plasma boundary and the magnetic surfaces are being simulated, with the currents in the coils, the total plasma current, its current density function and the magnetic permeability of the iron being the data for the problem; a certain number of the characteristic parameter of the equilibrium configuration is calculated. Secondly, a simple method of calculation is adopted in the determination of equilibrium fields and currents in iron core HL-1M tokamak device. In the plasma equilibrium, the magnetic effect of the air gaps in the iron core and the iron magnetic shielded plate are considered in HL-1M device. Reliable data are provided for designing and constructing the poloidal field system of HL-1M device. A good computer code is constructed, which may be useful in operating on analysis for the future device

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  5. Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

    International Nuclear Information System (INIS)

    Paris, P.J.; Yassen, F.; Assis, A.S. de; Raposo, C.

    1988-07-01

    The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.) [pt

  6. Gas utilization in the Tokamak Fusion Test Reactor neutral beam injectors

    International Nuclear Information System (INIS)

    Kamperschroer, J.H.; Gammel, G.M.; Kugel, H.W.; Grisham, L.R.; Stevenson, T.N.; von Halle, A.; Williams, M.D.; Jones, T.T.C.

    1989-01-01

    Measurements of gas utilization were performed using hydrogen and deuterium beams in the Tokamak Fusion Test Reactor (TFTR) neutral beam test beamline to study the feasibility of operating tritium beams with existing ion sources under conditions of minimal tritium consumption. (i) It was found that the fraction of gas molecules introduced into the TFTR long-pulse ion sources that are converted to extracted ions (i.e., the ion source gas efficiency) was higher than with previous short-pulse sources. Gas efficiencies were studied over the range 33%--55%, and its effect on neutralization of the extracted ions was studied. At the high end of the gas efficiency range, the neutral fraction of the beam fell below that predicted from room-temperature molecular gas flow (similar to observations at the Joint European Torus). (ii) Beam isotope change studies were performed. No extracted hydrogen ions were observed in the first deuterium beam following a working gas change from H 2 to D 2 . There was no arc conditioning or gas injection preceding the first beam extraction attempt. (iii) Experiments were also performed to determine the reliability of ion source operation during the long waiting periods between pulses that are anticipated during tritium operation. It was found that an ion source conditioned to 120 kV could produce a clean beam pulse after a waiting period of 14 h by preceding the beam extraction with several acceleration voltage/filament warm-up pulses. It can be concluded that the operation of up to six ion sources on tritium gas should be compatible with on-site inventory restrictions established for D--T, Q = 1 experiments on TFTR

  7. Stress analysis of neutral beam pivot point bellows for tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Johnson, J.J.; Benda, B.J.; Tiong, L.W.

    1983-01-01

    The neutral beam pivot point bellows serves as an airtight flexible linkage between the torus duct and the neutral beam transition duct in Princeton University's Tokamak Fusion Test Reactor. The bellows considered here is basically rectangular in cross section with rounded corners; a unique shape. Its overall external dimensions are about 28 in. (about 711 mm) X about 35 in. (about 889 mm). The bellows is formed from 18 convolutions and is of the nested ripple type. It is about 11 in. (about 43.3 mm) in length, composed of Inconel 718, and each leaf has a thickness of 0.034 in. (.86 mm). The bellows is subjected to a series of design loading conditions -- vacuum, vacuum + 2 psi (.12 MPa), vacuum + stroke (10,000 cycles), vacuum + temperature increase + extension, extension to a stress of 120 ksi (838 MPa), and a series of rotational loading conditions induced in the bellows by alignment of the neutral beam injector. A stress analysis of the bellows was performed by the finite element method -- locations and magnitude of maximum stresses were calculated for all of the design loading conditions to check with allowable values and help guide placement of strain gauges during proof testing. A typical center convolution and end convolution were analyzed. Loading conditions were separated into symmetric and antisymmetric cases about the planes of symmetry of the cross-section. Iterative linear analyses were performed, i.e. compressive loading conditions led to predicted overlap of the leaves from linear analysis and restraints were added to prevent such overlap. This effect was found to be substantial in stress predicition and necessary to be taken into account. A total of eleven loading conditions and seven models were analyzed. The results showed peak stresses to be within allowable limits and the number of allowable cycles to be greater than the design condition

  8. Study of lower hybrid current drive system in tokamak fusion devices

    International Nuclear Information System (INIS)

    Maebara, Sunao

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 μP is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 μsec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10 -6 Pam 3 /sm 2 at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm -2 (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10 -5 Pam 3 /sm 2 which is low enough for an antenna material. (author)

  9. End points in discharge cleaning on TFTR (Tokamak Fusion Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, D.; Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Bush, C.E.; Gettelfinger, G.; Hawryluk, R.J.; Hill, K.W.; Janos, A.C.; Jobes, F.C.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) /approx/ 3 at /approx/ 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H/sub 2/O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after /approx/ 10/sup 5/ pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150/degree/C to about 250/degree/C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H/sub 2/O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab.

  10. End points in discharge cleaning on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Mueller, D.; Dylla, H.F.; Bell, M.G.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) approx 3 at approx 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H 2 O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after approx 10 5 pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 degree C to about 250 degree C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H 2 O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab

  11. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

    1991-01-01

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    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

  13. Fusion energy

    International Nuclear Information System (INIS)

    Gross, R.A.

    1984-01-01

    This textbook covers the physics and technology upon which future fusion power reactors will be based. It reviews the history of fusion, reaction physics, plasma physics, heating, and confinement. Descriptions of commercial plants and design concepts are included. Topics covered include: fusion reactions and fuel resources; reaction rates; ignition, and confinement; basic plasma directory; Tokamak confinement physics; fusion technology; STARFIRE: A commercial Tokamak fusion power plant. MARS: A tandem-mirror fusion power plant; and other fusion reactor concepts

  14. Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

    International Nuclear Information System (INIS)

    Diop, C.M'B.

    1984-01-01

    This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

  15. Culham Conceptual Tokamak Mark II. Design study of the layout of a twin-reactor fusion power station

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.; Harding, N.H.

    1981-07-01

    This report describes the building layout and outline design for the nuclear complex of a fusion reactor power station incorporating two Culham Conceptual Tokamak Reactors Mk.II. The design incorporates equipment for steam generation, process services for the fusion reactors and all facilities for routine and non-routine servicing of the nuclear complex. The design includes provision of temporary facilities for on site construction of the major reactor components and shows that these facilities may be used for disassembly of the reactors either for major repair and/or decommissioning. Preliminary estimates are included, which indicate the cost benefits to be obtained from incorporating two reactors in one nuclear complex and from increased wall loading. (author)

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

    International Nuclear Information System (INIS)

    Intrator, T.

    2000-01-01

    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''

  17. Study of lower hybrid current drive system in tokamak fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Maebara, Sunao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 {mu}P is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 {mu}sec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10{sup -6} Pam{sup 3}/sm{sup 2} at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm{sup -2} (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10{sup -5} Pam{sup 3}/sm{sup 2} which is low enough for an antenna material. (author)

  18. A study on tokamak fusion reactor - Numerical analyses of MHD equilibrium= and edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Lim, Ki Hang; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Duk Kyu [Seoul National University, Seoul (Korea, Republic of); Cho, Soo Won [Kyungki Unviersity, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-DMHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we compute the plasma equilibrium of double null type and calculate the external coil currents and the plasma parameters used for operation and control data. Also the numerical algorithm is developed to analyse the behavior of edge plasmas in poloidal and radial directions and the programming and debugging of a 2-D transport code are completed. Furthermore, a neutral particle transport code for the edge region is developed and then used for the analysis of the neutral transport phenomena giving the sources in the fluid equations, and expected to supply the input parameters for the edge plasma transport code. 34 refs., 5 tabs., 28 figs. (author)

  19. The simultaneous measurements of core and outer core density fluctuations in L-H transition using CO2 laser collective scattering diagnostic in the EAST superconducting tokamak

    International Nuclear Information System (INIS)

    Cao, G.M.; Li, Y.D.; Zhang, X.D.; Sun, P.J.; Hu, L.Q.; Li, J.G.; Wu, G.J.

    2013-01-01

    The H-mode is the projected basic operation scenario for the ITER tokamak. The turbulence de-correlation by the synergistic effect of zonal flow and equilibrium ExB flow shear is believed to be the reason for L-H transition, however, the detailed physical mechanism has not been identified so far. Tangential multi-channel CO 2 laser collective scattering diagnostic system (mainly k r measurement) was first installed to investigate electron density fluctuations on EAST tokamak. The measurements in a spontaneous dithering L-H transition show that in core plasma (0 < r/a < 0.5) the low-frequency fluctuations strengthen greatly before L-H transition; meanwhile in outer core plasma (0.2 < r/a < 1) the low-frequency fluctuations strengthen slightly. Bispectral analysis reveals that the coupling strength between low- and high-frequency fluctuations in both core and outer core plasma strengthens greatly before the transition, but the latter is greater than the former. The results indicate that the low-frequency fluctuations of the core and outer core plasma play active, but different, roles in the spontaneous L-H transition. (author)

  20. Ion cyclotron emission due to the newly-born fusion products induced fast Alfven wave radiative instabilities in tokamaks

    International Nuclear Information System (INIS)

    Arunasalam, V.

    1995-08-01

    The velocity distribution functions of the newly born (t = 0) charged fusion products of tokamak discharges can be approximated by a monoenergetic ring distribution with a finite v parallel such that v perpendicular ∼ v parallel ∼ v j where (M j V j 2 /2) = E j , the directed birth energy of the charged fusion product species j of mass M j . As the time t progresses these distribution functions will evolve into a Gaussian in velocity with thermal spreadings given by the perpendicular and parallel temperatures T perpendicularj (t) = T parallelj (t) with T j (t) increasing as t increases and finally reaches an isotropic saturation value of T perpendicularj (t ∼ τ j ) = T parallelj (t ∼ τ j ) = T j (t ∼ τ j ) ∼ [M j T d E j /(M j + M)] 1/2 , where T d is the temperature of the background deuterium plasma ions, M is the mass of a triton or a neutron for j = protons and alpha particles, respectively, and τ j ∼ τ sj /4 is the thermalization time of the fusion product species j in the background deuterium plasma and τ sj is the slowing-down time. For times t of the order of τ j their distributions can be approximated by a Gaussian in their total energy. Then for times t ≥ τ sj the velocity distributions of these fusion products will relax towards their appropriate slowing-down distributions. Here the authors will examine the radiative stability of all these distributions. The ion cyclotron emission from energetic ion produced by fusion reactions or neutral beam injection promises to be a useful diagnostic tool

  1. Characterization of HCoV-229E fusion core: Implications for structure basis of coronavirus membrane fusion

    International Nuclear Information System (INIS)

    Liu Cheng; Feng Youjun; Gao Feng; Zhang Qiangmin; Wang Ming

    2006-01-01

    Human coronavirus 229E (HCoV-229E), a member of group I coronaviruses, has been identified as one of the major viral agents causing respiratory tract diseases in humans for nearly 40 years. However, the detailed molecular mechanism of the membrane fusion mediated by the spike (S) protein of HCoV-229E remains elusive. Here, we report, for the first time, a rationally designed fusion core of HCoV-229E (HR1-SGGRGG-HR2), which was in vitro produced in GST prokaryotic expression system. Multiple lines of experimental data including gel-filtration, chemical cross-linking, and circular diagram (CD) demonstrated that the HCoV-229E fusion core possesses the typical properties of the trimer of coiled-coil heterodimer (six α-helix bundle). 3D structure modeling presents its most-likely structure, similar to those of coronaviruses that have been well-documented. Collectively, HCoV-229E S protein belongs to the type I fusion protein, which is characterized by the existence of two heptad-repeat regions (HR1 and HR2), furthermore, the available knowledge concerning HCoV-229E fusion core may make it possible to design small molecule or polypeptide drugs targeting the membrane fusion, a crucial step of HCoV-229E infection

  2. Influence of INCONEL 625 composition on the activation characteristics of the vacuum vessel of experimental fusion tokamaks

    International Nuclear Information System (INIS)

    Cambi, G.; Cepraga, D.G.; Boeriu, S.; Maganzani, I.

    1995-01-01

    The radioactive inventory, the decay heat and the contact dose rate of permanent components such as the vacuum vessel of two experimental fusion tokamaks, the compact IGNITOR-ULT and the ITER-EDA fusion machines, are evaluated by using the ENEA-Bologna integrated methodology. The vacuum vessel material considered is the INCONEL 625. The neutron flux is calculated using the VITAMIN-C 171-group library, based on EFF-2 data and the 1-D transport code XSDRNPM in the S 8 -P 3 approximation. The ANITA-2 code, using updated cross sections and decay data libraries based on EAF-3 and IRDF90 evaluation files is used for activation calculations. The fusion neutron source has been normalised to a neutron first wall load of 2 MW/m 2 and 1 MW/m 2 for IGNITOR-ULT and ITER, respectively. The material irradiation have been described by multistep time histories, resulting in the designed total fluence. Variations in the composition of INCONEL 625 have been assessed and their impact on the activation characteristics are discussed, also from the point of view of waste disposal. (orig.)

  3. Computer simulation of charged fusion-product trajectories and detection efficiency expected for future experiments within the COMPASS tokamak

    International Nuclear Information System (INIS)

    Kwiatkowski, Roch; Malinowski, Karol; Sadowski, Marek J

    2014-01-01

    This paper presents results of computer simulations of charged particle motions and detection efficiencies for an ion-pinhole camera of a new diagnostic system to be used in future COMPASS tokamak experiments. A probe equipped with a nuclear track detector can deliver information about charged products of fusion reactions. The calculations were performed with a so-called Gourdon code, based on a single-particle model and toroidal symmetry. There were computed trajectories of fast ions (> 500 keV) in medium-dense plasma (n e  < 10 14  cm −3 ) and an expected detection efficiency (a ratio of the number of detected particles to that of particles emitted from plasma). The simulations showed that charged fusion products can reach the new diagnostic probe, and the expected detection efficiency can reach 2 × 10 −8 . Based on such calculations, one can determine the optimal position and orientation of the probe. The obtained results are of importance for the interpretation of fusion-product images to be recorded in future COMPASS experiments. (paper)

  4. A step towards controlled fusion reactors: Tore Supra tokamak with superconducting magnets

    International Nuclear Information System (INIS)

    Turck, B.

    1988-01-01

    Tore Supra technology has to solve all the problems related to the development and the installaion of superconducting coils and associated cryogenic devices. Tore Supra will allow to get a significative experience to prepare next machines. Specifications and needs of tokamaks concerning the superconducting coils of future machines are recalled [fr

  5. Fusion Energy-Production from a Deuterium-Tritium Plasma in the Jet Tokamak

    NARCIS (Netherlands)

    Rebut, P. H.; Gibson, A.; Huguet, M.; Adams, J. M.; Alper, B.; Altmann, H.; Andersen, A.; Andrew, P.; Angelone, M.; Aliarshad, S.; Baigger, P.; Bailey, W.; Balet, B.; Barabaschi, P.; Barker, P.; Barnsley, R.; Baronian, M.; Bartlett, D. V.; Baylor, L.; Bell, A. C.; Benali, G.; Bertoldi, P.; Bertolini, E.; Bhatnagar, V.; Bickley, A. J.; Binder, D.; Bindslev, H.; Bonicelli, T.; Booth, S. J.; Bosia, G.; Botman, M.; Boucher, D.; Boucquey, P.; Breger, P.; Brelen, H.; Brinkschulte, H.; Brooks, D.; Brown, A.; Brown, T.; Brusati, M.; Bryan, S.; Brzozowski, J.; Buchse, R.; Budd, T.; Bures, M.; Businaro, T.; Butcher, P.; Buttgereit, H.; Caldwellnichols, C.; Campbell, D. J.; Card, P.; Celentano, G.; Challis, C. D.; Chankin, A. V.; Cherubini, A.; Chiron, D.; Christiansen, J.; Chuilon, P.; Claesen, R.; Clement, S.; Clipsham, E.; Coad, J. P.; Coffey, I. H.; Colton, A.; Comiskey, M.; Conroy, S.; Cooke, M.; Cooper, D.; Cooper, S.; Cordey, J. G.; Core, W.; Corrigan, G.; Corti, S.; Costley, A. E.; Cottrell, G.; Cox, M.; Cripwell, P.; Dacosta, O.; Davies, J.; Davies, N.; de Blank, H.; De Esch, H.; Dekock, L.; Deksnis, E.; Delvart, F.; Dennehinnov, G. B.; Deschamps, G.; Dickson, W. J.; Dietz, K. J.; Dmitrenko, S. L.; Dmitrieva, M.; Dobbing, J.; Doglio, A.; Dolgetta, N.; Dorling, S. E.; Doyle, P. G.; Duchs, D. F.; Duquenoy, H.; Edwards, A.; Ehrenberg, J.; Ekedahl, A.; Elevant, T.; Erents, S.K.; Eriksson, L. G.; Fajemirokun, H.; Falter, H.; Freiling, J.; Freville, F.; Froger, C.; Froissard, P.; Fullard, K.; Gadeberg, M.; Galetsas, A.; Gallagher, T.; Gambier, D.; Garribba, M.; Gaze, P.; Giannella, R.; Gill, R. D.; Girard, A.; Gondhalekar, A.; Goodall, D.; Gormezano, C.; Gottardi, N. A.; Gowers, C.; Green, B. J.; Grievson, B.; Haange, R.; Haigh, A.; Hancock, C. J.; Harbour, P. J.; Hartrampf, T.; Hawkes, N. C.; Haynes, P.; Hemmerich, J. L.; Hender, T.; Hoekzema, J.; Holland, D.; Hone, M.; Horton, L.; How, J.; Huart, M.; Hughes, I.; Hughes, T. P.; Hugon, M.; Huo, Y.; Ida, K.; Ingram, B.; Irving, M.; Jacquinot, J.; Jaeckel, H.; Jaeger, J. F.; Janeschitz, G.; Jankovicz, Z.; Jarvis, O. N.; Jensen, F.; Jones, E. M.; Jones, H. D.; Jones, Lpdf; Jones, S.; Jones, T. T. C.; Junger, J. F.; Junique, F.; Kaye, A.; Keen, B. E.; Keilhacker, M.; Kelly, G. J.; Kerner, W.; Khudoleev, A.; Konig, R.; Konstantellos, A.; Kovanen, M.; Kramer, G.; Kupschus, P.; Lasser, R.; Last, J. R.; Laundy, B.; Laurotaroni, L.; Laveyry, M.; Lawson, K.; Lennholm, M.; Lingertat, J.; Litunovski, R. N.; Loarte, A.; Lobel, R.; Lomas, P.; Loughlin, M.; Lowry, C.; Lupo, J.; Maas, A. C.; Machuzak, J.; Macklin, B.; Maddison, G.; Maggi, C. F.; Magyar, G.; Mandl, W.; Marchese, V.; Marcon, G.; Marcus, F.; Mart, J.; Martin, D.; Martin, E.; Martinsolis, R.; Massmann, P.; Matthews, G.; McBryan, H.; McCracken, G.; McKivitt, J.; Meriguet, P.; Miele, P.; Miller, A.; Mills, J.; Mills, S. F.; Millward, P.; Milverton, P.; Minardi, E.; Mohanti, R.; Mondino, P. L.; Montgomery, D.; Montvai, A.; Morgan, P.; Morsi, H.; Muir, D.; Murphy, G.; Myrnas, R.; Nave, F.; Newbert, G.; Newman, M.; Nielsen, P.; Noll, P.; Obert, W.; Obrien, D.; Orchard, J.; Orourke, J.; Ostrom, R.; Ottaviani, M.; Pain, M.; Paoletti, F.; Papastergiou, S.; Parsons, W.; Pasini, D.; Patel, D.; Peacock, A.; Peacock, N.; Pearce, R. J. M.; Pearson, D.; Peng, J. F.; Desilva, R. P.; Perinic, G.; Perry, C.; Petrov, M.; Pick, M. A.; Plancoulaine, J.; Poffe, J. P.; Pohlchen, R.; Porcelli, F.; Porte, L.; Prentice, R.; Puppin, S.; Putvinskii, S.; Radford, G.; Raimondi, T.; Deandrade, M. C. R.; Reichle, R.; Reid, J.; Richards, S.; Righi, E.; Rimini, F.; Robinson, D.; Rolfe, A.; Ross, R. T.; Rossi, L.; Russ, R.; Rutter, P.; Sack, H. C.; Sadler, G.; Saibene, G.; Salanave, J. L.; Sanazzaro, G.; Santagiustina, A.; Sartori, R.; Sborchia, C.; Schild, P.; Schmid, M.; Schmidt, G.; Schunke, B.; Scott, S. M.; Serio, L.; Sibley, A.; Simonini, R.; Sips, A.C.C.; Smeulders, P.; Smith, R.; Stagg, R.; Stamp, M.; Stangeby, P.; Stankiewicz, R.; Start, D. F.; Steed, C. A.; Stork, D.; Stott, P.E.; Stubberfield, P.; Summers, D.; Summers, H.; Svensson, L.; Tagle, J. A.; Talbot, M.; Tanga, A.; Taroni, A.; Terella, C.; Terrington, A.; Tesini, A.; Thomas, P. R.; Thompson, E.; Thomsen, K.; Tibone, F.; Tiscornia, A.; Trevalion, P.; Tubbing, B.; Vanbelle, P.; Vanderbeken, H.; Vlases, G.; von Hellermann, M.; Wade, T.; Walker, C.; Walton, R.; Ward, D.; Watkins, M. L.; Watkins, N.; Watson, M. J.; Weber, S.; Wesson, J.; Wijnands, T. J.; Wilks, J.; Wilson, D.; Winkel, T.; Wolf, R.; Wong, D.; Woodward, C.; Wu, Y.; Wykes, M.; Young, D.; Young, I. D.; Zannelli, L.; Zolfaghari, A.; Zwingmann, W.

    1992-01-01

    The paper describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixtures. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to

  6. Studies of tokamak fusion reactor dynamics. Progress report, June 1, 1975--February 15, 1976

    International Nuclear Information System (INIS)

    Mills, R.G.; Gralnick, S.L.

    1976-01-01

    An investigation of the effect of plasma shape and position on the inductive coupling between the plasma and the external poloidal field coils is briefly described. Research on a multi-node time-dependent point kinetics code with which to study the operating dynamics of a tokamak reactor is also mentioned

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

    International Nuclear Information System (INIS)

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

    1989-05-01

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

  8. Investigating fusion plasma instabilities in the Mega Amp Spherical Tokamak using mega electron volt proton emissions (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Perez, R. V., E-mail: rvale006@fiu.edu; Boeglin, W. U.; Angulo, A.; Avila, P.; Leon, O.; Lopez, C. [Department of Physics, Florida International University, 11200 SW 8 ST, CP204, Miami, Florida 33199 (United States); Darrow, D. S. [Princeton Plasma Physics Laboratory, James Forrestal Campus, P.O. Box 451, Princeton, New Jersey 08543 (United States); Cecconello, M.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, Uppsala SE-751 20 (Sweden); Allan, S. Y.; Akers, R. J.; Keeling, D. L.; McClements, K. G.; Scannell, R.; Conway, N. J. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrasse 2, D-85748, Garching (Germany); Jones, O. M. [CCFE, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Michael, C. A. [Australian National University, Canberra ACT 0200 (Australia)

    2014-11-15

    The proton detector (PD) measures 3 MeV proton yield distributions from deuterium-deuterium fusion reactions within the Mega Amp Spherical Tokamak (MAST). The PD’s compact four-channel system of collimated and individually oriented silicon detectors probes different regions of the plasma, detecting protons (with gyro radii large enough to be unconfined) leaving the plasma on curved trajectories during neutral beam injection. From first PD data obtained during plasma operation in 2013, proton production rates (up to several hundred kHz and 1 ms time resolution) during sawtooth events were compared to the corresponding MAST neutron camera data. Fitted proton emission profiles in the poloidal plane demonstrate the capabilities of this new system.

  9. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  10. Alpha-driven magnetohydrodynamics (MHD) and MHD-induced alpha loss in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1997-02-01

    Alpha-driven toroidal Alfven eigenmodes (TAEs) are observed as predicted by theory in the post neutral beam phase in high central q (safety factor) deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR). The mode location, poloidal structure and the importance of q profile for TAE instability are discussed. So far no alpha particle loss due to these modes was detected due to the small mode amplitude. However, alpha loss induced by kinetic ballooning modes (KBMs) was observed in high confinement D-T discharges. Particle orbit simulation demonstrates that the wave-particle resonant interaction can explain the observed correlation between the increase in alpha loss and appearance of multiple high-n (n ≥ 6, n is the toroidal mode number) modes

  11. Finite element and node point generation computer programs used for the design of toroidal field coils in tokamak fusion devices

    International Nuclear Information System (INIS)

    Smith, R.A.

    1975-06-01

    The structural analysis of toroidal field coils in Tokamak fusion machines can be performed with the finite element method. This technique has been employed for design evaluations of toroidal field coils on the Princeton Large Torus (PLT), the Poloidal Diverter Experiment (PDX), and the Tokamak Fusion Test Reactor (TFTR). The application of the finite element method can be simplified with computer programs that are used to generate the input data for the finite element code. There are three areas of data input where significant automation can be provided by supplementary computer codes. These concern the definition of geometry by a node point mesh, the definition of the finite elements from the geometric node points, and the definition of the node point force/displacement boundary conditions. The node point forces in a model of a toroidal field coil are computed from the vector cross product of the coil current and the magnetic field. The computer programs named PDXNODE and ELEMENT are described. The program PDXNODE generates the geometric node points of a finite element model for a toroidal field coil. The program ELEMENT defines the finite elements of the model from the node points and from material property considerations. The program descriptions include input requirements, the output, the program logic, the methods of generating complex geometries with multiple runs, computational time and computer compatibility. The output format of PDXNODE and ELEMENT make them compatible with PDXFORC and two general purpose finite element computer codes: (ANSYS) the Engineering Analysis System written by the Swanson Analysis Systems, Inc., and (WECAN) the Westinghouse Electric Computer Analysis general purpose finite element program. The Fortran listings of PDXNODE and ELEMENT are provided

  12. Fusion Plasma Theory: Task 1, Magnetic confinement Fusion Plasma Theory

    International Nuclear Information System (INIS)

    Callen, J.D.

    1993-01-01

    The research performed under this grant during the current year has concentrated on few tokamak plasma confinement issues: applications of our new Chapman-Enskog-like approach for developing hybrid fluid/kinetic descriptions of tokamak plasmas; multi-faceted studies as part of our development of a new interacting island paradigm for the tokamak equilibrium'' and transport; investigations of the resolution power of BES and ECE diagnostics for measuring core plasma fluctuations; and studies of net transport in the presence of fluctuating surfaces. Recent progress and publications in these areas, and in the management of the NERSC node and the fusion theory workstations are summarized briefly in this report

  13. Joint research using small tokamaks

    International Nuclear Information System (INIS)

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

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  14. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  15. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  16. Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

    International Nuclear Information System (INIS)

    Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.

    2000-01-01

    If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

  17. Effect of eddy currents in the toroidal field coils of a tokamak with an air-core transformer

    International Nuclear Information System (INIS)

    Tani, Keiji; Kobayashi, Tomofumi; Tamura, Sanae

    1975-02-01

    The effect of eddy currents in the copper parts of the toroidal field coils is evaluated for a tokamak with the air-core transformer windings located inside the bore of the toroidal field coils. By introducing appropriate weights to the solutions obtained for a simplified cylindrical model, calculation is made of the induction toroidal electric field on the plasma axis in the presence of the eddy currents. The result shows that, to reduce the influence of the eddy currents on the induction one-turn voltage to the permissible level, it is necessary to choose the optimal number of turns and shape of the single conductor of the toroidal field coil. (auth.)

  18. Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Medley, S.S.

    1986-01-01

    The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10 -3 Torr is required to maintain the analyzer chamber at a pressure of -6 Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of -8 Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

  19. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    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

  20. Resistive requirements for the vacuum wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ehat, D.; Harkness, S.D.; Norem, J.; Stevens, H.; Turner, L.

    1978-01-01

    Most conceptual designs of tokamak power reactors have incorporated a ceramic insulator in the vacuum wall to make the wall electrically non-conducting. Such a material will have to be highly resistant to radiation damage at doses up to at least 10 MW-yr/m 2 while being compatible with a coolant and a first wall whose dimensions change due to thermal cycling and radiation damage. Thus there is considerable incentive to assess the consequences of eliminating the flux breaker from the design and having a conducting boundary instead. In this initial study the question of having a finite wall resistance has been examined in terms of its major implications on both the normal and abnormal operation of a tokamak reactor. This study has been conducted within the framework of the ANL-EPR-77 design although the results should provide some guidance for future reactors as well. The EPR design referred to is a 5 m major radius tokamak with an aspect ratio of 3.5, and with an equilibrium plasma current of 7.3 MA. The vacuum chamber is designed to accommodate a non-circular plasma with a height to width ratio of up to 1.65. The basic vacuum wall design is shown in Fig. 1. It is located about 0.4 M from the plasma boundary and has an irregular polygon shape made of sixteen sections, one per TF coil interval. Variations of this design having a range of resistance values have been used in the analysis

  1. Fokker--Planck/transport analyses of fusion plasmas in contemporary beam-driven tokamaks

    International Nuclear Information System (INIS)

    Mirin, A.A.; McCoy, M.G.; Killeen, J.; Rensink, M.E.; Shumaker, D.E.; Jassby, D.L.; Post, D.E.

    1978-04-01

    The properties of deuterium plasmas in experimental tokamaks heated and fueled by intense neutral-beam injection are evaluated with a Fokker-Planck/radial transport code coupled with a Monte Carlo neutrals treatment. Illustrative results are presented for the Poloidal Divertor Experiment at PPPL as a function of beam power and plasma recycling coefficient, R/sub c/. When P/sub beam/ = 8 MW at E/sub b/ = 60 keV, and R/sub c/ = 0.2, then approximately 0.5, [ 2 / 3 ] = 22 keV approximately 6 , and the D-D neutron intensity is 10 16 n/sec

  2. Observation of inward and outward particle convection in the core of electron cyclotron heated and current driven plasmas in the Tokamak a Configuration Variable

    International Nuclear Information System (INIS)

    Furno, I.; Weisen, H.

    2003-01-01

    In the Tokamak a Configuration Variable [F. Hofmann, J.B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 (1994)], inward or outward convection in the core of electron cyclotron heated and current driven plasmas is observed, depending on discharge conditions. In sawtoothing discharges with central electron cyclotron heating, outward convection is observed when a quasicontinuous m=1 kink mode is present, resulting in inverted sawteeth on the central electron density, while in the absence thereof, inward convection between successive sawtooth crashes leads to 'normal' sawteeth. The occurrence of a kink mode depends sensitively on plasma triangularity. When sawteeth are stabilized with central co- or counterelectron cyclotron current drive, stationary hollow electron density profiles are observed in the presence of m=1 modes, while peaked or flat profiles are observed in magnetohydrodynamic quiescent discharges. The observation of peaked density profiles in fully electron cyclotron driven plasmas demonstrates that pinch processes other than the Ware pinch must be responsible for these phenomena

  3. An advanced conceptual Tokamak fusion power reactor utilizing closed cycle helium gas turbines

    International Nuclear Information System (INIS)

    Conn, R.W.

    1976-01-01

    UWMAK-III is a conceptual Tokamak reactor designed to study the potential and the problems associated with an advanced version of Tokamaks as power reactors. Design choices have been made which represent reasonable extrapolations of present technology. The major features are the noncircular plasma cross section, the use of TZM, a molybdenum based alloy, as the primary structural material, and the incorporation of a closed-cycle helium gas turbine power conversion system. A conceptual design of the turbomachinery is given together with a preliminary heat exchanger analysis that results in relatively compact designs for the generator, precooler, and intercooler. This paper contains a general description of the UWMAK-III system and a discussion of those aspects of the reactor, such as the burn cycle, the blanket design and the heat transfer analysis, which are required to form the basis for discussing the power conversion system. The authors concentrate on the power conversion system and include a parametric performance analysis, an interface and trade-off study and a description of the reference conceptual design of the closed-cycle helium gas turbine power conversion system. (Auth.)

  4. Optimization of OH coil recharging scenario of quasi-steady operation in tokamak fusion reactor by lower hybrid wave current drive

    International Nuclear Information System (INIS)

    Sugihara, M.; Fujisawa, N.; Nishio, S.; Iida, H.

    1984-01-01

    Using simple physical model equations optimum plasma and rf parameters for an OH coil recharging scenario of quasi-steady operation in tokamak fusion reactors by lower hybrid wave current drive are studied. In this operation scenario, the minimization of the recharge time of OH coils or stored energy for it will be essential and can be realized by driving sufficient current without increasing the plasma temperature too much. Low density and broad spectrum are shown to be favorable for the minimization. In the case of FER (Fusion Experimental Reactor under design study in JAERI) baseline parameters, the minimum recharge time is 3-5 s/V s. (orig.)

  5. A key to improved ion core confinement in the JET tokamak: ion stiffness mitigation due to combined plasma rotation and low magnetic shear.

    Science.gov (United States)

    Mantica, P; Angioni, C; Challis, C; Colyer, G; Frassinetti, L; Hawkes, N; Johnson, T; Tsalas, M; deVries, P C; Weiland, J; Baiocchi, B; Beurskens, M N A; Figueiredo, A C A; Giroud, C; Hobirk, J; Joffrin, E; Lerche, E; Naulin, V; Peeters, A G; Salmi, A; Sozzi, C; Strintzi, D; Staebler, G; Tala, T; Van Eester, D; Versloot, T

    2011-09-23

    New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.

  6. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    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 nTτ's ∼ 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

  7. Monte Carlo analysis of the effects of penetrations on the performance of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-01-01

    Adjoint Monte Carlo calculations have been carried out to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections and the nuclear heating and radiation damage were obtained using the latest available response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 greater than in the fully shielded coils depending on the location. Substantial increases in the radation damage were also noted

  8. Transmutation of minor actinides in a spherical torus tokamak fusion reactor, FDTR

    International Nuclear Information System (INIS)

    Feng, K.M.; Zhang, G.S.; Deng, M.G.

    2003-01-01

    In this paper, a concept for the transmutation of minor actinide (MA) nuclear wastes based on a spherical torus (ST) tokamak reactor, FDTR, is put forward. A set of plasma parameters suitable for the transmutation blanket was chosen. The 2-D neutron transport code TWODANT, the 3-D Monte Carlo code MCNP/4B, the 1-D neutron transport and burn-up calculation code BISON3.0 and their associated data libraries were used to calculate the transmutation rate, the energy multiplication factor and the tritium breeding ratio of the transmutation blanket. The calculation results for the system parameters and the actinide series isotopes for different operation times are presented. The engineering feasibility of the center-post (CP) of FDTR has been investigated and the results are also given. A preliminary neutronics calculation based on an ST transmutation blanket shows that the proposed system has a high transmutation capability for MA wastes. (author)

  9. Testing of the rectangular pivot-point bellows for the PPPL tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Haughian, J.; Lou, K.; Greer, J.; Fong, M.; Scalise, D.T.

    1983-12-01

    The Neutral Beam Pivot Point Bellows (PPB) is installed in the duct which connects the Neutral Beam Enclosure to the Torus. This bellows, located at the pivot point, must fit the severely limited space available at the pivot-point location. Consequently, it has to be made rectangular in cross section with a large inside area for beam access. This leads to small convolutions with high stress concentrations. The function of the bellows is to permit change in the angular positioning of the neutral beam line with respect to the Tokamak, to isolate the Neutral Beam Line from the deflection of the Torus during bake out, and to allow for all misalignments. Internally the bellows will have a vacuum along with such gases such as hydrogen or deuterium. Tests parameters are described

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

    International Nuclear Information System (INIS)

    Kim, Seokho H.; Berry, Jan

    2011-01-01

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

  11. Lithium adsorption by the first wall of fusion reactor-tokamak

    International Nuclear Information System (INIS)

    Bakunin, O.G.

    1989-01-01

    Lithium adsorption by the first wall of fusion reactor under stationary conditions and in the absence of chemical reactions is considered. Possibility of achieving 70% coating of the wall with lithium which can lead to sufficient decrease of sputtering is shown. 5 refs.; 5 figs

  12. Evaluation of neutron streaming through injection ports in a tokamak-type fusion reactor

    International Nuclear Information System (INIS)

    Ide, Takahiro; Seki, Yasushi; Iida, Hiromasa

    1976-03-01

    The effects of neutron streaming through injection ports in the fusion reactor designed in JAERI have been studied, especially those on tritium breeding ratio and the shielding of the superconducting magnet. In placement of the injection ports in the blanket, the tritium breeding ratio decreases by up to 1.3%, and shielding problem of the superconducting magnet is very important. (auth.)

  13. Fusion product measurements of the local ion thermal diffusivity in the PLT tokamak

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Lovberg, J.; Strachan, J.D.; Bell, R.E.

    1986-03-01

    Measurement of the gradient of the d-d fusion rate profile in an ohmic PLT plasma is used to deduce the gradient of the ion temperature and, thus, the local ion thermal diffusivity through an energy balance analysis. The inferred ion diffusivity is consistent with neoclassical theory

  14. Annual report of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development for the period of April 1, 1977 to March 31, 1978

    International Nuclear Information System (INIS)

    1979-02-01

    Research and development works in fiscal year 1977 of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development are described. 1) Theoretical studies on tokamak confinement have continued with more emphasis on computations. A task was started of developing a computer code system for mhd behavior of tokamak plasmas. 2) Experimental studies of lower hybrid heating up to 140 kW were made in JFT-2. The ion temperature was increased by 50% -- 60% near the plasma center. Plasma-wall interactions (particle and thermal fluxes to the wall, and titanium gettering) were studied. In JFT-2a (DIVA) ion sputtering, arcing and evaporation were identified, and the impurity ion sputtering was found to be a dominant origin of metal impurities in the present tokamaks. High temperature and high-density plasma divertor actions were demonstrated; i.e. the divertor decreases the radiation power loss by a factor of 3 and increases the energy confinement time by a factor of 2.5. Various diagnostic instruments operated sufficiently to provide useful information for the research with JFT-2 and JFT-2a(DIVA). 3) JFT-2 and JFT-2a(DIVA) operated as scheduled. Technological improvements were made such as titanium coating of the chamber wall, discharge cleaning and pre-ionization. 4) Detailed design of the prototype JT-60 neutral beam injector was made. A 200 kW, 650 MHz radiofrequency heating system for JFT-2 was completed; a lower hybrid heating experiment in JFT-2 was successful 5) In particle-surface interactions, the sputtering and surface erosion were studied. 6) Improvement designs of a superconducting cluster test facility and a test module coil were made in the toroidal coil development. 7) Second preliminary design of the tokamak experimental fusion reactor JXFR started in April 1977. Safety analyses were made of the main components and system of JXFR on the basis of the first preliminary design. (J.P.N.)

  15. START: the creation of a spherical tokamak

    International Nuclear Information System (INIS)

    Sykes, Alan

    1992-01-01

    The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

  16. Characterization of the fusion core in zebrafish endogenous retroviral envelope protein

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Jian [State Key Laboratory of Virology, College of Life Sciences, Wuhan University, Wuhan, Hubei 430072 (China); State Key Laboratory of Virology, Wuhan Institute of Virology, Chinese Academy of Sciences, Wuhan, Hubei 430071 (China); Zhang, Huaidong [CAS Key Laboratory of Special Pathogens and Biosafety, Wuhan Institute of Virology, Chinese Academy of Sciences, Wuhan, Hubei 430071 (China); Gong, Rui, E-mail: gongr@wh.iov.cn [CAS Key Laboratory of Special Pathogens and Biosafety, Wuhan Institute of Virology, Chinese Academy of Sciences, Wuhan, Hubei 430071 (China); Xiao, Gengfu, E-mail: xiaogf@wh.iov.cn [State Key Laboratory of Virology, College of Life Sciences, Wuhan University, Wuhan, Hubei 430072 (China); State Key Laboratory of Virology, Wuhan Institute of Virology, Chinese Academy of Sciences, Wuhan, Hubei 430071 (China)

    2015-05-08

    Zebrafish endogenous retrovirus (ZFERV) is the unique endogenous retrovirus in zebrafish, as yet, containing intact open reading frames of its envelope protein gene in zebrafish genome. Similarly, several envelope proteins of endogenous retroviruses in human and other mammalian animal genomes (such as syncytin-1 and 2 in human, syncytin-A and B in mouse) were identified and shown to be functional in induction of cell–cell fusion involved in placental development. ZFERV envelope protein (Env) gene appears to be also functional in vivo because it is expressible. After sequence alignment, we found ZFERV Env shares similar structural profiles with syncytin and other type I viral envelopes, especially in the regions of N- and C-terminal heptad repeats (NHR and CHR) which were crucial for membrane fusion. We expressed the regions of N + C protein in the ZFERV Env (residues 459–567, including predicted NHR and CHR) to characterize the fusion core structure. We found N + C protein could form a stable coiled-coil trimer that consists of three helical NHR regions forming a central trimeric core, and three helical CHR regions packing into the grooves on the surface of the central core. The structural characterization of the fusion core revealed the possible mechanism of fusion mediated by ZFERV Env. These results gave comprehensive explanation of how the ancient virus infects the zebrafish and integrates into the genome million years ago, and showed a rational clue for discovery of physiological significance (e.g., medicate cell–cell fusion). - Highlights: • ZFERV Env shares similar structural profiles with syncytin and other type I viral envelopes. • The fusion core of ZFERV Env forms stable coiled-coil trimer including three NHRs and three CHRs. • The structural mechanism of viral entry mediated by ZFERV Env is disclosed. • The results are helpful for further discovery of physiological function of ZFERV Env in zebrafish.

  17. First wall/blanket/shield design and power conversion for the ARIES-IV tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Conn, R.W.; Najmabadi, F.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10 MPa base pressure. The coolant flows poloidally in two loops, one inboard and one outboard. The coolant channels are circular tubes that form shells and are placed between two purge plates; the space between two adjacent tubes and the plate is purge gas flow area. The solid breeder is Li 2 O, and Be is used as neutron multiplier to ensure adequate TBR. Beryllium and Li 2 O are placed in between the adjacent tube shells. A computer code was developed to perform and optimize thermal-hydraulic design. Minimization of blanket thickness and the amount of Be, and the maximization of breeder zone thickness were done by iteration with neutronics. The gross thermal efficiency is 49%. The cost of electricity is 68 mills/kWh. The use of low activation SiC composite as the structural material, Li 2 O as the solid breeder, and avoidance of tungsten in the divertor has resulted in a good safety performance, and LSA rating of 1. Overall, SiC/He/Li 2 O ARIES-IV design is expected to have attractive economic and safety advantages

  18. Cryogenic distillation: a fuel enrichment system for near-term tokamak-type D-T fusion reactors

    International Nuclear Information System (INIS)

    Misra, B.; Davis, J.F.

    1980-02-01

    The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio

  19. Design of Fire/Gas Penetration Seals and fire exposure tests for Tokamak Fusion Test Reactor experimental areas

    International Nuclear Information System (INIS)

    Cavalluzzo, S.

    1983-01-01

    A Fire/Gas Penetration Seal is required in every penetration through the walls and ceilings into the Test Cell housing the Tokamak Fusion Test Reactor (TFTR), as well as other adjacent areas to protect the TFTR from fire damage. The penetrations are used for field coil lead stems, diagnostics systems, utilities, cables, trays, mechanical devices, electrical conduits, vacuum liner, air conditioning ducts, water pipes, and gas pipes. The function of the Fire/Gas Penetration Seals is to prevent the passage of fire and products of combustion through penetrations for a period of time up to three hours and remain structurally intact during fire exposure. The Penetration Seal must withstand, without rupture, a fire hose water stream directed at the hot surface. There are over 3000 penetrations ranging in size from several square inches to 100 square feet, and classified into 90 different types. The material used to construct the Fire/Gas Penetration Seals consist of a single and a two-component room temperature vulcanizing (RTV) silicone rubber compound. Miscellaneous materials such as alumina silica refractory fibers in board, blanket and fiber forms are also used in the construction and assembly of the Seals. This paper describes some of the penetration seals and the test procedures used to perform the three-hour fire exposure tests to demonstrate the adequacy of the seals

  20. Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

    1978-01-01

    A new type of superconducting magnet system for large fusion reactors is described. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil. (author)

  1. Demountable low stress high field toroidal field magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.; Hsieh, D.; Lehner, J.; Suenaga, M.

    1977-01-01

    A new type of superconducting magnet system for large fusion reactors is described in this report. Instead of winding large planar or multi-axis coils, as has been proposed in previous fusion reactor designs, the superconducting coils are made by joining together several prefabricated conductor sections. The joints can be unmade and sections removed if they fail. Conductor sections can be made at a factory and shipped to the reactor site for assembly. The conductor stress level in the assembled coil can be kept small by external support of the coil at a number of points along its perimeter, so that the magnetic forces are transmitted to an external warm reinforcement structure. This warm reinforcement structure can also be the primary containment for the fusion reactor, constructed similar to a PCRV (Prestressed Concrete Reactor Vessel) used in fission reactors. Low thermal conductivity, high strength supports are used to transfer the magnetic forces to the external reinforcement through a hydraulic system. The hydraulic supports are movable and can be programmed to accommodate thermal contraction and to minimize stress in the superconducting coil

  2. Design of a tokamak fusion reactor first wall armor against neutral beam impingement

    International Nuclear Information System (INIS)

    Myers, R.A.

    1977-12-01

    The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

  3. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  4. Low-noise cable for diagnostics, control and instrumentation of the ASDEX tokamak fusion experiment

    International Nuclear Information System (INIS)

    Gernhardt, J.

    1988-11-01

    ASDEX (Axially Symmetric Divertor EXperiment) is a large tokamak (R=1.65 m; a=0.4 m) with an air transformer. The relatively large stray field, Bz=10 mT=(100 G); for ρ=5 m, Bz=40 mT=(400 G); for ρ=3 m, BΦ=0.3 T=(3 kG); for ρ=3m, compared with that of an iron transformer, and the cable length 1≤30 m from the experiment to the control room, make mainly the magnetically induced and capacitively coupled noise signal in the cable relatively high. As a result of neutral injection (> 4 MW; 40 kV) and lower hybrid ion cyclotron and Alfven wave heating strong E-fields are produced and noise is coupled into the cables. These magnetic and electric field gradients during the plasma shot vary with time and location. This report tries to show how these noise signals can be reduced without reducing the broadcast frequency of the signal. The Electro Magnetic Compatibility and Interference (EMC, EMI) are discussed. The cost of diagnostic cable, connectors and cable ducts without mounting is approximately DM 700,000.--. (orig.)

  5. The testing of the Rectangular Pivot-point bellows for the PPPL Tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Haughian, J.; Fong, M.; Greer, J.; Lou, K.; Scalise, D.T.

    1983-01-01

    The Neutral Beam Pivot Point Bellows (PPB) is installed in the duct which connects the Neutral Beam Enclosure to the Torus. This bellows, located at the pivot point, must fit the severely limited space available at the pivot-point location. Consequently, it has to be made rectangular in cross section with a large inside area for beam access. This leads to small convolutions with high stress concentrations. The function of the bellows is to permit change in the angular positioning of the neutral beam line with respect to the Tokamak, to isolate the Neutral Beam Line from the deflection of the Torus during bake out, and to allow for all misalignments. Internally the bellows will have a vacuum along with such gases such as hydrogen or deuterium. Externally, air or nitrogen gas will be present. It is constructed of Inconel 718 convolutions welded together to provide a clear rectangular opening of 23.4 by 32.2 inches, joined to a 625 Inconel flange at each end

  6. A study on the fusion reactor - Study of ICRF coupling in the KAIST tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Duk In; Chang, Hong Young; Lee, Sun Chil; Jun, Sang Jin; Kwon, Gi Chung; Seo, Sung Hun; Heo, Sung Hoi; You, Kwang Il; Song, Soo Bin; Lee, Sung Chul; Kim, Min Chul; Lee, Chan Hui [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-09-01

    Research objectives are to design and fabricate the antenna, measure t property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling, Main heating method is ohmic heating at the KAIST tokamak. The power of the plasma produced by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, the RF system`s output power is about 10 kW and frequency range is 7 {approx} 30 MHz. In the first year, a 1 kW RF preamplifier was bought. In this year, a CW 2 kW RF main amp. and RF power monitoring system was bought. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. The matching network composed of two VVC (35 kV), 100 {approx} 1000 pF match firmly up to 50 kW power. We studied the measurement method of antenna impedance theoretically, and measured power efficiency and antenna impedance in the helicon plasma. 32 refs., 5 tabs., 29 figs. (author)

  7. ORNL TNS Program: plasma engineering considerations and innovations for a medium field tokamak fusion reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Attenberger, S.E.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-12-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments

  8. An electrically conducting first wall for the fusion engineering device-A (FED-A) tokamak

    International Nuclear Information System (INIS)

    Cramer, B.A.; Fuller, G.M.

    1983-01-01

    The first wall of the tokamak FED-A device was designed to satisfy two conflicting requirements. They are a low electrical resistance to give a long eddy-current decay time and a high neutron transparency to give a favorable tritium breeding ratio. The tradeoff between these conflicting requirements resulted in a copper alloy first wall that satisfied the specific goals for FED-A, i.e., a minimum eddy-current decay time of 0.5 sec and a tritium breeding ratio of at least 1.2. Aluminum alloys come close to meeting the requirements and would also probably work. Stainless steel will not work in this application because shells thin enough to satisfy temperature and stress limits are not thick enough to give a long eddy-current decay time and to avoid disruption induced melting. The baseline first wall design is a rib-stiffened, double-wall construction. The total wall thickness is 1.5 cm, including a water coolant thickness of 0.5 cm. The first wall is divided into twelve 30-degree sectors. Flange rings at the ends of each sector are bolted together to form the torus. Structural support is provided at the top center of each sector

  9. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn 3 conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended

  10. Conceptual study on high performance blanket in a spherical tokamak fusion-driven transmuter

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Yican

    2000-01-01

    A preliminary conceptual design on high performance dual-cooled blanket of fusion-driven transmuter is presented based on neutronic calculation. The dual-cooled system has some attractive advantages when utilized in transmutation of HLW (High Level Wastes). The calculation results show that this kind of blanket could safely transmute about 6 ton minor actinides (produced by 170 GW(e) Year PWRs approximately) and 0.4 ton fission products per year, and output 12 GW thermal power. In addition, the variation of power and critical factor of this blanket is relatively little during its 1-year operation period. This blanket is also tritium self-sustainable

  11. Developing maintainability for tokamak fusion power systems. Phase II report. Volume II: study results

    International Nuclear Information System (INIS)

    Fuller, G.M.; Zahn, H.S.; Mantz, H.C.; Kaletta, G.R.; Waganer, L.M.; Carosella, L.A.; Conlee, J.L.

    1978-11-01

    In this second phase the impact of unscheduled maintenance, several vacuum wall arrangements, and maintenance of other reactor interfacing subsystems and maintenance equipment are added to the evaluation of the maintainability of the fusion power reactor concepts. Four concepts are normalized to common performance parameters and evaluated for their capability to achieve availability and cost of electricity goals considering both scheduled and unscheduled maintenance. The results of this evaluation are used to generate a series of maintainability design guidelines and to select the more desirable features and design options which are used to configure a preliminary reactor concept having improved maintainability

  12. Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

    International Nuclear Information System (INIS)

    Giannone, L.; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T.

    2009-01-01

    The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

  13. Biochemical and biophysical characterization of the transmissible gastroenteritis coronavirus fusion core

    International Nuclear Information System (INIS)

    Ma Guangpeng; Feng Youjun; Gao Feng; Wang Jinzi; Liu Cheng; Li Yijing

    2005-01-01

    Transmissible gastroenteritis coronavirus (TGEV) is one of the most destructive agents, responsible for the enteric infections that are lethal for suckling piglets, causing enormous economic loss to the porcine fostering industry every year. Although it has been known that TGEV spiker protein is essential for the viral entry for many years, the detail knowledge of the TGEV fusion protein core is still very limited. Here, we report that TGEV fusion core (HR1-SGGRGG-HR2), in vitro expressed in GST prokaryotic expression system, shares the typical properties of the trimer of coiled-coil heterodimer (six α-helix bundle), which has been confirmed by a combined series of biochemical and biophysical evidences including size exclusion chromatography (gel-filtration), chemical crossing, and circular diagram. The 3D homologous structure model presents its most likely structure, extremely similar to those of the coronaviruses documented. Taken together, TGEV spiker protein belongs to the class I fusion protein, characterized by the existence of two heptad-repeat (HR) regions, HR1 and HR2, and the present knowledge about the truncated TGEV fusion protein core may facilitate in the design of the small molecule or polypeptide drugs targeting the membrane fusion between TGEV and its host

  14. Cryogenic aspects of a demountable toroidal field magnet system for tokamak type fusion reactors

    International Nuclear Information System (INIS)

    Hsieh, S.Y.; Powell, J.; Lehner, J.

    1977-01-01

    A new concept for superconducting Toroidal Field (TF) magnet construction is presented. It is termed the ''Demountable Externally Anchored Low Stress'' (DEALS) magnet system. In contrast to continuous wound conventional superconducting coils, each magnet coil is made from several straight coil segments to form a polygon which can be joined and disjoined to improve reactor maintenance accessibility or to replace failed coil segments if necessary. A design example is presented of a DEALS magnet system for a UWMAK II size reactor. The overall magnet system is described, followed by a detailed analysis of the major heat loads in order to assess the refrigeration requirements for the concept. Despite the increased heat loads caused by high current power leads (200,000 amps) and the coil warm reinforcement support system, the analysis shows that at most, only about one percent (approximately 20 Mw) of the plant electrical output (approximately 2,000 Mw) is needed to operate the magnet cryogenic system. The advantages and the drawbacks of the DEALS magnet system are also discussed. The advantages include: capability to replace failed coils, increased accessibility to the blanket shield assembly, reduced reliability requirements for the magnet, much lower stress in conductor, easier application of improved high field brittle superconductors like Nb 3 Sn, improved magnet safety features, etc. The drawbacks are the increased refrigeration requirements and the necessity of a movable coil support system. A comparison with a conventional magnet system is made. It is concluded that the benefits of the DEALS approach far outweigh its penalties, and that the DEALS concept is the most practical, economical way to construct TF magnet systems for Tokamak reactors

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

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Hans W. [Princeton Univ., NJ (United States)

    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 α-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 α-particle loss has led to a better understanding of α-particle dynamics in tokamaks. Intuitively, one might suppose that confined marginally passing α-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° lost alpha scintillator detector were observed during the shifts. It is calculated that the energy gained by an α-particle during the inward shift is sufficient to explain this result. However, an unexpected loss of partially thermalized α-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.

  16. Study of design parameters for minimizing the cost of electricity of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Tokimatsu, K.; Yamaji, K.; Katsurai, M.; Okano, K.; Yoshida, T.

    1998-01-01

    The impact of the design parameters on the cost of electricity (COE) is studied through a parameter survey in order to minimize the COE. Three kinds of operating modes are considered; first stability (FS), second stability (SS) and reversed shear (RS). The COE is calculated by a coupled physics-engineering-cost computer system code. Deuterium-tritium type, 1000 MW(e) at electric bus bar, steady state tokamak reactors with aspect ratios A from 3 to 4.5 are assumed. Several criteria are used for the parameter survey; for example, (a) the thermal to electrical conversion efficiency is assumed to be 34.5% using water as a coolant; (b) the average neutron wall load must not exceed 5 MW/m 2 for plasma major radius R p >5 m; (c) a 2 MeV neutral beam injector (NBI) is applied. It is found that the RS operating mode most minimizes the COE among the three operating modes by reducing the cost of the current drive and the coils and structures. The cost-minimized RS reactor can attain high f bs , high β N and low q 95 at the same time, which results in a short R p of 5.1 m, a low B max of the maximum magnetic toroidal field (TF) of the TF coils of 13 T and a low A of 3.0. It can be concluded that this cost-minimized RS reactor is the most cost-minimized within the frameworks of this study. This cost-minimized RS reactor has two advantages: one is that a B max =13 T TF coil can be made by use of ITER coil technology and the other is that the same cooling technology as that of ITER (water cooling) can be used. (author)

  17. Fuelling effect of tangential compact toroid injection in STOR-M Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Onchi, T.; Liu, Y., E-mail: tao668@mail.usask.ca [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Dreval, M. [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Inst. of Plasma Physics NSC KIPT, Kharkov (Ukraine); McColl, D. [Univ. of Saskatchewan, Dept. of Physics and Engineering Physics, Saskatoon, Saskatchewan (Canada); Asai, T. [Inst. of Plasma Physics NSC KIPT, Kharkov (Ukraine); Wolfe, S. [Nihon Univ., Dept. of Physics, Tokyo (Japan); Xiao, C.; Hirose, A. [Univ. of Saskatchewan, Saskatoon, Saskatchewan (Canada)

    2012-07-01

    Compact torus injection (CTI) is the only known candidate for directly fuelling the core of a tokamak fusion reactor. Compact torus (CT) injection into the STOR-M tokamak has induced improved confinement accompanied by an increase in the electron density, reduction in Hα emission, and suppression of the saw-tooth oscillations. The measured change in the toroidal flow velocity following tangential CTI has demonstrated momentum injection into the STOR-M plasma. (author)

  18. Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

    Science.gov (United States)

    Fukuda, Takeshi

    The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

  19. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Grotz, S.; Cheng, E.T.; Sharafat, S.; Cooke, P.I.H.

    1988-03-01

    TITAN-II is a compact, high power density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MWm/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a passive safety assurance design. 13 refs., 3 figs., 1 tab.

  20. Overview of the TITAN-II reversed-field pinch aqueous fusion power core design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Creedon, R.L.; Cheng, E.T. (General Atomic Co., San Diego, CA (USA)); Grotz, S.P.; Sharafat, S.; Cooke, P.I.H. (California Univ., Los Angeles (USA). Dept. of Mechanical, Aerospace and Nuclear Engineering; California Univ., Los Angeles, CA (USA). Inst. for Plasma and Fusion Research); TITAN Research Group

    1989-04-01

    TITAN-II is a compact, high-power-density Reversed-Field Pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO/sub 3/ and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the Reversed-Field Pinch reactor to operate at a neutron wall loading of 18 MW/m/sup 2/. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design. (orig.).

  1. Plasma core electron density and temperature measurements using CVI line emissions in TCABR Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, F. do, E-mail: fellypen@ifi.unicamp.br [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Centro de Componentes Semicondutores; Machida, M. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Instituto de Fisica Gleb Wataghin; Severo, J.H.F.; Sanada, E.; Ronchi, G. [Universidade de Sao Paulo (USP), SP (Brazil). Instituto de Fisica

    2015-08-15

    In this work, we present results of electron temperature (T{sub e} ) and density (n {sub e} ) measurements obtained in Tokamak Chauffage Alfven Bresilien (TCABR) tokamak using visible spectroscopy from CVI line emissions which occurs mainly near the center of the plasma column. The presented method is based on a well-known relationship between the particle flux (Γ {sub ion}) and the photon flux (ø {sub ion}) emitted by an ion species combined with ionizations per photon atomic data provided by the atomic data and analysis structure (ADAS) database. In the experiment, we measured the photon fluxes of three different CVI spectral line emissions, 4685.2, 5290.5, and 6200.6 Å (one line per shot). Using this method it was possible to find out the temporal evolution of T{sub e} and n{sub e} in the plasma. The results achieved are in good agreement with T{sub e} and n{sub e} measurements made using other diagnostic tools. (author)

  2. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

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

  3. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

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

  4. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  5. Feasibility studies on plasma vertical position control by ex-vessel coils in ITER-like tokamak fusion reactors

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Sugihara, Masayoshi; Shimomura, Yasuo

    1993-01-01

    Feasibility of the plasma vertical position control by control coils installed outside the vacuum vessel (ex-vessel) in a tokamak fusion reactor is examined for an ITER-like device. When a pair of ex-vessel control coils is made of normal conductor material and located near the outmost superconducting (SC) poloidal field (PF) coils, the applied voltage of several hundred volts on the control coils is the maximum allowable value which is limited by the maximum allowable induced voltage and eddy current heating on the SC PF coils, under the conditions that the SC PF coils are connected in series and a partitioning connection is employed for each of these PF coils. A proportional and derivative (PD) controller with and without voltage limitation has been employed to examine the feasibility. Indices of settling time and overshoot are introduced to measure the controllability of the control system. Based on these control schemes and indices, higher elongation (κ=2) and moderate elongation (κ=1.6) plasmas are examined for normal and deteriorated (low beta value and peaked current profile) plasma conditions within the restriction of applied voltage and current of control coils. The effect of the time constant of the passive stabilizer is also examined. The major results are: (1) A plasma with an elongation of 2.0 inevitably requires a passive stabilizer close to the plasma surface, (2) in case of a higher elongation than κ=2, even the ex-vessel control coil system is marginally controllable under normal plasma conditions, while it is difficult to control the deteriorated plasma conditions, (3) the time constant of the passive stabilizer is not an essential parameter for the controllability, (4) when the elongation is reduced down to 1.6, the ex-vessel control coil system can control the plasma even under deteriorated plasma conditions. (orig.)

  6. In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

    International Nuclear Information System (INIS)

    C.A. Gentile; C.H. Skinner; K.M. Young; M. Nishi; S. Langish; et al

    1999-01-01

    The Princeton Plasma Physics Laboratory (PPPL) Engineering and Research Staff in collaboration with members of the Japan Atomic Energy Research Institute (JAERI), Tritium Engineering Laboratory have commenced in-situ tritium measurements of the TFTR bumper limiter. The Tokamak Fusion Test Reactor (TFTR) operated with tritium from 1993 to 1997. During this time ∼ 53,000 Ci of tritium was injected into the TFTR vacuum vessel. After the cessation of TFTR plasma operations in April 1997 an aggressive tritium cleanup campaign lasting ∼ 3 months was initiated. The TFTR vacuum vessel was subjected to a regimen of glow discharge cleaning (GDC) and dry nitrogen and ''moist air'' purges. Currently ∼ 7,500 Ci of tritium remains in the vacuum vessel largely contained in the limiter tiles. The TFTR limiter is composed of 1,920 carbon tiles with an average weight of ∼ 600 grams each. The location and distribution of tritium on the TFTR carbon tiles are of considerable interest. Future magnetically confined fusion devices employing carbon as a limiter material may be considerably constrained due to potentially large tritium inventories being tenaciously held on the surface of the tiles. In-situ tritium measurements were conducted in TFTR bay L during August and November 1998. During the bay L measurement campaign open wall ion chambers and ultra thin thermoluminscent dosimeters (TLD) affixed to a boom and end effector were deployed into the vacuum vessel. The detectors were designed to make contact with the surface of the bumper limiter tile and to provide either real time (ion chamber) or passive (TLD) indication of the surface tritium concentration. The open wall ion chambers were positioned onto the surface of the tile in a manner which employed the surface of the tile as one of the walls of the chamber. The ion chambers, which are (electrically) gamma insensitive, were landed at four positions per tile. The geometry for landing the TLD's provided measurement at 24

  7. Development of thin foil Faraday collector as a lost alpha particle diagnostic for high yield D-T tokamak fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Van Belle, P; Jarvis, O N; Sadler, G J [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Cecil, F E [Colorado School of Mines, Golden, CO (United States)

    1994-07-01

    Alpha particle confinement is necessary for ignition of a D-T tokamak fusion plasma and for first wall protection. Due to high radiation backgrounds and temperatures, scintillators and semiconductor detectors may not be used to study alpha particles which are lost to the first wall during the D-T programs on JET and ITER. An alternative method of charged particle spectrometry capable of operation in these harsh environments, is proposed: it consists of thin foils of electrically isolated conductors with the flux of alpha particles determined by the positive current flowing from the foils. 2 refs., 3 figs.

  8. Comparison of explicit calculations for n = 3 to 8 dielectronic satellites of the FeXXV Kα resonance line with experimental data from the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Decaux, V.; Bitter, M.; Hsuan, H.; Hill, K.W.; von Goeler, S.; Park, H.; Bhalla, C.P.

    1991-12-01

    Dielectronic satellite spectra of the FeXXV Kα resonance line observed from the Tokamak Fusion Test Reactor (TFTR) plasmas have been compared with recent explicit calculations for the n = 3 to 8 dielectronic satellites as well as the earlier theoretical predictions, which were based on the 1/n 3 scaling law for n > 4 satellites. The analysis has been performed by least-squares fits of synthetic spectra to the experimental data. The synthetic spectra constructed from both theories are in good agreement with the observed data. However, the electron temperature values obtained from the fit of the present explicit calculations are in better agreement with independent measurements. 20 refs., 4 figs

  9. Conceptual design of an electrical power module for the tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Jassby, D.L.; Bullis, R.; Sedgeley, D.; Caldwell, C.S.; Pettus, W.G.; Schluderberg, D.C.

    1979-01-01

    The TFTR Engineering Test Station (ETS) can support blanket modules with a fusion-neutron view area of 0.5 m/sup 2/. If the TFTR magnetic systems and beam injectors can operate with pulse lengths of 5 s, once every 300 s, the time-averaged neutron power incident on a module will be 1.5 kW, which can be enhanced by a suitable blanket energy multiplier. A preliminary conceptual design of a dual-loop steam-generating power system that can be housed in the ETS has been carried out. The optimal heat transfer fluid in the primary loop is an organic liquid, which allows an operating temperature of 700/degree/F at low pressure. The primary coolant must be preheated electrically to operating temperature. A ballast tank levels the temperature at the steam generator, so that the secondary loop is in steady-state operation. With a natural-uranium blanket multiplier, the time-averaged net electrical power is 1.2 kW(e). 8 refs

  10. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

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

  11. Nonlinear fluid equations for fully toroidal electromagnetic waves for the core tokamak plasma

    Science.gov (United States)

    Weiland, J.; Liu, C. S.; Liu

    2013-12-01

    The rather general set of fluid equations with full curvature effects (Shukla and Weiland, Phys. Rev. A 40, 341 (1989)) has been modified to apply to the core and generalized to include also microtearing modes.

  12. The science program of the TCV tokamak: exploring fusion reactor and power plant concepts

    Science.gov (United States)

    Coda, S.; TCV Team

    2015-10-01

    TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic electron cyclotron resonance heating (ECRH) to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was employed to control the current density profile using electron cyclotron current drive. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace. Systematic studies of suprathermal electrons and ions in the presence of ECRH were performed. The L-H threshold power was measured to be ˜50-75% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core. The geodesic acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested successfully. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. The edge gradients were found to govern the scaling of confinement with current, power, density and triangularity. The dynamical interplay of confinement and magnetohydrodynamic modes leading to the density limit in TCV was documented. The heat flux profile decay lengths and heat load profile on the wall were documented in limited plasmas. In the snowflake (SF

  13. A study on the fusion reactor - Development of MHD stability and transport code for KT-2 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Koo; Shin, Kyo Jin [Pohang University of Science and Tecnology, Pohang (Korea, Republic of)

    1996-08-01

    MHD Stability analyses for KT-2 Tokamak were carried out by using CART (Resistive 3-D) Code. Linear Growth rates and linear perturbed eigen function of both N=0 axisymmetric mode and N=1 kink modes of highly elongated tokamak plasmas, in the presence of a conducting wall at various distances are computed and linear and nonlinear evolution of N=0 axisymmetric modes are simulated. 26 refs., 25 figs. (author)

  14. Diagnostic of the spatial and velocity distribution of alpha particles in tokamak fusion reactor using beat-wave generated lower hybrid wave. Progress report, 1994-1995

    International Nuclear Information System (INIS)

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

    1995-01-01

    The alpha particle population from fusion reactions in a DT tokamak reactor can have dramatic effects on the pressure profiles, energetic particle confinement, and the overall stability of the plasma; thus leading to important design consideration of a fusion reactor based on the tokamak concept. In order to fully understand the effects of the alpha population, a non-invasive diagnostic technique suitable for use in a reacting plasma environment needs to be developed to map out both the spatial and velocity distribution of the alphas. The proposed experimental goals for the eventual demonstration of LH wave interaction with a fast ion population is given in the reduced 3 year plan in table 1. At present time the authors are approaching the 8th month in their first year of this project. Up to now, their main effort has been concentrated in the operation of the two beat wave sources in burst mode. The second priority in the experimental project is the probe diagnostics and computer aided data acquisition system. The progress made so far is given, and they are ready to perform the beat-wave generated lower hybrid wave experiment. Some theoretical calculation had been reported at APS meetings. More refined theoretical models are being constructed in collaboration with Drs. J. Rogers and E. Valeo at PPPL

  15. Reconstruction and analysis of temperature and density spatial profiles inertial confinement fusion implosion cores

    International Nuclear Information System (INIS)

    Mancini, R. C.

    2007-01-01

    We discuss several methods for the extraction of temperature and density spatial profiles in inertial confinement fusion implosion cores based on the analysis of the x-ray emission from spectroscopic tracers added to the deuterium fuel. The ideas rely on (1) detailed spectral models that take into account collisional-radiative atomic kinetics, Stark broadened line shapes, and radiation transport calculations, (2) the availability of narrow-band, gated pinhole and slit x-ray images, and space-resolved line spectra of the core, and (3) several data analysis and reconstruction methods that include a multi-objective search and optimization technique based on a novel application of Pareto genetic algorithms to plasma spectroscopy. The spectroscopic analysis yields the spatial profiles of temperature and density in the core at the collapse of the implosion, and also the extent of shell material mixing into the core. Results are illustrated with data recorded in implosion experiments driven by the OMEGA and Z facilities

  16. Conceptual analysis of a tokamak reactor with lithium dust jet

    International Nuclear Information System (INIS)

    Kuteev, B.V.; Krylov, S.V.; Sergeev, V.Yu.; Skokov, V.G.; Timokhin, V.M.

    2010-01-01

    The steady-state operation of tokamak reactors requires radiating a substantial part of the fusion energy dissipated in plasma to make more uniform the heat loads onto the first wall and to reduce the erosion of the divertor plates. One of the approaches to realize this goal uses injection of lithium dust jet into the scrape-off layer (SOL). A quantitative conceptual analysis of the reactor parameters with lithium dust jet injection is presented here. The effects of the lithium on the core and SOL plasma are considered. The first results of developing the lithium jet injection technology and its application to the T-10 tokamak are also presented.

  17. Fusion Canada

    International Nuclear Information System (INIS)

    1987-07-01

    This first issue of a quarterly newsletter announces the startup of the Tokamak de Varennes, describes Canada's national fusion program, and outlines the Canadian Fusion Fuels Technology Program. A map gives the location of the eleven principal fusion centres in Canada. (L.L.)

  18. Integrated predictive modeling of high-mode tokamak plasmas using a combination of core and pedestal models

    International Nuclear Information System (INIS)

    Bateman, Glenn; Bandres, Miguel A.; Onjun, Thawatchai; Kritz, Arnold H.; Pankin, Alexei

    2003-01-01

    A new integrated modeling protocol is developed using a model for the temperature and density pedestal at the edge of high-mode (H-mode) plasmas [Onjun et al., Phys. Plasmas 9, 5018 (2002)] together with the Multi-Mode core transport model (MMM95) [Bateman et al., Phys. Plasmas 5, 1793 (1998)] in the BALDUR integrated modeling code to predict the temperature and density profiles of 33 H-mode discharges. The pedestal model is used to provide the boundary conditions in the simulations, once the heating power rises above the H-mode power threshold. Simulations are carried out for 20 discharges in the Joint European Torus and 13 discharges in the DIII-D tokamak. These discharges include systematic scans in normalized gyroradius, plasma pressure, collisionality, isotope mass, elongation, heating power, and plasma density. The average rms deviation between experimental data and the predicted profiles of temperature and density, normalized by central values, is found to be about 10%. It is found that the simulations tend to overpredict the temperature profiles in discharges with low heating power per plasma particle and to underpredict the temperature profiles in discharges with high heating power per particle. Variations of the pedestal model are used to test the sensitivity of the simulation results

  19. Fusion Canada issue 17

    International Nuclear Information System (INIS)

    1992-05-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on increased funding for the Canadian Fusion Program, news of the compact Toroid fuelling gun, an update on Tokamak de Varennes, the Canada - U.S. fusion meeting, measurements of plasma flow velocity, and replaceable Tokamak divertors. 4 figs

  20. Fusion Canada issue 17

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-05-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on increased funding for the Canadian Fusion Program, news of the compact Toroid fuelling gun, an update on Tokamak de Varennes, the Canada - U.S. fusion meeting, measurements of plasma flow velocity, and replaceable Tokamak divertors. 4 figs.

  1. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  2. A Functional Core of IncA Is Required for Chlamydia trachomatis Inclusion Fusion.

    Science.gov (United States)

    Weber, Mary M; Noriea, Nicholas F; Bauler, Laura D; Lam, Jennifer L; Sager, Janet; Wesolowski, Jordan; Paumet, Fabienne; Hackstadt, Ted

    2016-04-01

    Chlamydia trachomatis is an obligate intracellular pathogen that is the etiological agent of a variety of human diseases, including blinding trachoma and sexually transmitted infections. Chlamydiae replicate within a membrane-bound compartment, termed an inclusion, which they extensively modify by the insertion of type III secreted proteins called Inc proteins. IncA is an inclusion membrane protein that encodes two coiled-coil domains that are homologous to eukaryotic SNARE (soluble N-ethylmaleimide-sensitive factor attachment receptor) motifs. Recent biochemical evidence suggests that a functional core, composed of SNARE-like domain 1 (SLD-1) and part of SNARE-like domain 2 (SLD-2), is required for the characteristic homotypic fusion of C. trachomatis inclusions in multiply infected cells. To verify the importance of IncA in homotypic fusion in Chlamydia, we generated an incA::bla mutant. Insertional inactivation of incA resulted in the formation of nonfusogenic inclusions, a phenotype that was completely rescued by complementation with full-length IncA. Rescue of homotypic inclusion fusion was dependent on the presence of the functional core consisting of SLD-1 and part of SLD-2. Collectively, these results confirm in vitro membrane fusion assays identifying functional domains of IncA and expand the genetic tools available for identification of chlamydia with a method for complementation of site-specific mutants. Chlamydia trachomatis replicates within a parasitophorous vacuole termed an inclusion. The chlamydial inclusions are nonfusogenic with vesicles in the endocytic pathway but, in multiply infected cells, fuse with each other to form a single large inclusion. This homotypic fusion is dependent upon the presence of a chlamydial inclusion membrane-localized protein, IncA. Specificity of membrane fusion in eukaryotic cells is regulated by SNARE (soluble N-ethylmaleimide sensitive factor attachment receptor) proteins on the cytosolic face of vesicles and target

  3. Tokamak ARC damage

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

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

  5. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  6. A study on the fusion reactor - Development of x-ray spectrometer for diagnosis of tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hong Young; Choi, Duk In; Seo, Sung Hun; Kwon, Gi Chung; Jun, Sang Jin; Heo, Sung Hoi; Lee, Chan Hui [Korea Advanced Institute of Science and Technolgoy, Taejon (Korea, Republic of)

    1996-09-01

    This report of research is on the development of X-ray Photo-Electron Spectrometer (PES) for diagnosis of tokamak plasma. The spectrometer utilizes the fact that the energy of photo-electron is given by the difference between the energy of X-ray and the binding energy of materials. In the research of this year, we constructed two spectrometers; one is operated in KAIST tokamak and the other in KT1 tokamak. In addition, we reviewed the characteristics of the x-ray filter, the photo-electric effect of carbon foils and the detection efficiency of MCP and x-ray radiation of plasma. We measured the x-ray radiation in tokamak and diagnosed the qualitative plasma parameters from the analysis of data. The major interesting plasma parameters, which we can diagnose with the spectrometer, are the electron temperature, Z{sub eff}, the spatial distribution of x-ray radiation and etc. 27 refs., 2 tabs., 20 figs. (author)

  7. Fluorescent protein-tagged Vpr dissociates from HIV-1 core after viral fusion and rapidly enters the cell nucleus.

    Science.gov (United States)

    Desai, Tanay M; Marin, Mariana; Sood, Chetan; Shi, Jiong; Nawaz, Fatima; Aiken, Christopher; Melikyan, Gregory B

    2015-10-29

    HIV-1 Vpr is recruited into virions during assembly and appears to remain associated with the viral core after the reverse transcription and uncoating steps of entry. This feature has prompted the use of fluorescently labeled Vpr to visualize viral particles and to follow trafficking of post-fusion HIV-1 cores in the cytoplasm. Here, we tracked single pseudovirus entry and fusion and observed that fluorescently tagged Vpr gradually dissociates from post-fusion viral cores over the course of several minutes and accumulates in the nucleus. Kinetics measurements showed that fluorescent Vpr released from the cores very rapidly entered the cell nucleus. More than 10,000 Vpr molecules can be delivered into the cell nucleus within 45 min of infection by HIV-1 particles pseudotyped with the avian sarcoma and leukosis virus envelope glycoprotein. The fraction of Vpr from cell-bound viruses that accumulated in the nucleus was proportional to the extent of virus-cell fusion and was fully blocked by viral fusion inhibitors. Entry of virus-derived Vpr into the nucleus occurred independently of envelope glycoproteins or target cells. Fluorescence correlation spectroscopy revealed two forms of nuclear Vpr-monomers and very large complexes, likely involving host factors. The kinetics of viral Vpr entering the nucleus after fusion was not affected by point mutations in the capsid protein that alter the stability of the viral core. The independence of Vpr shedding of capsid stability and its relatively rapid dissociation from post-fusion cores suggest that this process may precede capsid uncoating, which appears to occur on a slower time scale. Our results thus demonstrate that a bulk of fluorescently labeled Vpr incorporated into HIV-1 particles is released shortly after fusion. Future studies will address the question whether the quick and efficient nuclear delivery of Vpr derived from incoming viruses can regulate subsequent steps of HIV-1 infection.

  8. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolas, T.

    2013-01-01

    The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr

  9. Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

    1985-12-01

    A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions

  10. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  11. Structural characterization of the fusion core in syncytin, envelope protein of human endogenous retrovirus family W

    International Nuclear Information System (INIS)

    Gong Rui; Peng Xiaoxue; Kang Shuli; Feng Huixing; Huang Jianying; Zhang Wentao; Lin Donghai; Tien Po; Xiao Gengfu

    2005-01-01

    Syncytin is a captive retroviral envelope protein, possibly involved in the formation of the placental syncytiotrophoblast layer generated by trophoblast cell fusion at the maternal-fetal interface. We found that syncytin and type I viral envelope proteins shared similar structural profiling, especially in the regions of N- and C-terminal heptad repeats (NHR and CHR). We expressed the predicted regions of NHR (41 aa) and CHR (34 aa) in syncytin as a native single chain (named 2-helix protein) to characterize it. 2-helix protein exists as a trimer and is highly α-helix, thermo-stable, and denatured by low pH. NHR and CHR could form a protease-resistant complex. The complex structure built by the molecular docking demonstrated that NHR and CHR associated in an antiparallel manner. Overall, the 2-helix protein could form a thermo-stable coiled coil trimer. The fusion core structure of syncytin was first demonstrated in endogenous retrovirus. These results support the explanation how syncytin mediates cytotrophoblast cell fusion involved in placental morphogenesis

  12. Evaluation of compatibility of flowing liquid lithium curtain for blanket with core plasma in fusion reactors

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua; Peng Lilin; Yan Jiancheng

    2003-01-01

    A global model analysis of the compatibility of flowing liquid lithium curtain for blanket with core plasma has been performed. The relationships between the surface temperature of lithium curtain and mean effective plasma charges, fuel dilution and produced fusion power have been obtained. Results show that under normal circumstances, the evaporation of liquid lithium does not affect Z eff seriously, but affects fuel dilution and fusion power sensitively. The authors have investigated the relationships between the flow velocity of liquid lithium and its surface temperature rise based on the conditions of the option II of the fusion experimental breeder (FEB-E) design with reversed shear configuration and fairly high power density. The authors concluded that the effects of evaporation from liquid lithium curtain for FEB-E on plasma are negligible even if the flow velocity of liquid lithium is as low as 0.5 m·s -1 . Finally, the sputtering yield of liquid lithium saturated by hydrogen isotopes is briefly discussed

  13. Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

    International Nuclear Information System (INIS)

    Sato, Satoshi

    2003-09-01

    In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

  14. Fusion reactor design studies

    International Nuclear Information System (INIS)

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  15. Fusion Canada issue 23

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-01-01

    A short bulletin from the National Fusion Program highlighting in this issue TdeV tokamak updates, fusion research in Korea, CCFM program review, TdeV divertor plasma, and CFFTP program review. 4 figs.

  16. Fusion Canada issue 27

    International Nuclear Information System (INIS)

    1995-03-01

    A short bulletin from the National Fusion Program highlighting in this issue ITER reactor siting, a major upgrade for TdeV tokamak, Ceramic Breeders: new tritium mapping technique and Joint Fusion Symposium. 2 figs

  17. Fusion Canada issue 23

    International Nuclear Information System (INIS)

    1994-01-01

    A short bulletin from the National Fusion Program highlighting in this issue TdeV tokamak updates, fusion research in Korea, CCFM program review, TdeV divertor plasma, and CFFTP program review. 4 figs

  18. Fusion Canada issue 6

    International Nuclear Information System (INIS)

    1989-02-01

    A short bulletin from the National Fusion Program. Included in this issue is a funding report for CFFTP, a technical update for Tokamak de Varennes and a network for university research by the National Fusion Program. 4 figs

  19. Tokamak transmutation of (nuclear) waste (TTW): Parametric studies

    International Nuclear Information System (INIS)

    Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.

    1994-01-01

    Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses

  20. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

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

    1985-12-01

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

  1. Fusion devices

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1977-01-01

    Three types of thermonuclear fusion devices currently under development are reviewed for an electric utilities management audience. Overall design features of laser fusion, tokamak, and magnetic mirror type reactors are described and illustrated. Thrusts and trends in current research on these devices that promise to improve performance are briefly reviewed. Twenty photographs and drawings are included

  2. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. World's largest DC flywheel generator for the toroidal field power supply of JAERI's JFT-2M Tokamak nuclear fusion reactor

    International Nuclear Information System (INIS)

    Tani, Takashi; Nakanishi, Yuji; Horita, Tsuyoshi; Kawase, Chiharu; Oyabu, Isao; Kishimoto, Takeshi.

    1996-01-01

    Mitsubishi Electric has delivered the world's largest DC generator for the toroidal field coil power supply of the JFT-2M Tokamak at the Japan Atomic Energy Research Institute. The unit rotates at 225 or 460 rpm, providing a maximum rated output of 2,700 V, 19,000 A and 51.3 MW. The toroidal field is a DC field, so use of a DC generator permits a simpler design consuming less floor space than an AC drive system. The generator was manufactured following extensive studies on commutation, mechanical strength and insulation. (author)

  4. Fusion

    CERN Document Server

    Mahaffey, James A

    2012-01-01

    As energy problems of the world grow, work toward fusion power continues at a greater pace than ever before. The topic of fusion is one that is often met with the most recognition and interest in the nuclear power arena. Written in clear and jargon-free prose, Fusion explores the big bang of creation to the blackout death of worn-out stars. A brief history of fusion research, beginning with the first tentative theories in the early 20th century, is also discussed, as well as the race for fusion power. This brand-new, full-color resource examines the various programs currently being funded or p

  5. Overview of Fusion-Fission Hybrid Reactor Design Study in China

    International Nuclear Information System (INIS)

    Huang Jinhua; Feng Kaiming; Deng Baiquan; Deng, P.Zh.; Zhang Guoshu; Hu Gang; He Kaihui; Wu Yican; Qiu Lijian; Huang Qunying; Xiao Bingjia; Liu Xiaoping; Chen Yixue; Kong, M.H.

    2002-01-01

    The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented

  6. Simulation of lower hybrid current drive in enhanced reversed shear plasmas in the tokamak fusion test reactor using the lower hybrid simulation code

    International Nuclear Information System (INIS)

    Kaita, R.; Bernabei, S.; Budny, R.

    1996-01-01

    The Enhanced Reversed Shear (ERS) mode has already shown great potential for improving the performance of the Tokamak Fusion Test Reactor (TFTR) and other devices. Sustaining the ERS, however, remains an outstanding problem. Lower hybrid (LH) current drive is a possible method for modifying the current profile and controlling its time evolution. To predict its effectiveness in TFTR, the Lower Hybrid Simulation Code (LSC) model is used in the TRANSP code and the Tokamak Simulation Code (TSC). Among the results from the simulations are the following. (1) Single-pass absorption is expected in TFTR ERS plasmas. The simulations show that the LH current follows isotherms of the electron temperature. The ability to control the location of the minimum in the q profile (q min ) has been demonstrated by varying the phase velocity of the launched LH waves and observing the change in the damping location. (2) LH current drive can been used to sustain the q min location. The tendency of qmin to drift inward, as the inductive current diffuses during the formation phase of the reversed shear discharge, is prevented by the LH current driven at a fixed radial location. If this results in an expanded plasma volume with improved confinement as high power neutral beam injection is applied, the high bootstrap currents induced during this phase can then maintain the larger qmin radius. (3) There should be no LH wave damping on energetic beam particles. The values of perpendicular index of refraction in the calculations never exceed about 20, while ions at TFR injection energies are resonant with waves having values closer to 100. Other issues being addressed in the study include the LH current drive efficiency in the presence of high bootstrap currents, and the effect of fast electron diffusion on LH current localization

  7. Draft program plant for TNS: The Next Step after the tokamak fusion test reactor. Part III. Project specific RD and D needs

    International Nuclear Information System (INIS)

    1977-03-01

    Research and development needs for the TNS systems are described according to the following chapters: (1) tokamak system, (2) electrical power systems, (3) plasma heating systems, (4) tokamak support systems, (5) instrumentation, control, and data systems, and (6) program recommendations

  8. Experimental identification of nonlinear coupling between (intermediate, small)-scale microturbulence and an MHD mode in the core of a superconducting tokamak

    Science.gov (United States)

    Sun, P. J.; Li, Y. D.; Ren, Y.; Zhang, X. D.; Wu, G. J.; Xu, L. Q.; Chen, R.; Li, Q.; Zhao, H. L.; Zhang, J. Z.; Shi, T. H.; Wang, Y. M.; Lyu, B.; Hu, L. Q.; Li, J.; The EAST Team

    2018-01-01

    In this paper, we present clear experimental evidence of core region nonlinear coupling between (intermediate, small)-scale microturbulence and an magnetohydrodynamics (MHD) mode during the current ramp-down phase in a set of L-mode plasma discharges in the experimental advanced superconducting tokamak (EAST, Wan et al (2006 Plasma Sci. Technol. 8 253)). Density fluctuations of broadband microturbulence (k\\perpρi˜2{-}5.2 ) and the MHD mode (toroidal mode number m = -1 , poloidal mode number n = 1 ) are measured simultaneously, using a four-channel tangential CO2 laser collective scattering diagnostic in core plasmas. The nonlinear coupling between the broadband microturbulence and the MHD mode is directly demonstrated by showing a statistically significant bicoherence and modulation of turbulent density fluctuation amplitude by the MHD mode.

  9. Assessment of feasibility of helium ash exhaust and heat removal by pumped-limiter in tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hitoki, Shigehisa; Sugihara, Masayoshi; Saito, Seiji; Fujisawa, Noboru

    1985-01-01

    A detailed calculation of the behavior of fuel and He particles in tokamak reactor with pumped-limiter is performed by one-dimensional tokamak transport code. Energy of neutral particles flowing back from limiter chamber is calculated by two-dimensional Monte Carlo neutral code. Feasibility of He ash exhaust and heat removal by the pumped-limiter are analyzed. Following features of the pumped-limiter are clarified: (1) Electron temperature decays rapidly in radial direction in scrape-off layer, while density profile is broader than that of temperature. (2) Helium accumulation in main plasma can be kept at desired level by rather short limiter and moderate pumping system. (3) Minimum amount of tritium pumped out little depends on limiter length. (4) Although high temperature plasma in scrape-off layer could be realized by large pumping and ideal pellet injection, it is not sufficiently high to reduce the erosion of the limiter surface and the leading edge. In conclusion, He ash exhaust may be possible by the pumped-limiter, while the heat load and erosion will be so high that the pumped-limiter may not be applicable unless the boundary plasma is cooled by radiation or by some other means. (author)

  10. A study on the fusion reactor - Numerical analyses of MHD equilibrium and= edge plasma transport in tokamak fusion reactor with divertor configurations

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Kang, Kyung Doo; Ryu, Ji Myung; Kim, Deok Kyu; Chung, TaeKyun; Chung, Mo Se [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

    1995-08-01

    In the present project for developing the numerical codes of 2-D MHD equilibrium, edge plasma transport and neutral particle transport for the tokamak plasmas, we computed the MHD equilibria of single and double null configurations and determined the external coil currents and the plasma parameters used for operation and control data. Also we numerically acquired the distributions of edge plasma parameters in poloidal and radial directions= and the design-related values according to the various operating conditions using the developed plasma transport code. Furthermore, a neutral particle transport code for the edge region is developed and them used for the analysis of the neutral particle behavior yielding the source terms in the fluid transport equations, and expected to supply the input parameters for the edge plasma transport code. 53 refs., 12 tabs., 44 figs. (author)

  11. Results from deuterium-tritium tokamak confinement experiments

    International Nuclear Information System (INIS)

    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

  12. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  13. Health physics around a controlled fusion research device: the Tokamak at Fontenay-aux-Roses (T.F.R.)

    International Nuclear Information System (INIS)

    1977-01-01

    The X and neutron dosimetry measurement near the magnetic confinement device for hot plasma, called T.F.R. (Tokamak, Fontenay-aux-Roses) are presented. The biological shielding consists of an ordinary concrete wall 30 cm thick; the dose rate is thus limited at 10 -1 mrem per discharge (corresponding to 10 mrem per day) in the whole area frequented by people during T.F.R. operation. A numerical calculation, taking into account the true geometry and X ray reflexion by the walls and roof, and normalized to the measurements, gives some indications on the electron beam which produces X rays. The photoneutron source (up to 10 10 neutrons per dischage) and the activation of the vacuum vessel result from high energy electrons (>= 10 MeV) supporting a 10 to 1,000 A current [fr

  14. Fusion Canada issue 28

    International Nuclear Information System (INIS)

    1995-06-01

    A short bulletin from the National Fusion Program highlighting in this issue the Canada - US fusion meeting in Montreal, fusion breeder work in Chile, new management at CFFTP, fast electrons in tokamaks: new data from TdeV, a program review of CCFM and Velikhov to address Montreal fusion meeting. 1 fig

  15. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  16. The ARIES-II and ARIES-IV second-stability tokamak reactors

    International Nuclear Information System (INIS)

    Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

    1992-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I 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. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

  17. ORR core re-configuration measurements to increase the fast neutron flux in the Magnetic Fusion Energy (MFE) experiments

    International Nuclear Information System (INIS)

    Hobbs, R.W.; Stinnett, R.M.; Sims, T.M.

    1985-06-01

    A study has been made of the relative increases obtainable in the fast neutron flux in the Magnetic Fusion Energy (MFE) experiment positions by reconfiguring the current ORR core. The study was made at the request of the MFE program to examine the percentage increase possible in the current displacement per atom (dpa) rate (assumed proportional to the fast flux). The principle methods investigated to increase the fast flux consisted of reducing the current core size (number of fuel elements) to increase the core average power density and arrangement of the fuel elements in the reduced-size core to tilt the core power distribution towards the MFE positions. The study concluded that fast fluxes in the E-3 core position could be increased by approximately 15 to 20% over current values and in E-5 by approximately 45 to 55%

  18. A preliminary study of a D-T tokamak fusion reactor with advanced blanket using the compact fusion advanced Brayton (CFAB) cycle

    International Nuclear Information System (INIS)

    Yoshikawa, K.; Ishikawa, M.; Umoto, J.; Fukuyama, A.; Mitarai, O.; Okamoto, M.; Sekimoto, H.; Nagatsu, M.

    1995-01-01

    Preliminary key issues for a synchrotron radiation-enhanced compact fusion advanced Brayton (CFAB) cycle fusion reactor similar to the CFAR (compact fusion advanced Rankine) cycle reactor are presented. These include plasma operation windows as a function of the first wall reflectivity and related issues, to estimate an allowance for deterioration of the first wall reflectivity due to dpa effects. It was found theoretically that first wall reflectivities down to 0.8 are still adequate for operation at an energy confinement scaling of 3 times Kaye-Goldston. Measurements of the graphite first wall reflectivities at Nagoya University indicate excellent reflectivities in excess of 90% for CC-312, PCC-2S, and PD-330S in the submillimeter regime, even at high temperatures in excess of 1000K. Some engineering issues inherent to the CFAB cycle are also discussed briefly in comparison with the CFAR cycle which uses hazardous limited-resource materials but is capable of using mercury as coolant for high heat removal. The CFAB cycle using helium coolant is found to achieve higher net plant conversion efficiencies in excess 60% using a non-equilibrium magnetohydrodynamic disk generator in the moderate pressure range, even at the cost of a relatively large pumping power, and at the penalty of high temperature materials, although excellent heat removal characteristics in the moderate pressure range need to be guaranteed in the future. (orig.)

  19. Structural safety assessment of a tokamak-type fusion facility for a through crack to cause cooling water leakage and plasma disruption

    International Nuclear Information System (INIS)

    Nakahira, Masataka

    2004-01-01

    A tokamak-type fusion machine has inherent safety associated with plasma shutdown. A small water leak can cause a plasma disruption although there is another possibility to terminate plasma without disruption. This plasma disruption will induce electromagnetic (EM) forces acting in the vacuum vessel (VV). From a radiological safety viewpoint, the VV is designed to form a physical barrier that encloses tritium and activated dust. If the VV can sustain an unstable fracture by EM forces from a through crack to cause the small leak, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a systematic approach to assure the structural safety is developed. A new analytical model to evaluate the through crack and leak rate of cooling water is proposed, with verification by experimental leak measurements. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is concluded that EM forces induced by the small leak to terminate plasma will not cause unstable fracture of the VV; thus the inherent safety is demonstrated. (author)

  20. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

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

    Energy Technology Data Exchange (ETDEWEB)

    Froio, A.; Bonifetto, R.; Carli, S.; Quartararo, A.; Savoldi, L., E-mail: laura.savoldi@polito.it; Zanino, R.

    2016-09-15

    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

  2. Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-05-01

    Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  4. Identification of a human protein-derived HIV-1 fusion inhibitor targeting the gp41 fusion core structure.

    Directory of Open Access Journals (Sweden)

    Lijun Chao

    Full Text Available The HIV-1 envelope glycoprotein (Env gp41 plays a crucial role in the viral fusion process. The peptides derived from the C-terminal heptad repeat (CHR of gp41 are potent HIV fusion inhibitors. However, the activity of these anti-HIV-1 peptides in vivo may be attenuated by their induction of anti-gp41 antibodies. Thus, it is essential to identify antiviral peptides or proteins with low, or no, immunogenicity to humans. Here, we found that the C-terminal fragment (aa 462-521 of the human POB1 (the partner of RalBP1, designated C60, is an HIV-1 fusion inhibitor. It bound to N36, the peptide derived from the N-terminal heptad repeat (NHR of gp41, and to the six-helix bundle (6-HB formed by N36 and C34, a CHR-peptide, but it did not bind to C34. Unlike the CHR-peptides, C60 did not block gp41 6-HB formation. Rather, results suggest that C60 inhibits HIV-1 fusion by binding to the 6-HB, in particular, the residues in the gp41 NHR domain that are exposed on the surface of 6-HB. Since 6-HB plays a crucial role in the late stage of fusion between the viral envelope and endosomal membrane during the endocytic process of HIV-1, C60 may serve as a host restriction factor to suppress HIV-1 entry into CD4+ T lymphocytes. Taken together, it can be concluded from these results that C60 can be used as a lead for the development of anti-HIV-1 therapeutics or microbicides for the treatment and prevention of HIV-1 infection, as well as a molecular probe to study the fusogenic mechanism of HIV-1.

  5. Compact Reversed-Field Pinch Reactors (CRFPR): fusion-power-core integration study

    International Nuclear Information System (INIS)

    Copenhaver, C.; Krakowski, R.A.; Schnurr, N.M.

    1985-08-01

    Using detailed two-dimensional neutronics studies based on the results of a previous framework study (LA-10200-MS), the fusion-power-core (FPC) integration, maintenance, and radio-activity/afterheat control are examined for the Compact Reversed-Field Pinch Reactor (CRFPR). While maintaining as a base case the nominal 20-MW/m 2 neutron first-wall loading design, CRFPR(20), the cost and technology impact of lower-wall-loading designs are also examined. The additional detail developed as part of this follow-on study also allows the cost estimates to be refined. The cost impact of multiplexing lower-wall-loading FPCs into a approx. 1000-MWe(net) plant is also examined. The CRFPR(20) design remains based on a PbLi-cooled FPC with pressurized-water used as a coolant for first-wall, pumped-limiter, and structural-shield systems. Single-piece FPC maintenance of this steady-state power plant is envisaged and evaluated on the basis of a preliminary layout of the reactor building. This follow-on study also develops the groundwork for assessing the feasibility and impact of impurity/ash control by magnetic divertors as an alternative to previously considered pumped-limiter systems. Lastly, directions for future, more-detailed power-plant designs based on the Reversed-Field Pinch are suggested

  6. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  7. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  8. Bringing fusion electric power closer

    International Nuclear Information System (INIS)

    Kintner, E.

    1977-01-01

    A review of the controlled fusion research program is given. The tokamak research program is described. Beam injection heating, control systems, and the safety of fusion reactors are topics that are also discussed

  9. The quest for fusion energy

    International Nuclear Information System (INIS)

    Johnson, J.L.

    1997-10-01

    A brief history of the magnetic fusion program from the point of view of a stellarator enthusiast who worked at a major tokamak laboratory. The reason that success in the magnetic fusion energy program is essential is presented. (author)

  10. Application and Continued Development of Thin Faraday Collectors as a Lost Ion Diagnostic for Tokamak Fusion Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    F. Ed Cecil

    2011-06-30

    This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

  11. Time-dependent tritium inventories and flow rates in fuel cycle components of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Kuan, W.

    1995-01-01

    Time-dependent inventories and flow rates for several components of the fuel cycle are modeled and studied through the use of a new modular-type model for the dynamic simulation of the fuel cycle in a fusion reactor. The complex dynamic behavior in the modeled subsystems is analyzed using this new model. Preliminary results using fuel cycle design configurations similar to ITER are presented and analyzed. The inventories and flow rates inside the primary vacuum pumping, fuel cleanup unit and isotope separation system are studied. Ways to minimize the tritium inventory are also assessed. This was performed by looking at various design options that could be used to minimize tritium inventory for specific components. (orig.)

  12. A thin foil Faraday collector as a lost alpha detector for high yield d-t tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Cecil, F. Ed

    2011-01-01

    This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

  13. JSPS-CAS Core University Program seminar on summary of 10-year collaborations in plasma and nuclear fusion research area

    International Nuclear Information System (INIS)

    Toi, Kazuo; Wang Kongjia

    2011-07-01

    The JSPS-CAS Core University Program (CUP) seminar on “Summary of 10-year Collaborations in Plasma and Nuclear Fusion Research Area” was held from March 9 to March 11, 2011 in the Okinawa Prefectural Art Museum, Naha city, Okinawa, Japan. The collaboration program on plasma and nuclear fusion started from 2001 under the auspices of Japanese Society of Promotion of Science (JSPS) and Chinese Academy of Sciences (CAS). This year is the last year of the CUP. This seminar was organized in the framework of the CUP. In the seminar, 29 oral talks were presented, having 14 Chinese and 30 Japanese participants. These presentations covered key topics related to the collaboration categories: (1) improvement of core plasma properties, (2) basic research on fusion reactor technologies, and (3) theory and numerical simulation. This seminar aims at summarizing the results obtained through the collaborations for 10 years, and discussing future prospects of China-Japan collaboration in plasma and nuclear fusion research areas. (author)

  14. Compact magnetic fusion systems

    Energy Technology Data Exchange (ETDEWEB)

    Linford, R.K.

    1983-12-01

    If the core (first wall, blanket, shield, and magnet coils) of fusion reactor systems could be made smaller in mass and volume for a given net electric power output than is usually predicted for the mainline tokamak/sup 1/ and mirror concepts, the cost of the technological development of the core and the construction of power plants might be significantly reduced. Although progress in plasma physics and engineering approaches should continue to yield improvements in reactor designs, certain physics features of the mainline concepts may prevent major reductions in the size of the core without straining the limits of technology. However, more than a factor of ten reduction in volume and mass of the core, at constant output power, may be possible for a class of toroidal confinement concepts in which the confining magnetic fields are supported more by currents flowing in the plasma than those in the external coils. In spite of this dramatic increase in power density (ratio of total thermal output power to the volume of the core), the design of compact systems need not rely on any materials requirements that are qualitatively more difficult than those proposed for the lower-power-density mainline fusion concepts. In some respects compact systems require less of an extension of existing technology, e.g. magnetics.

  15. Compact magnetic fusion systems

    International Nuclear Information System (INIS)

    Linford, R.K.

    1983-01-01

    If the core (first wall, blanket, shield, and magnet coils) of fusion reactor systems could be made smaller in mass and volume for a given net electric power output than is usually predicted for the mainline tokamak 1 and mirror concepts, the cost of the technological development of the core and the construction of power plants might be significantly reduced. Although progress in plasma physics and engineering approaches should continue to yield improvements in reactor designs, certain physics features of the mainline concepts may prevent major reductions in the size of the core without straining the limits of technology. However, more than a factor of ten reduction in volume and mass of the core, at constant output power, may be possible for a class of toroidal confinement concepts in which the confining magnetic fields are supported more by currents flowing in the plasma than those in the external coils. In spite of this dramatic increase in power density (ratio of total thermal output power to the volume of the core), the design of compact systems need not rely on any materials requirements that are qualitatively more difficult than those proposed for the lower-power-density mainline fusion concepts. In some respects compact systems require less of an extension of existing technology, e.g. magnetics

  16. Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

    International Nuclear Information System (INIS)

    Lee, Gil Soo

    2006-02-01

    To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

  17. The software-defined fast post-processing for GEM soft x-ray diagnostics in the Tungsten Environment in Steady-state Tokamak thermal fusion reactor

    Science.gov (United States)

    Krawczyk, Rafał Dominik; Czarski, Tomasz; Linczuk, Paweł; Wojeński, Andrzej; Kolasiński, Piotr; GÄ ska, Michał; Chernyshova, Maryna; Mazon, Didier; Jardin, Axel; Malard, Philippe; Poźniak, Krzysztof; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol

    2018-06-01

    This article presents a novel software-defined server-based solutions that were introduced in the fast, real-time computation systems for soft X-ray diagnostics for the WEST (Tungsten Environment in Steady-state Tokamak) reactor in Cadarache, France. The objective of the research was to provide a fast processing of data at high throughput and with low latencies for investigating the interplay between the particle transport and magnetohydrodynamic activity. The long-term objective is to implement in the future a fast feedback signal in the reactor control mechanisms to sustain the fusion reaction. The implemented electronic measurement device is anticipated to be deployed in the WEST. A standalone software-defined computation engine was designed to handle data collected at high rates in the server back-end of the system. Signals are obtained from the front-end field-programmable gate array mezzanine cards that acquire and perform a selection from the gas electron multiplier detector. A fast, authorial library for plasma diagnostics was written in C++. It originated from reference offline MATLAB implementations. They were redesigned for runtime analysis during the experiment in the novel online modes of operation. The implementation allowed the benchmarking, evaluation, and optimization of plasma processing algorithms with the possibility to check the consistency with reference computations written in MATLAB. The back-end software and hardware architecture are presented with data evaluation mechanisms. The online modes of operation for the WEST are discussed. The results concerning the performance of the processing and the introduced functionality are presented.

  18. Annual report of Division of Thermonuclear Fusion Research and Division of Large Tokamak Development for the period of April 1, 1976 to March 31, 1977

    International Nuclear Information System (INIS)

    1978-02-01

    Research and development activities in the two divisions are closely related. 1) Theoretical and computational studies continued on tokamak confinement and heating related to experimental problems. Studies on NBI heating in JT-60 were completed. 2) Experimental studies on impurities, density control and effects of density fluctuations were made in JFT-2. Neutral beams up to 30 keV and 8 A were injected into JFT-2 plasma perpendicularly. The ion temperature was increased by 10% - 15%, which is in agreement with the prediction by classical Fokker-Planck theory. In JFT-2a(DIVA), plasma-wall interaction (behavior of heavy and light impurities) was studies. The divertor of DIVA reduced the plasma-wall interaction and hence the radiation loss due to heavy impurities by a factor of 3. A grazing-incidence vacuum monochromator was first used in impurity studies in JFT-2 and JFT-2a. 3) Technological improvements were made raising efficiencies of operation, maintenance and plasma research. 4) Neutral beam injector test stand ITS-2 of 100 keV was completed. Construction of a 200 kW, 650 MHz radiofrequency heating system for JFT-2 was started. 5) Sputterings of molybdenum and pyrolytic graphite by low-energy protons and chemical reaction rates of pyrolytic graphite with protons were measured. Honeycomb structure greatly reduced the sputtered particles. 6) The superconducting magnet development group made the design of cluster test apparatus and the development of large current superconductor. 7) Phase-I preliminary design of experimental fusion reactor JXFR was completed and preliminary safety evaluation of JXFR was made. 8) Detailed design of JT-60 was completed in November 1976. Engineering development contracts were all completed by March 1977. 9) Engineering studies and tests on critical components of JT-4 with non-circular plasma cross section and divertors were made, after the preliminary design in fiscal year 1975. (auth.)

  19. Magnetic confinement by Tokamak: physical aspects

    International Nuclear Information System (INIS)

    Tachon, J.

    1980-01-01

    After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

  20. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  1. Task toward a Realization of Commercial Tokamak Fusion Plants in 2050 -The Role of ITER and the Succeeding Developments- 5.Challenge to Innovative Technologies and the Expected Market Appeal

    Science.gov (United States)

    Tobita, Kenji; Konishi, Satoshi; Tokimatsu, Koji; Nishio, Satoshi; Hiwatari, Ryoji

    This section describes the future of fusion energy in terms of its impact on the global energy supply and global warming mitigation, the possible entry scenarios of fusion into future energy market, and innovative technologies for deploying and expanding fusion's share in the market. Section 5.1 shows that fusion energy can contribute to the stabilization of atmospheric CO2 concentration if fusion is introduced into the future energy market at a competitive price. Considerations regarding fusion's entry scenarios into the energy market are presented in Sec. 5.2, suggesting that fusion should replace fossil energy sources and thus contribute to global warming mitigation. In this sense, first generation fusion power plants should be a viable energy source with global appeal and be so attractive as to be employed in developing countries rather than in developed countries. Favorable factors lending to this purpose are fusion's stability as a power source, and its security, safety, and environmental frendliness as well as its cost-of-electricity. The requirements for core plasma to expand the share of fusion in the market in the latter half of this century are given in Sec.5.3, pointing out the importance of high beta access with low aspect ratio and plasma profile control. From this same point of view, innovative fusion technologies worthy of further development are commented on in Sec. 5.4, addressing the high temperature blanket, hydrogen production, high temperature superconductors, and hot cell maintenance.

  2. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  3. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  4. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  5. Fusion energy

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The efforts of the Chemical Technology Division in fusion energy include the areas of fuel handling, processing, and containment. Current studies are concerned largely with the development of vacuum pumps for fusion reactors and experiments and with development and evaluation of techniques for recovering tritium from solid or liquid breeding blankets. In addition, a small effort is devoted to support of the ORNL design of a major Tokamak experiment, The Next Step (TNS)

  6. Development of Operation Scenario for Spherical Tokamak at SNU

    International Nuclear Information System (INIS)

    Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

    2009-01-01

    Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

  7. The need and prospects for improved fusion reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

    Conceptual fusion reactor studies over the past 10-15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100-200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed

  8. Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Abiteboul, J.

    2012-10-01

    The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

  9. Draft program plan for TNS: the next step after the tokamak fusion test reactor. Part I. Summary

    International Nuclear Information System (INIS)

    Roberts, M.

    1977-10-01

    A draft program plan for TNS has been prepared which consists of two basic parts--an R and D Needs Assessment and a Project Plan with schedules and necessary implementation steps. In this brief but intensive effort, questions concerning (1) the present basis for the TNS program, (2) the principal gaps in the supporting program, and (3) the necessary actions to be taken to implement the TNS program were examined. The study supported the thesis that the physics and technology bases do exist from which to start the TNS design as a central fusion program goal. Specific recommendations are made to emphasize those physics, technology, and engineering areas in which there are program gaps. In the project engineering study, a basic schedule with close support from the R and D program is developed from which recommendations on administrative actions and areas for further elucidation are made. This document presents in summary form the findings of the study, the development of the principal theses, and the recommendation to ERDA-DMFE

  10. Temperature distributions in a Tokamak vacuum vessel of fusion reactor after the loss-of-vacuum-events occurred

    International Nuclear Information System (INIS)

    Takase, K.; Kunugi, T.; Shibata, M.; Seki, Y.

    1998-01-01

    If a loss-of-vacuum-event (LOVA) occurred in a fusion reactor, buoyancy-driven exchange flows would occur at breaches of a vacuum vessel (VV) due to the temperature difference between the inside and outside of the VV. The exchange flows may bring mixtures of activated materials and tritium in the VV to the outside through the breaches, and remove decay heat from the plasma-facing components of the VV. Therefore, the LOVA experiments were carried out under the condition that one or two breaches was opened and that the VV was heated to a maximum 200 C, using a small-scaled LOVA experimental apparatus. Air and helium gas were provided as working fluids. Fluid and wall temperature distributions in the VV were measured and the flow patterns in the VV were estimated by using these temperature distributions. It was found that: (1) the exchange mass in the VV depended on the breach positions; (2) the exchange flow at the single breach case became a counter-current flow when the breach was at the roof of the VV and a stratified flow when it was at the side wall; (3) and that at the double breach case, a one-way flow between two breaches was formed. (orig.)

  11. User's perspective on fusion

    International Nuclear Information System (INIS)

    Ashworth, C.P.

    1976-01-01

    The need in fusion, from the electric utilities viewpoint, is for fusion to be a real option, not huge, complicated nuclear plants costing $10 billion each and requiring restructuring the energy industry to provide and use them. A course for future fusion reactor work in order to be a real option is discussed. The advantages of alternate concepts to the tokamak are presented

  12. Fusion Canada issue 9

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-11-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on availability of Canadian Tritium, an ITER update, a CCFM update on Tokamak and the new team organization, an international report on Fusion in Canada and a Laser Fusion Project at the University of Toronto. 3 figs.

  13. Fusion reactors - types - problems

    International Nuclear Information System (INIS)

    Schmitter, K.H.

    1979-07-01

    A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

  14. Fusion Canada issue 9

    International Nuclear Information System (INIS)

    1989-11-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on availability of Canadian Tritium, an ITER update, a CCFM update on Tokamak and the new team organization, an international report on Fusion in Canada and a Laser Fusion Project at the University of Toronto. 3 figs

  15. Advanced tokamak physics in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

    2000-12-01

    Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

  16. Magnetic fusion

    International Nuclear Information System (INIS)

    2002-01-01

    This document is a detailed lecture on thermonuclear fusion. The basic physics principles are recalled and the technological choices that have led to tokamaks or stellarators are exposed. Different aspects concerning thermonuclear reactors such as safety, economy and feasibility are discussed. Tore-supra is described in details as well as the ITER project

  17. MHD stability of an almost circular tokamak

    International Nuclear Information System (INIS)

    Roy, A.

    1990-10-01

    In a tokamak, the ratio β between the plasma pressure and that of the magnetic field is limited by the appearance of instabilities. The magnetic field in a tokamak reactor will always be limited by technological constraints. It is therefore crucial to know what factors have an effect on the β limit, since a zero resistivity plasma fluid model allows for theoretical reproduction of the β limits observed experimentally. Theoretical studies have shown that the distributions of pressure and current density may have a substantial effect on the β limit. The effect of the current density and pressure distributions on the β limit has been studied for tokamak with a circular core section. The best results are obtained when the current density is concentrated in the centre of the section and is nil at the periphery. But the second region of stability against ballooning modes cannot be obtained in a circular tokamak owing to the destabilisation of the universal modes. This study was then extended to the stability of plasmas the section of which is almost circular and has a point of reflection. Such configurations are vital for fusion since they allow systems in which the confinement time does not deteriorate with an increase in the additional heating power. The β limit was calculated for different positions of the reflection point. The results show that when it is displaced from the interior towards the exterior of the torus, the stability of the overall modes is progressively improved until it is vertical. But if the point of reflection is further displaced from this vertical position towards the exterior of the torus, localised modes close to the edge of the plasma are destabilised and bring about a drop in the β limit. (author) figs., tabs., 80 refs

  18. (Fusion energy research)

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, C.A. (ed.)

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer.

  19. [Fusion energy research

    International Nuclear Information System (INIS)

    Phillips, C.A.

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer

  20. Low temperature plasma near a tokamak reactor limiter

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Ignat, D.W.

    1987-01-01

    This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule

  4. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

    Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)

  5. Modeling of impurity transport in the core plasma

    International Nuclear Information System (INIS)

    Hulse, R.A.

    1992-01-01

    This paper presents a brief overview of computer modeling of impurity transport in the core region of controlled thermonuclear fusion plasmas. The atomic processes of importance in these high temperature plasmas and the numerical formulation of the model are described. Selected modeling examples are then used to highlight some features of the physics of impurity behavior in large tokamak fusion devices, with an emphasis on demonstrating the sensitivity of such modeling to uncertainties in the rate coefficients used for the atomic processes. This leads to a discussion of current requirements and opportunities for generating the improved sets of comprehensive atomic data needed to support present and future fusion impurity modeling studies

  6. Magnetic fusion; La fusion magnetique

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    This document is a detailed lecture on thermonuclear fusion. The basic physics principles are recalled and the technological choices that have led to tokamaks or stellarators are exposed. Different aspects concerning thermonuclear reactors such as safety, economy and feasibility are discussed. Tore-supra is described in details as well as the ITER project.

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

    International Nuclear Information System (INIS)

    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- 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

  8. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

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

  9. Fusion Canada issue 11

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-06-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on operation at Tokamak de Varennes, CRITIC irradiations at AECL, Tritium systems at TFTR, physics contribution at ITER. 4 figs.

  10. Summary of experimental core turbulence characteristics in ohmic and electron cyclotron resonance heated discharges in T-10 tokamak plasmas

    International Nuclear Information System (INIS)

    Vershkov, V.A.; Shelukhin, D.A.; Soldatov, S.V.; Urazbaev, A.O.; Grashin, S.A.; Eliseev, L.G.; Melnikov, A.V.

    2005-01-01

    This report summarizes the results of experimental turbulence investigations carried out at T-10 for more than 10 years. The turbulence characteristics were investigated using correlation reflectometry, multipin Langmuir probe (MLP) and heavy ion beam probe diagnostics. The reflectometry capabilities were analysed using 2D full-wave simulations and verified by direct comparison using a MLP. The ohmic and electron cyclotron resonance heated discharges show the distinct transition from the core turbulence, having complex spectral structure, to the unstructured one in the scrape-off layer. The core turbulence includes 'broad band, quasi-coherent' features, arising due to the excitation of rational surfaces with high poloidal m-numbers, with a low frequency near zero and specific oscillations at 15-30 kHz. All experimentally measured properties of low frequency and high frequency quasi-coherent oscillations are in good agreement with predictions of linear theory for the ion temperature gradient/dissipative trapped electron mode instabilities. Significant local changes in the turbulence characteristics were observed at the edge velocity shear layer and in the core near q = 1 radius after switching off the electron cyclotron resonance heating (ECRH). The local decrease in the electron heat conductivity and decrease in the turbulence level could be evidence of the formation of an electron internal transport barrier. The dynamic behaviour of the core turbulence was also investigated for the case of fast edge cooling and the beginning phase of ECRH

  11. Possible fusion reactor

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1976-05-01

    A scheme to improve performance characteristics of a tokamak-type fusion reactor is proposed. Basically, the tokamak-type plasma could be moved around so that the plasma could be heated by compression, brought to the region where the blanket surrounds the plasma, and moved so as to keep wall loading below the acceptable limit. This idea should be able to help to economize a fusion reactor

  12. Internal transport barrier formation and pellet injection simulation in helical and tokamak reactors

    International Nuclear Information System (INIS)

    Higashiyama, You; Yamazaki, Kozo; Arimoto, Hideki; Garcia, Jeronimo

    2008-01-01

    In the future fusion reactor, plasma density peaking is important for increase in the fusion power gain and for achievement of confinement improvement mode. Density control and internal transport barrier (ITB) formation due to pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. First, pellet injection simulation is carried out, including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of high-field side (HFS) pellet injection is clarified. Second, ITB simulation with pellet injection is carried out with the confinement improvement model based on the E x B shear effects, and it is found that deep pellet penetration is helpful for ITB formation as well as plasma core fuelling in the reversed-shear tokamak and helical reactors. (author)

  13. A Tutorial on Basic Principles of Microwave Reflectometry Applied to Fluctuation Measurements in Fusion Plasmas

    International Nuclear Information System (INIS)

    Nazikian, R.; Kramer, G.J.; Valeo, E.

    2001-01-01

    Microwave reflectometry is now routinely used for probing the structure of magnetohydrodynamic and turbulent fluctuations in fusion plasmas. Conditions specific to the core of tokamak plasmas, such as small amplitude of density irregularities and the uniformity of the background plasma, have enabled progress in the quantitative interpretation of reflectometer signals. In particular, the extent of applicability of the 1-D [one-dimensional] geometric optics description of the reflected field is investigated by direct comparison to 1-D full wave analysis. Significant advances in laboratory experiments are discussed which are paving the way towards a thorough understanding of this important measurement technique. Data is presented from the Tokamak Fusion Test Reactor [R. Hawryluk, Plasma Physics and Controlled Fusion 33 (1991) 1509] identifying the validity of the geometric optics description of the scattered field and demonstrating the feasibility of imaging turbulent fluctuations in fusion scale devices

  14. Maturation of the viral core enhances the fusion of HIV-1 particles with primary human T cells and monocyte-derived macrophages

    International Nuclear Information System (INIS)

    Jiang Jiyang; Aiken, Christopher

    2006-01-01

    HIV-1 infection requires fusion of viral and cellular membranes in a reaction catalyzed by the viral envelope proteins gp120 and gp41. We recently reported that efficient HIV-1 particle fusion with target cells is linked to maturation of the viral core by an activity of the gp41 cytoplasmic domain. Here, we show that maturation enhances the fusion of a variety of recombinant viruses bearing primary and laboratory-adapted Env proteins with primary human CD4 + T cells. Overall, HIV-1 fusion was more dependent on maturation for viruses bearing X4-tropic envelope proteins than for R5-tropic viruses. Fusion of HIV-1 with monocyte-derived macrophages was also dependent on particle maturation. We conclude that the ability to couple fusion to particle maturation is a common feature of HIV-1 Env proteins and may play an important role during HIV-1 replication in vivo

  15. Multitude of Core-Localized Shear Alfvén Waves in a High-Temperature Fusion Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Nazikian, R. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Berk, H. L. [Univ. of Texas, Austin, TX (United States); Budny, R. V. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Burrell, K. H. [General Atomics, San Diego, CA (United States); Doyle, E. J. [Univ. of California, Los Angeles, CA (United States); Fonck, R. J. [Univ. of Wisconsin, Madison, WI (United States); Gorelenkov, N. N. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Holcomb, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kramer, G. J. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Jayakumar, R. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); La Haye, R. J. [General Atomics, San Diego, CA (United States); McKee, G. R. [Univ. of Wisconsin, Madison, WI (United States); Makowski, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Peebles, W. A. [Univ. of California, Los Angeles, CA (United States); Rhodes, T. L. [Univ. of California, Los Angeles, CA (United States); Solomon, W. M. [Princeton Plasma Physics Laboratory (PPPL), Princeton, NJ (United States); Strait, E. J. [General Atomics, San Diego, CA (United States); VanZeeland, M. A. [Oak Ridge Institute for Science and Education (ORISE), Oak Ridge, TN (United States); Zeng, L. [Univ. of California, Los Angeles, CA (United States)

    2006-03-01

    Evidence is provided for a multitude of discrete frequency Alfvén waves in the core of magnetically confined high-temperature fusion plasmas. Multiple diagnostic instruments verify wave excitation over a wide spatial range from the device size at the longest wavelengths down to the thermal ion Larmor radius. At the shortest scales, the poloidal wavelengths are like the scale length of electrostatic drift wave turbulence. Theoretical analysis verifies a dominant interaction of the modes with particles in the thermal ion distribution traveling well below the Alfvén velocity.

  16. Characterization of BIV Env core: Implication for mechanism of BIV-mediated cell fusion

    International Nuclear Information System (INIS)

    Li Shu; Zhu Jieqing; Peng Yu; Cui Shanshan; Wang Chunping; Gao, George F.; Tien Po

    2005-01-01

    Entry of lentiviruses, such as human immunodeficiency virus type 1 (HIV-1) and simian immunodeficiency virus (SIV), requires folding of two heptad repeat regions (HR1 and HR2) of gp41 into a trimer-of-hairpins, which subsequently brings virus and cell membrane into fusion. This motif is a generalized feature of viral fusion proteins and has been exploited in generating antiviral fusion agents. In the present paper, we report structural characters of Env protein from another lentivirus, bovine immunodeficiency virus (BIV), which contributes to a good animal model of HIV. BIV HR1 and HR2 regions are predicted by two different programs and expressed separately or conjointly in Escherichia coli. Biochemical and biophysical analyses show that the predicted HRs of BIV Env can form a stable trimer-of-hairpins or six-helix bundle just like that formed by feline immunodeficiency virus Env. Cell fusion assay demonstrates that the HR2 peptide of BIV can efficiently inhibit the virus-mediated cell fusion

  17. Fusion Machines

    International Nuclear Information System (INIS)

    Weynants, R.R.

    2004-01-01

    A concise overview is given of the principles of inertial and magnetic fusion, with an emphasis on the latter in view of the aim of this summer school. The basis of magnetic confinement in mirror and toroidal geometry is discussed and applied to the tokamak concept. A brief discussion of the reactor prospects of this configuration identifies which future developments are crucial and where alternative concepts might help in optimising the reactor design. The text also aims at introducing the main concepts encountered in tokamak research that will be studied and used in the subsequent lectures

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

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

    This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

  19. An optical liquid level sensor based on core-offset fusion splicing method using polarization-maintaining fiber

    Science.gov (United States)

    Lou, Weimin; Chen, Debao; Shen, Changyu; Lu, Yanfang; Liu, Huanan; Wei, Jian

    2016-01-01

    A simple liquid level sensor using a small piece of hydrofluoric acid (HF) etched polarization maintaining fiber (PMF), with SMF-PMF-SMF fiber structure based on Mach- Zehnder interference (MZI) mechanism is proposed. The core-offset fusion splicing method induced cladding modes interfere with the core mode. Moreover, the changing liquid level would influence the optical path difference of the MZI since the effective refractive indices of the air and the liquid is different. Both the variations of the wavelength shifts and power intensity attenuation corresponding to the liquid level can be obtained with a sensitivity of 0.4956nm/mm and 0.2204dB/mm, respectively.

  20. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  1. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  2. Tokamak reactor startup power

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J A

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  4. Scattering measurements in Tokamak type devices

    International Nuclear Information System (INIS)

    Matoba, Tohru

    1975-03-01

    Theories, experiments and proposals for light scattering in Tokamak type devices are reviewed. Thomson scattering, measuring method of the current density distribution by scattering and resonance fluorescence are summarily described. These methods may be useful for diagnosis of the fusion plasmas. The report may help planning of the measuring apparatus for the fusion plasmas in future. (auth.)

  5. African Swine Fever Virus Undergoes Outer Envelope Disruption, Capsid Disassembly and Inner Envelope Fusion before Core Release from Multivesicular Endosomes.

    Directory of Open Access Journals (Sweden)

    Bruno Hernáez

    2016-04-01

    Full Text Available African swine fever virus (ASFV is a nucleocytoplasmic large DNA virus (NCLDV that causes a highly lethal disease in domestic pigs. As other NCLDVs, the extracellular form of ASFV possesses a multilayered structure consisting of a genome-containing nucleoid successively wrapped by a thick protein core shell, an inner lipid membrane, an icosahedral protein capsid and an outer lipid envelope. This structural complexity suggests an intricate mechanism of internalization in order to deliver the virus genome into the cytoplasm. By using flow cytometry in combination with pharmacological entry inhibitors, as well as fluorescence and electron microscopy approaches, we have dissected the entry and uncoating pathway used by ASFV to infect the macrophage, its natural host cell. We found that purified extracellular ASFV is internalized by both constitutive macropinocytosis and clathrin-mediated endocytosis. Once inside the cell, ASFV particles move from early endosomes or macropinosomes to late, multivesicular endosomes where they become uncoated. Virus uncoating requires acidic pH and involves the disruption of the outer membrane as well as of the protein capsid. As a consequence, the inner viral membrane becomes exposed and fuses with the limiting endosomal membrane to release the viral core into the cytosol. Interestingly, virus fusion is dependent on virus protein pE248R, a transmembrane polypeptide of the inner envelope that shares sequence similarity with some members of the poxviral entry/fusion complex. Collective evidence supports an entry model for ASFV that might also explain the uncoating of other multienveloped icosahedral NCLDVs.

  6. Modeling, numerical simulation and optimal control of plasma configuration evolution for the NET Tokamak and for the future generation of nuclear fusion

    International Nuclear Information System (INIS)

    Bourbon, F.

    1993-03-01

    The present thesis treats of the modeling, of the numerical simulation and equilibrium optimal control and configuration evolution of the plasma in a tokamak. In the first chapter, we establish the problem of equilibrium of the plasma by the finite elements method and a Newton algorithm. Then, we investigate the problem of equilibrium control: determination of currents in circuits of the tokamak to reach a given configuration. In the third chapter, we develop a model for evolution of the plasma configuration during a discharge. We solve this question by a Newton algorithm; with the fourth chapter, we treat the monitoring of plasma configuration evolution

  7. Tokamak Fusion Test Reactor Project

    International Nuclear Information System (INIS)

    1983-01-01

    Where the project's major contracts for buildings and equipment have experienced significant increases in cost, even though the contracts were for fixed prices, IG concluded that a major reason for the increases was the failure to adequately specify requirements prior to awarding the contracts. IG recommended that the Department should develop guidelines on what constitutes an adequate specification for a fixed-price contract and what controls should be placed over change orders. The Assistant Secretary for Management and Administration indicated in his comments to the draft report that an upcoming revision of the Accounting Practices and Procedures Handbook would establish controls over reallocations of construction funds to operating funds. Recommendations also propose that particular attention be given to ensuring that an agreed-upon plan and budget for completing the project is established, that a performance measurement system be implemented by the Princeton Plasma Physics Laboratory, and that improvements be made in the laboratory's Cost/Schedule Performance Report. Improvements are also needed in the quality assurance and safety programs of the Princeton Plasma Physics Laboratory. A number of recommendations from previous quality-assurance and safety reviews, performed by personnel from the Princeton Plasma Physics Laboratory and the Department of Energy, have not been implemented. Comments to the draft report also address these outstanding issues

  8. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.; Bell, R.E.

    2001-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  9. A comparative study of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

    International Nuclear Information System (INIS)

    Synakowski, E.J.; Beer, M.A.; Bell, R.E.

    1999-01-01

    Confinement bifurcations and subsequent plasma dynamics in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Observations suggest a framework in which ExB shear plays a dominant role in the barrier dynamics. In TFTR, bifurcations from the reverse shear (RS) into the enhanced reverse shear (ERS) regime with high power balanced neutral beam heating (above 25 MW at 4.8 T) resemble edge H mode transitions observed on DIII-D. In both, radial electric field (E r ) excursions precede confinement changes and are manifest as localized changes in the impurity poloidal rotation. Reduced transport follows the excursions, and in both cases strong E r shear is reinforced by the plasma pressure. These characteristics are contrasted with DIII-D negative central shear (NCS) barrier evolution with unidirectional beam injection. There, the improved confinement region can develop slowly, depending on the neutral beam input power and torque. Rapid expansion and deepening of this region follows an increase in the neutral beam heating power. The initial formation phase is modulated by confinement steps and interruptions. An analog for these steps is found in TFTR RS plasmas. Although these do not dominate the TFTR plasma evolution during low power (7 MW) heating, they can represent significant transport reductions when additional heating is applied. In both devices, no strong excursion in E r precedes these latter confinement bifurcations. The triggering event of these steps may be related to current profile relaxation, but it is not always connected with simple integral or half-integer values of the minimum in the q profile. Finally, variations of E r and the ExB shear through the application of unidirectional injection on TFTR yielded plasmas with confinement characteristics and barrier dynamics similar to those of DIII-D NCS plasmas. The data underscore that the physics responsible for the enhanced confinement states is fundamentally

  10. The role of the spherical tokamak in clarifying tokamak physics

    International Nuclear Information System (INIS)

    Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

    1999-01-01

    The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

  11. Electrostatic Architecture of the Infectious Salmon Anemia Virus (ISAV) Core Fusion Protein Illustrates a Carboxyl-Carboxylate pH Sensor.

    Science.gov (United States)

    Cook, Jonathan D; Soto-Montoya, Hazel; Korpela, Markus K; Lee, Jeffrey E

    2015-07-24

    Segment 5, ORF 1 of the infectious salmon anemia virus (ISAV) genome, encodes for the ISAV F protein, which is responsible for viral-host endosomal membrane fusion during a productive ISAV infection. The entry machinery of ISAV is composed of a complex of the ISAV F and ISAV hemagglutinin esterase (HE) proteins in an unknown stoichiometry prior to receptor engagement by ISAV HE. Following binding of the receptor to ISAV HE, dissociation of the ISAV F protein from HE, and subsequent endocytosis, the ISAV F protein resolves into a fusion-competent oligomeric state. Here, we present a 2.1 Å crystal structure of the fusion core of the ISAV F protein determined at low pH. This structure has allowed us to unambiguously demonstrate that the ISAV entry machinery exhibits typical class I viral fusion protein architecture. Furthermore, we have determined stabilizing factors that accommodate the pH-dependent mode of ISAV transmission, and our structure has allowed the identification of a central coil that is conserved across numerous and varied post-fusion viral glycoprotein structures. We then discuss a mechanistic model of ISAV fusion that parallels the paramyxoviral class I fusion strategy wherein attachment and fusion are relegated to separate proteins in a similar fashion to ISAV fusion. © 2015 by The American Society for Biochemistry and Molecular Biology, Inc.

  12. First wall and blanket stresses induced by cyclic fusion core operations

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.; Kostoff, R.N.

    1981-01-01

    An analysis is made of cyclic thermal loads and stresses for the complete range of operating conditions. Two critical components were examined; the solid wall adjacent to the fusion plasma (first wall) and the fuel elements in the high power density region of the blanket. Simple closed form expressions were derived for temperature increases and thermal stresses that may be evaluated conveniently and rapidly and the values compared for different systems

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Cost assessment of a generic magnetic fusion reactor

    International Nuclear Information System (INIS)

    Sheffield, J.; Dory, R.A.; Cohn, S.M.; Delene, J.G.; Parsly, L.F.; Ashby, D.E.T.F.; Reiersen, W.T.

    1986-03-01

    A generic reactor model is used to examine the economic viability of generating electricity by magnetic fusion. The simple model uses components that are representative of those used in previous reactor studies of deuterium-tritium-burning tokamaks, stellarators, bumpy tori, reversed-field pinches (RFPs), and tandem mirrors. Conservative costing assumptions are made. The generic reactor is not a tokamak; rather, it is intended to emphasize what is common to all magnetic fusion rectors. The reactor uses a superconducting toroidal coil set to produce the dominant magnetic field. To this extent, it is not as good an approximation to systems such as the RFP in which the main field is produced by a plasma current. The main output of the study is the cost of electricity as a function of the weight and size of the fusion core - blanket, shield, structure, and coils. The model shows that a 1200-MW(e) power plant with a fusion core weight of about 10,000 tonnes should be competitive in the future with fission and fossil plants. Studies of the sensitivity of the model to variations in the assumptions show that this result is not sensitively dependent on any given assumption. Of particular importance is the result that a fusion reactor of this scale may be realized with only moderate advances in physics and technology capabilities

  15. The fusion reactor

    International Nuclear Information System (INIS)

    Brennan, M.H.

    1974-01-01

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  16. Fusion Canada issue 8

    International Nuclear Information System (INIS)

    1989-08-01

    A short bulletin from the National Fusion Program. Included in this issue are Canada-ITER contributions, NET Fuel Processing Loop, Bilateral Meeting for Canada-Europe, report from Tokamak de Varennes and a report from the University of Toronto on materials research for Fusion Reactors. 3 figs

  17. Fusion Canada issue 15

    International Nuclear Information System (INIS)

    1991-10-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on the 1996 IAEA Fusion Conference site, operations at the Tokamak de Varennes including divertor pumping of impurities and pumping of carbon monoxide and methane, a discussion of the CFFTP and it's role. 1 fig

  18. Fusion Canada issue 22

    International Nuclear Information System (INIS)

    1993-10-01

    A short bulletin from the National Fusion Program highlighting in this issue a bi-lateral meeting between Canada and Japan, water and hydrogen detritiation, in-situ tokamak surface analysis, an update of CCFM/TdeV and tritium accounting Industry guidance in Fusion, fast probe for plasma-surface interaction. 4 figs

  19. Fusion Canada issue 15

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-10-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on the 1996 IAEA Fusion Conference site, operations at the Tokamak de Varennes including divertor pumping of impurities and pumping of carbon monoxide and methane, a discussion of the CFFTP and it`s role. 1 fig.

  20. Fusion Canada issue 8

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-08-01

    A short bulletin from the National Fusion Program. Included in this issue are Canada-ITER contributions, NET Fuel Processing Loop, Bilateral Meeting for Canada-Europe, report from Tokamak de Varennes and a report from the University of Toronto on materials research for Fusion Reactors. 3 figs.

  1. Controlled thermonuclear fusion

    CERN Document Server

    Bobin, Jean Louis

    2014-01-01

    The book is a presentation of the basic principles and main achievements in the field of nuclear fusion. It encompasses both magnetic and inertial confinements plus a few exotic mechanisms for nuclear fusion. The state-of-the-art regarding thermonuclear reactions, hot plasmas, tokamaks, laser-driven compression and future reactors is given.

  2. Fusion Canada issue 4

    International Nuclear Information System (INIS)

    1988-05-01

    A short bulletin from the National Fusion Program. Included in this issue is a technical update on Tokamak de Varennes, a report on the Beatrix II Breeding Materials Test Program, the Tritium glovebox system for UPM, Saudi Arabia, a broad update of the Canadian Fusion Fuels Technology Project is also included. 1 fig

  3. Fusion Canada issue 12

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-10-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on Darlington`s Tritium Removal Facility, work at universities on Deuterium Diffusivity in Beryllium, Fusion Studies, confinement research and the operation of divertors at Tokamak de Varennes. 5 figs.

  4. Fusion Canada issue 22

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-10-01

    A short bulletin from the National Fusion Program highlighting in this issue a bi-lateral meeting between Canada and Japan, water and hydrogen detritiation, in-situ tokamak surface analysis, an update of CCFM/TdeV and tritium accounting Industry guidance in Fusion, fast probe for plasma-surface interaction. 4 figs.

  5. Fusion Canada issue 4

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-05-01

    A short bulletin from the National Fusion Program. Included in this issue is a technical update on Tokamak de Varennes, a report on the Beatrix II Breeding Materials Test Program, the Tritium glovebox system for UPM, Saudi Arabia, a broad update of the Canadian Fusion Fuels Technology Project is also included. 1 fig.

  6. Fusion Canada issue 14

    International Nuclear Information System (INIS)

    1991-05-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on a fusion cooperation agreement between Japan and Canada, an update at Tokamak de Varennes on plasma biasing experiments and boronization tests and a collaboration between Canada and the U.S. on a compact toroid fuelling gun. 4 figs

  7. Fusion Canada issue 12

    International Nuclear Information System (INIS)

    1990-10-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on Darlington's Tritium Removal Facility, work at universities on Deuterium Diffusivity in Beryllium, Fusion Studies, confinement research and the operation of divertors at Tokamak de Varennes. 5 figs

  8. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  9. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.; California Univ., Los Angeles

    1984-01-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

  10. Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Fraboulet, D.

    1996-09-17

    Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.

  11. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  12. A Study of Performance in Low-Power Tokamak Reactor with Integrated Predictive Modeling Code

    International Nuclear Information System (INIS)

    Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

    2009-07-01

    Full text: A fusion hybrid or a small fusion power output with low power tokamak reactor is presented as another useful application of nuclear fusion. Such tokamak can be used for fuel breeding, high-level waste transmutation, hydrogen production at high temperature, and testing of nuclear fusion technology components. In this work, an investigation of the plasma performance in a small fusion power output design is carried out using the BALDUR predictive integrated modeling code. The simulations of the plasma performance in this design are carried out using the empirical-based Mixed Bohm/gyro Bohm (B/gB) model, whereas the pedestal temperature model is based on magnetic and flow shear (δ α ρ ζ 2 ) stabilization pedestal width scaling. The preliminary results using this core transport model show that the central ion and electron temperatures are rather pessimistic. To improve the performance, the optimization approach are carried out by varying some parameters, such as plasma current and power auxiliary heating, which results in some improvement of plasma performance

  13. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  14. A study on the fusion reactor - A study on wave physics of fast wave heating and the current drive in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Su Won; Yeom, Hyun Ju [Kyonggi University, Suwon (Korea, Republic of); Hong, Sang Hee; Chung, Mo Se [Seoul National University, Seoul (Korea, Republic of)

    1996-09-01

    A full 3-dimensional code for fast wave heating and the current drive has been developed ant its results are compared with those of FASTWA for Phaedrus-T tokamak. The finite Larmour radius expansion and the order reduction method have been used to derive the wave equation in the toroidal coordinate from the Maxwell-Vlasov equations. By expanding the fields in poloidal Fourier series, the wave equations are reduced to the system of ordinary differential equations in the radial axis, which are then numerically integrated via the shooting method. In addition, the convergence of the solutions and energy conservation are discussed. Finally, and example calculation of the current drive is presented for the advanced superconducting tokamak which is in its conceptual design phase. 17 refs., 10 tabs., 31 figs. (author)

  15. Laser fusion overview

    International Nuclear Information System (INIS)

    Nuckolls, J.

    1976-01-01

    Because of recent breakthroughs in the target area, and in the glass laser area, the scientific feasibility of laser fusion--and of inertial fusion--may be demonstrated in the early 1980's. Then the development in that time period of a suitable laser (or storage ring or other driving source) would make possible an operational inertial fusion reactor in this century. These are roughly the same time scales as projected by the Tokamak magnetic confinement approach. It thus appears that the 15-20 year earlier start by magnetic confinement fusion may be overcome. Because inertial confinement has been demonstrated, and inertial fusion reactors may operate on smaller scales than Tokamaks, laser fusion may have important technical and economic advantages

  16. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  17. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1998-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  18. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  19. Thermonuclear fusion

    International Nuclear Information System (INIS)

    Weisse, J.

    2000-01-01

    This document takes stock of the two ways of thermonuclear fusion research explored today: magnetic confinement fusion and inertial confinement fusion. The basic physical principles are recalled first: fundamental nuclear reactions, high temperatures, elementary properties of plasmas, ignition criterion, magnetic confinement (charged particle in a uniform magnetic field, confinement and Tokamak principle, heating of magnetized plasmas (ohmic, neutral particles, high frequency waves, other heating means), results obtained so far (scale laws and extrapolation of performances, tritium experiments, ITER project), inertial fusion (hot spot ignition, instabilities, results (Centurion-Halite program, laser experiments). The second part presents the fusion reactor and its associated technologies: principle (tritium production, heat source, neutron protection, tritium generation, materials), magnetic fusion (superconducting magnets, divertor (role, principle, realization), inertial fusion (energy vector, laser adaptation, particle beams, reaction chamber, stresses, chamber concepts (dry and wet walls, liquid walls), targets (fabrication, injection and pointing)). The third chapter concerns the socio-economic aspects of thermonuclear fusion: safety (normal operation and accidents, wastes), costs (costs structure and elementary comparison, ecological impact and external costs). (J.S.)

  20. Steady State Advanced Tokamak (SSAT): The mission and the machine

    International Nuclear Information System (INIS)

    Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

    1992-03-01

    Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

  1. Focus on nuclear fusion research

    Czech Academy of Sciences Publication Activity Database

    Křenek, Petr; Mlynář, Jan

    2011-01-01

    Roč. 61, - (2011), s. 62-63 ISSN 0375-8842 Institutional research plan: CEZ:AV0Z20430508 Keywords : ITER * COMPASS * fusion energy * tokamak * EURATOM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.ipp.cas.cz/Tokamak/clanky/energetika_COMPASS.pdf

  2. Proceedings of JSPS-CAS core university program seminar on PWI/PFC and fusion technologies

    International Nuclear Information System (INIS)

    Hino, Tomoaki; Chen Junling

    2009-01-01

    The present symposium has the topics on plasma wall interaction, plasma facing materials and components, core plasma behavior, blanket, tritium, superconductor, ITER related R and D and theory. The number of the presented papers was 39, and the participants were 45, 18 from Japan and 27 from China. Very aggressive discussions were conducted in this symposium, and the understanding for important issues toward to ITER and Demonstration Reactor was deepened. The 29 of the papers are indexed individually. (J.P.N.)

  3. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  4. Tokamak Physics Experiment (TPX) design

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1995-01-01

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

  5. A programmatic framework for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

    1993-01-01

    Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

  6. Alternate fusion -- continuous inertial confinement

    International Nuclear Information System (INIS)

    Barnes, D.C.; Turner, L.; Nebel, R.A.

    1993-01-01

    The authors argue that alternate approaches to large tokamak confinement are appropriate for fusion applications if: (1) They do not require magnetic confinement of a much higher quality than demonstrated in tokamaks; (2) Their physics basis may be succinctly stated and experimentally tested; (3) They offer near-term applications to important technical problems; and (4) Their cost to proof-of-principle is low enough to be consistent with current budget realities. An approach satisfying all of these criteria is presented. Fusion systems based on continuous inertial confinement are described. In these approaches, the inertia of a nonequilibrium plasma is used to produce local concentrations of plasma density in space and/or time. One implementation (inertial electrostatic confinement) which has been investigated both experimentally and theoretically uses a system of electrostatic grids to accelerate plasma ions toward a spherical focus. This system produced a steady 2 x 10 10 D-T neutrons/second with an overall fusion gain of 10 -5 in a sphere of about 9 cm radius. Recent theoretical developments show how to raise the fusion gain to order unity or greater by replacing the internal grids by a combination of applied magnetic and electrostatic fields. In these approaches, useful thermonuclear conditions may be produced in a system as small as a few mm radius. Confinement is that of a nonneutralized plasma. A pure electron plasma with a radial beam velocity distribution is absolutely confined by an applied Penning trap field. Spherical convergence of the confined electrons forms a deep virtual cathode near r = 0, in which thermonuclear ions are absolutely confined at useful densities. The authors have examined the equilibrium, stability, and classical relaxation of such systems, and obtained many positive physics results. Equilibria exist for both pure electron and partially charge-neutralized systems with arbitrarily high core-plasma densities

  7. Tokamak WEST připraven ke startu!

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Květen (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * WEST * Tora Supra * divertor Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/2014-tokamak-west-pripraven-ke- start u

  8. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

    Porkolab, M.

    1985-10-01

    Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

  9. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

    htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma

  10. Fusion technology: The Iter fusion experiment

    International Nuclear Information System (INIS)

    Dietz, K.J.

    1994-01-01

    Plans for the Iter international fusion experiment, in which the European Union, Japan, Canada, Russia, Sweden, Switzerland, and the USA cooperate, were begun in 1985, and construction work started in early 1994. These activities serve for the preparation of the design and construction documents for a research reactor in which a stable fusion plasma is to be generated. This is to be the basis for the construction of a fusion reactor for electricity generation. Preparatory work was performed in the Tokamak experiments with JET and TFTR. The fusion power of 1.5 GW will be attained, thus enabling Iter to keep a deuterium-tritium plasma burning. (orig.) [de

  11. Magnetic Fusion Advisory Committee report on recommended fusion program priorities and strategy

    International Nuclear Information System (INIS)

    1983-09-01

    The Magnetic Fusion Advisory Committee recommends a new program strategy with the following principal features: (1) Initiation in FY86 of the Tokamak Fusion Core Experiment (TFCX), a moderate-cost tokamak reactor device (less than $1 B PACE) designed to achieve ignition and long-pulse equilibrium burn. Careful trade-off studies are needed before making key design choices in interrelated technology areas. Cost reductions relative to earlier plans can be realized by exploiting new plasma technology, by locating the TFCX at the TFTR site, and by assigning responsibility for complementary reactor engineering tasks to other sectors of the fusion program. (2) Potential utilization of the MFTF Upgrade to provide a cost-effective means for quasi-steady-state testing of blanket and power-system components, complementary to TFCX. This will depend on future assessments of the data base for tandem mirrors. (3) Vigorous pursuit of the broad US base program in magnetic confinement, including new machine starts, where appropriate, at approximately the present total level of support. (4) Utilization of Development and Technology programs in plasma and magnet technology in support of specific hardware requirements of the TFCX and of other major fusion facilities, so as to minimize overall program cost

  12. Fusion core start-up, ignition, and burn simulations of reversed-field pinch (RFP) reactors

    International Nuclear Information System (INIS)

    Chu, Y.Y.

    1988-01-01

    A transient reactor simulation model is developed to investigate and simulate the start-up, ignition, and burn of a reversed-field pinch reactor. The simulation is based upon a spatially averaged plasma balance model with field profiles obtained from MHD quasi-equilibrium analysis. Alpha particle heating is estimated from Fokker-Planck calculations. The instantaneous plasma current is derived from a self-consistent circuit analysis for plasma/coil/eddy current interactions. The simulation code is applied to the TITAN RFP reactor design which features a compact, high-power-density reversed-field pinch fusion system. A contour analysis is performed using the steady-state global plasma balance. The results are presented with contours of constant plasma current. A saddle point is identified in the contour plot which determined the minimum value of plasma current required to achieve ignition. In the simulations of the TITAN RFP reactor, the OH-driven super-conducting EF coils are found to deviate from the required equilibrium values as the induced plasma current increases. A set of basic results from the simulation of TITAN RFP reactor yield a picture of RFP plasma operation in a reactor. Investigations of eddy currents are also presented and have very important in reactor design

  13. Fusion reactor critical issues

    International Nuclear Information System (INIS)

    1987-11-01

    The document summarizes the results of a series of INTOR-related meetings organized by the IAEA in 1985-1986 with the following topics: Impurity control modelling, non-inductive current-drive, confinement in tokamaks with intense heating and DEMO requirements. These results are useful to the specialists involved in research on large fusion machines or in the design activity on the next generation tokamaks. Refs, figs and tabs

  14. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

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

  15. Results of nuclear fusion development

    International Nuclear Information System (INIS)

    Yamamoto, Kenzo

    1975-01-01

    Compared with the nuclear fission research which followed that in advanced countries, Japan has treaded on its own track in nuclear fusion development; in the former, she had been far behind other leading countries. Characteristic of the efforts in Japan is the collaboration with educational institutions. Works are now carried out mainly in Tokamak plasma confinement, though other means being studied simultaneously. The nation's fusion research program is the realization of a fusion reactor at the turn of the present century, based on the world-level results attained with Tokamak. Past developments in the nuclear fusion research, the current status, and aspects for the future are discribed. (Mori, K.)

  16. THE GENERAL ATOMICS FUSION THEORY PROGRAM ANNUAL REPORT FOR GRANT YEAR 2004

    International Nuclear Information System (INIS)

    PROJECT STAFF

    2004-01-01

    The dual objective of the fusion theory program at General Atomics (GA) is to significantly advance our scientific understanding of the physics of fusion plasmas and to support the DIII-D and other tokamak experiments. The program plan is aimed at contributing significantly to the Fusion Energy Science and the Tokamak Concept Improvement goals of the Office of Fusion Energy Sciences (OFES)

  17. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  18. Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

    Science.gov (United States)

    Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

    2018-02-01

    A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

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

    Science.gov (United States)

    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

  20. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.