WorldWideScience

Sample records for fusion target designs

  1. Fusion target design

    International Nuclear Information System (INIS)

    Bangerter, R.O.

    1978-01-01

    Most detailed fusion target design is done by numerical simulation using large computers. Although numerical simulation is briefly discussed, this lecture deals primarily with the way in which basic physical arguments, driver technology considerations and economical power production requirements are used to guide and augment the simulations. Physics topics discussed include target energetics, preheat, stability and symmetry. A specific design example is discussed

  2. Target design for heavy ion beam fusion

    International Nuclear Information System (INIS)

    Meyer-ter-Vehn, J.; Metzler, N.

    1981-07-01

    Target design for Heavy Ion Beam Fusion and related physics are discussed. First, a modified version of the Kidder-Bodner model for pellet gain is presented and is used to define the working point (Esub(beam) = 4.8 MJ, Gain 83) for a reactor size target. Secondly, stopping of heavy ions in hot dense plasma is investigated and numerical results for stopping powers and ranges of 10 GeV Bi-ions in Pb, Li, and PbLi-alloy are given. Finally, results of an explicit implosion calculation, using the 1-D code MINIHY, are discussed in detail. The hydrodynamic efficiency is found to be about 5%. Special attention is given to the shock sequence leading to the ignition configuration. Also the growth of Rayleigh-Taylor instability at the absorber-pusher interface is estimated. (orig.)

  3. Charged particle fusion targets

    International Nuclear Information System (INIS)

    Bangerter, R.O.; Meeker, D.J.

    1977-01-01

    The power, voltage, energy and other requirements of electron and ion beam fusion targets are reviewed. Single shell, multiple shell and magnetically insulated target designs are discussed. Questions of stability are also considered. In particular, it is shown that ion beam targets are stabilized by an energy spread in the ion beam

  4. Muon-catalyzed fusion experiment target and detector system. Preliminary design report

    International Nuclear Information System (INIS)

    Jones, S.E.; Watts, K.D.; Caffrey, A.J.; Walter, J.B.

    1982-03-01

    We present detailed plans for the target and particle detector systems for the muon-catalyzed fusion experiment. Requirements imposed on the target vessel by experimental conditions and safety considerations are delineated. Preliminary designs for the target vessel capsule and secondary containment vessel have been developed which meet these requirements. In addition, the particle detection system is outlined, including associated fast electronics and on-line data acquisition. Computer programs developed to study the target and detector system designs are described

  5. Conceptual design considerations and neutronics of lithium fall laser fusion target chambers

    International Nuclear Information System (INIS)

    Meier, W.R.; Thomson, W.B.

    1978-01-01

    Atomics International and Lawrence Livermore Laboratory are involved in the conceptual design of a laser fusion power plant incorporating the lithium fall target chamber. In this paper we discuss some of the more important design considerations for the target chamber and evaluate its nuclear performance. Sizing and configuration of the fall, hydraulic effects, and mechanical design considerations are addressed. The nuclear aspects examined include tritium breeding, energy deposition, and radiation damage

  6. Design, construction, and characterization of high-performance membrane fusion devices with target-selectivity.

    Science.gov (United States)

    Kashiwada, Ayumi; Yamane, Iori; Tsuboi, Mana; Ando, Shun; Matsuda, Kiyomi

    2012-01-31

    Membrane fusion proteins such as the hemagglutinin glycoprotein have target recognition and fusion accelerative domains, where some synergistically working elements are essential for target-selective and highly effective native membrane fusion systems. In this work, novel membrane fusion devices bearing such domains were designed and constructed. We selected a phenylboronic acid derivative as a recognition domain for a sugar-like target and a transmembrane-peptide (Leu-Ala sequence) domain interacting with the target membrane, forming a stable hydrophobic α-helix and accelerating the fusion process. Artificial membrane fusion behavior between the synthetic devices in which pilot and target liposomes were incorporated was characterized by lipid-mixing and inner-leaflet lipid-mixing assays. Consequently, the devices bearing both the recognition and transmembrane domains brought about a remarkable increase in the initial rate for the membrane fusion compared with the devices containing the recognition domain alone. In addition, a weakly acidic pH-responsive device was also constructed by replacing three Leu residues in the transmembrane-peptide domain by Glu residues. The presence of Glu residues made the acidic pH-dependent hydrophobic α-helix formation possible as expected. The target-selective liposome-liposome fusion was accelerated in a weakly acidic pH range when the Glu-substituted device was incorporated in pilot liposomes. The use of this pH-responsive device seems to be a potential strategy for novel applications in a liposome-based delivery system. © 2011 American Chemical Society

  7. Directions for reactor target design based on the US heavy ion fusion systems assessment

    International Nuclear Information System (INIS)

    Wilson, D.C.; Dudziak, D.; Magelssen, G.; Zuckerman, D.; Dreimeyer, D.

    1986-01-01

    We studied areas of major uncertainty in target design using the cost of electricity as our figure of merit. Net electric power from the plant was fixed at 1000 MW to eliminate large effects due to economies of scale. The system is relatively insensitive to target gain. Factors of three changes in gain cause only 8 to 12% changes in electricity cost. An increase in the peak power needed to drive targets poses only a small cost risk, but requires many more beamlets be transported to the target. A shortening of the required ion range causes both cost and beamlet difficulties. A factor of 4 decrease in the required range at a fixed driver energy increases electricity cost by 44% and raises the number of beamlets to 240. Finally, the heavy ion fusion system can accommodate large increases in target costs. To address the major uncertainties, target design should concentrate on the understanding requirements for ion range and peak driver power

  8. Design and fabrication of foam-insulated cryogenic target for wet-wall laser fusion reactor

    International Nuclear Information System (INIS)

    Norimatsu, T.; Takeda, T.; Nagai, K.; Mima, K.; Yamanaka, T.

    2003-01-01

    A foam insulated cryogenic target was proposed for use in a future laser fusion reactor with a wet wall. This scheme can protect the solid DT layer from melting due to surface heating by adsorption of metal vapor without significant reduction in the target gain. Design spaces for the injection velocity and the acceptable vapor pressure in the reactor are discussed. Basic technology to fabricate such structure was demonstrated by emulsion process. Concept of a cryogenic fast-ignition target with a gold guiding cone was proposed together with direct injection filling of liquid DT. (author)

  9. Target design for high fusion yield with the double Z-pinch-driven hohlraum

    International Nuclear Information System (INIS)

    Vesey, R. A.; Herrmann, M. C.; Lemke, R. W.; Desjarlais, M. P.; Cuneo, M. E.; Stygar, W. A.; Bennett, G. R.; Campbell, R. B.; Christenson, P. J.; Mehlhorn, T. A.; Porter, J. L.; Slutz, S. A.

    2007-01-01

    A key demonstration on the path to inertial fusion energy is the achievement of high fusion yield (hundreds of MJ) and high target gain. Toward this goal, an indirect-drive high-yield inertial confinement fusion (ICF) target involving two Z-pinch x-ray sources heating a central secondary hohlraum is described by Hammer et al. [Phys. Plasmas 6, 2129 (1999)]. In subsequent research at Sandia National Laboratories, theoretical/computational models have been developed and an extensive series of validation experiments have been performed to study hohlraum energetics, capsule coupling, and capsule implosion symmetry for this system. These models have been used to design a high-yield Z-pinch-driven ICF target that incorporates the latest experience in capsule design, hohlraum symmetry control, and x-ray production by Z pinches. An x-ray energy output of 9 MJ per pinch, suitably pulse-shaped, is sufficient for this concept to drive 0.3-0.5 GJ capsules. For the first time, integrated two-dimensional (2D) hohlraum/capsule radiation-hydrodynamics simulations have demonstrated adequate hohlraum coupling, time-dependent radiation symmetry control, and the successful implosion, ignition, and burn of a high-yield capsule in the double Z-pinch hohlraum. An important new feature of this target design is mode-selective symmetry control: the use of burn-through shields offset from the capsule that selectively tune certain low-order asymmetry modes (P 2 ,P 4 ) without significantly perturbing higher-order modes and without a significant energy penalty. This paper will describe the capsule and hohlraum design that have produced 0.4-0.5 GJ yields in 2D simulations, provide a preliminary estimate of the Z-pinch load and accelerator requirements necessary to drive the system, and suggest future directions for target design work

  10. Targets for heavy ion fusion

    International Nuclear Information System (INIS)

    Clauser, M.J.

    1978-01-01

    This paper describes some of the basic principles of fusion target implosions, using some simple targets designed for irradiation by ion beams. Present estimates are that ion beams with 1-5 MJ, and 100-500 TW will be required to ignite high gain targets. (orig.) [de

  11. Reduced cost design of liquid lithium target for international fusion material irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

    2001-01-01

    The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based D-Li neutron source to produce intense high energy neutrons (2 MW/m 2 ) up to 200 dpa and a sufficient irradiation volume (500 cm 3 ) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid lithium flow with a speed of 20 m/s. Following Conceptual Design Activity (1995-1998), a design study with focus on cost reduction without changing its original mission has been done in 1999. The following major changes to the CAD target design have been considered in the study and included in the new design: i) number of the Li target has been changed from 2 to 1, ii) spare of impurity traps of the Li loop was removed although the spare will be stored in a laboratory for quick exchange, iii) building volume was reduced via design changes in lithium loop length. This paper describes the reduced cost design of the lithium target system and recent status of Key Element Technology activities. (author)

  12. High performance inertial fusion targets

    International Nuclear Information System (INIS)

    Nuckolls, J.H.; Bangerter, R.O.; Lindl, J.D.; Mead, W.C.; Pan, Y.L.

    1978-01-01

    Inertial confinement fusion (ICF) target designs are considered which may have very high gains (approximately 1000) and low power requirements (< 100 TW) for input energies of approximately one megajoule. These include targets having very low density shells, ultra thin shells, central ignitors, magnetic insulation, and non-ablative acceleration

  13. Optimized beryllium target design for indirectly driven inertial confinement fusion experiments on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Simakov, Andrei N., E-mail: simakov@lanl.gov; Wilson, Douglas C.; Yi, Sunghwan A.; Kline, John L.; Batha, Steven H. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Clark, Daniel S.; Milovich, Jose L.; Salmonson, Jay D. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States)

    2014-02-15

    For indirect drive inertial confinement fusion, Beryllium (Be) ablators offer a number of important advantages as compared with other ablator materials, e.g., plastic and high density carbon. In particular, the low opacity and relatively high density of Be lead to higher rocket efficiencies giving a higher fuel implosion velocity for a given X-ray drive; and to higher ablation velocities providing more ablative stabilization and reducing the effect of hydrodynamic instabilities on the implosion performance. Be ablator advantages provide a larger target design optimization space and can significantly improve the National Ignition Facility (NIF) [J. D. Lindl et al., Phys. Plasmas 11, 339 (2004)] ignition margin. Herein, we summarize the Be advantages, briefly review NIF Be target history, and present a modern, optimized, low adiabat, Revision 6 NIF Be target design. This design takes advantage of knowledge gained from recent NIF experiments, including more realistic levels of laser-plasma energy backscatter, degraded hohlraum-capsule coupling, and the presence of cross-beam energy transfer.

  14. Recycling issues facing target and RTL materials of inertial fusion designs

    International Nuclear Information System (INIS)

    El-Guebaly, L.; Wilson, P.; Sawan, M.; Henderson, D.; Varuttamaseni, A.

    2005-01-01

    Designers of heavy ion (HI) and Z-pinch inertial fusion power plants have explored the potential of recycling the target and recyclable transmission line (RTL) materials as an alternate option to disposal in a geological repository. This work represents the first time a comprehensive recycling assessment was performed on both machines with an exact pulse history. Our results offer two divergent conclusions on the recycling issue. For the HI concept, target recycling is not a 'must' requirement and the preferred option is the one-shot use scenario as target materials represent a small waste stream, less than 1% of the total nuclear island waste. We recommend using low-cost hohlraum materials once-through and then disposing of them instead of recycling expensive materials such as Au and Gd. On the contrary, RTL recycling is a 'must' requirement for the Z-pinch concept in order to minimize the RTL inventory and enhance the economics. The RTLs meet the low level waste and recycling dose requirements with a wide margin when recycled for the entire plant life even without a cooling period. While recycling offers advantages to the Z-pinch system, it adds complexity and cost to the HI designs

  15. High performance inertial fusion targets

    International Nuclear Information System (INIS)

    Nuckolls, J.H.; Bangerter, R.O.; Lindl, J.D.; Mead, W.C.; Pan, Y.L.

    1977-01-01

    Inertial confinement fusion (ICF) designs are considered which may have very high gains (approximately 1000) and low power requirements (<100 TW) for input energies of approximately one megajoule. These include targets having very low density shells, ultra thin shells, central ignitors, magnetic insulation, and non-ablative acceleration

  16. Electron beam fusion targets

    International Nuclear Information System (INIS)

    Clauser, M.J.; Sweeney, M.A.

    1975-01-01

    R The behavior of the DT filled gold shells when irradiated by a variety of pulse shapes was studied. In these pulses the power (and beam current) was varied, but the voltage was kept constant at 1 MeV. In general the performance of the target, for a given peak power, was not significantly affected by the pulse shape. Pulses with rise times of up to half the implosion time do not significantly degrade the target performance. The use of the ''optimal pulse'' of laser fusion with a fixed peak power does not appear to improve the performance of these targets. The main function of the ''optimal pulse'' is to produce a large rho r of the target during the thermonuclear burn. In e-beam targets a total rho r of 5--10 g/cm 2 can be obtained without pulse shaping; the problem here is one of achieving high enough temperatures to ignite the DT. (U.S.)

  17. Target support for inertial confinement fusion

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1995-08-01

    General Atomics (GA) plays an important industrial support role for the US Inertial Confinement Fusion (ICF) program in the area of target technology. This includes three major activities: target fabrication support, target handling systems development, and target chamber design. The work includes target fabrication for existing ICF experiments, target and target system development for future experiments, and target research and target chamber design for experiments on future machines, such as the National Ignition Facility (NIF)

  18. 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

  19. Inertial confinement fusion target

    International Nuclear Information System (INIS)

    Bourdier, A.

    2001-12-01

    A simple, zero-dimensional model describing the temporal behaviour of an imploding-shell, magnetized fuel inertial confinement fusion target is formulated. The addition of a magnetic field to the fuel reduces thermal conduction losses. As a consequence, it might lead to high gains and reduce the driver requirements. This beneficial effect of the magnetic field on thermonuclear gains is confirmed qualitatively by the zero-dimensional model results. Still, the extent of the initial-condition space for which significant gains can occur is not, by far, as large as previously reported. One-dimensional CEA code simulations which confirm this results are also presented. Finally, we suggest to study the approach proposed by Hasegawa. In this scheme, the laser target is not imploded, and the life-time of the plasma can be very much increased. (author)

  20. Progress towards a high-gain and robust target design for heavy ion fusion

    Energy Technology Data Exchange (ETDEWEB)

    Henestroza, Enrique; Grant Logan, B. [Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States)

    2012-07-15

    Recently [E. Henestroza et al., Phys. Plasmas 18, 032702 (2011)], a new inertial-fusion target configuration, the X-target, using one-sided axial illumination has been explored. This class of target uses annular and solid-profile heavy ion beams to compress and ignite deuterium-tritium (DT) fuel that fills the interior of metal cases that have side-view cross sections in the shape of an 'X.' X-targets using all-DT-filled metal cases imploded by three annular ion beams resulted in fuel densities of {approx}50 g/cm{sup 3} at peak compression, and fusion gains of {approx}50, comparable to heavy ion driven hohlraum targets [D. A. Callahan-Miller and M. Tabak, Phys. Plasmas 7, 2083 (2000)]. This paper discusses updated X-target configurations that incorporate inside the case a propellant (plastic) and a pusher (aluminum) surrounding the DT fuel. The updated configurations are capable of assembling higher fuel areal densities {approx}2 g/cm{sup 2} using two annular beams to implode the target to peak DT densities {approx}100 g/cm{sup 3}, followed by a fast-ignition solid ion beam which heats the high-density fuel to thermonuclear temperatures in {approx}200 ps to start the burn propagation, obtaining gains of {approx}300. These targets have been modeled using the radiation-hydrodynamics code HYDRA [M. M. Marinak et al., Phys. Plasmas 8, 2275 (2001)] in two- and three- dimensions to study the properties of the implosion as well as the ignition and burn propagation phases. At typical Eulerian mesh resolutions of a few microns, the aluminum-DT interface shows negligible Rayleigh-Taylor (RT) and Richtmyer-Meshkov instability growth; also, the shear flow of the DT fuel as it slides along the metal X-target walls, which drives the RT and Kelvin Helmholtz instabilities, does not have a major effect on the burning rate. An analytic estimate of the RT instability process at the Al-DT interface shows that the aluminum spikes generated during the pusher deceleration phase

  1. Inertial fusion reactor designs

    International Nuclear Information System (INIS)

    Meier, W.

    1987-01-01

    In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

  2. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    Science.gov (United States)

    Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.

    2001-05-01

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si

  3. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G.; Perlado, J.M. E-mail: mperlado@denim.upm.es; Alonso, E.; Alonso, M.; Dominguez, E.; Rubiano, J.G.; Gil, J.M.; Gomez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martinez-Val, J.M.; Minguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P

    2001-05-21

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT{sub x} fuel with a small tritium initial content (x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures ({>=}100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower T{sub e} and to enhance radiation losses, reducing the plasma temperature, T{sub i}. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination

  4. Inertial-confinement-fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1981-01-01

    Inertial confinement fusion (ICF) targets are made as simple flat discs, as hollow shells or as complicated multilayer structures. Many techniques have been devised for producing the targets. Glass and metal shells are made by using drop and bubble techniques. Solid hydrogen shells are also produced by adapting old methods to the solution of modern problems. Some of these techniques, problems and solutions are discussed. In addition, the applications of many of the techniques to fabrication of ICF targets is presented

  5. Target production for inertial fusion energy

    International Nuclear Information System (INIS)

    Woodworth, J.G.; Meier, W.

    1995-03-01

    Inertial fusion energy (IFE) power plants will require the ignition and burn of 5-10 fusion fuel targets every second. The technology to economically mass produce high-quality, precision targets at this rate is beyond the current state of the art. Techniques that are scalable to high production rates, however, have been identified for all the necessary process steps, and many have been tested in laboratory experiments or are similar to current commercial manufacturing processes. In this paper, we describe a baseline target factory conceptual design and estimate its capital and operating costs. The result is a total production cost of ∼16 cents per target. At this level, target production represents about 6% of the estimated cost of electricity from a 1-GW e IFE power plant. Cost scaling relationships are presented and used to show the variation in target cost with production rate and plant power level

  6. Implosion physics, alternative targets design and neutron effects on inertial fusion systems

    International Nuclear Information System (INIS)

    Velarde, G.; Martinez-Val, J.M.; Perlado, J.M.

    2001-01-01

    A new radiation transport code has been coupled with an existing multimaterial fluidynamics code using Adaptive Mesh Refinement (AMR) and its testing is presented, solving ray effect and shadow problems in SN classical methods. Important advances in atomic physics, opacity calculations and NLTE calculations, participating in significant experiments (LULI/France), have been obtained. Our new 1D target simulation model allows considering the effect of inverse Compton scattering in DT x targets (x<3%) working in a catalytic regime, showing the effectiveness of such tritium-less targets. Neutron activation of all natural elements in IFE reactors for waste management and that of target debris in NIF-type facilities have been completed. Pulse activation in structural walls is presented with a new modeling. Tritium atmospheric dispersion results indicate large uncertainties in environmental responses and needs to treat the two chemical forms. We recognise recombination barriers (metastable defects) and compute first systematic high-energy displacement cascade analysis in SiC, and radiation damage pulses by atomistic models in metals. Using Molecular Dynamics we explain the experimental evidence of low-temperature amorphization by damage accumulation in SiC. (author)

  7. Conceptual design and issues of the laser inertial fusion test (LIFT) reactor—targets and chamber systems

    Science.gov (United States)

    Norimatsu, T.; Kozaki, Y.; Shiraga, H.; Fujita, H.; Okano, K.; Members of LIFT Design Team

    2017-11-01

    We present the conceptual design of an experimental laser fusion plant known as the laser inertial fusion test (LIFT) reactor. The conceptual design aims at technically connecting a single-shot experiment and a commercial power plant. The LIFT reactor is designed on a three-phase scheme, where each phase has specific goals and the dedicated chambers of each phase are driven by the same laser. Technical issues related to the chamber technology including radiation safety to repeat burst mode operation are discussed in this paper.

  8. Reactor potential for magnetized target fusion

    International Nuclear Information System (INIS)

    Dahlin, J.E.

    2001-06-01

    Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well

  9. Reactor potential for magnetized target fusion

    Energy Technology Data Exchange (ETDEWEB)

    Dahlin, J.E

    2001-06-01

    Magnetized Target Fusion (MTF) is a possible pathway to thermonuclear fusion different from both magnetic fusion and inertial confinement fusion. An imploding cylindrical metal liner compresses a preheated and magnetized plasma configuration until thermonuclear conditions are achieved. In this report the Magnetized Target Fusion concept is evaluated and a zero-dimensional computer model of the plasma, liner and circuit as a connected system is designed. The results of running this code are that thermonuclear conditions are achieved indeed, but only during a very short time. At peak compression the pressure from the compressed plasma and magnetic field is so large reversing the liner implosion into an explosion. The time period of liner motion reversal is termed the dwell time and is crucial to the performance of the fusion system. Parameters as liner thickness and plasma density are certainly of significant importance to the dwell time, but it seems like a reactor based on the MTF principle hardly can become economic if not innovative solutions are introduced. In the report two such solutions are presented as well.

  10. Design study of laser fusion rocket

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Shoyama, Hidetoshi; Kanda, Yukinori

    1991-01-01

    A design study was made on a rocket powered by laser fusion. Dependence of its flight performance on target gain, driver repetition rate and fuel composition was analyzed to obtain optimal design parameters of the laser fusion rocket. The results indicate that the laser fusion rocket fueled with DT or D 3 He has the potential advantages over other propulsion systems such as fission rocket for interplanetary travel. (author)

  11. Designing the Sniper: Improving Targeted Human Cytolytic Fusion Proteins for Anti-Cancer Therapy via Molecular Simulation

    Directory of Open Access Journals (Sweden)

    Anna Bochicchio

    2017-02-01

    Full Text Available Targeted human cytolytic fusion proteins (hCFPs are humanized immunotoxins for selective treatment of different diseases including cancer. They are composed of a ligand specifically binding to target cells genetically linked to a human apoptosis-inducing enzyme. hCFPs target cancer cells via an antibody or derivative (scFv specifically binding to e.g., tumor associated antigens (TAAs. After internalization and translocation of the enzyme from endocytosed endosomes, the human enzymes introduced into the cytosol are efficiently inducing apoptosis. Under in vivo conditions such enzymes are subject to tight regulation by native inhibitors in order to prevent inappropriate induction of cell death in healthy cells. Tumor cells are known to upregulate these inhibitors as a survival mechanism resulting in escape of malignant cells from elimination by immune effector cells. Cytosolic inhibitors of Granzyme B and Angiogenin (Serpin P9 and RNH1, respectively, reduce the efficacy of hCFPs with these enzymes as effector domains, requiring detrimentally high doses in order to saturate inhibitor binding and rescue cytolytic activity. Variants of Granzyme B and Angiogenin might feature reduced affinity for their respective inhibitors, while retaining or even enhancing their catalytic activity. A powerful tool to design hCFPs mutants with improved potency is given by in silico methods. These include molecular dynamics (MD simulations and enhanced sampling methods (ESM. MD and ESM allow predicting the enzyme-protein inhibitor binding stability and the associated conformational changes, provided that structural information is available. Such “high-resolution” detailed description enables the elucidation of interaction domains and the identification of sites where particular point mutations may modify those interactions. This review discusses recent advances in the use of MD and ESM for hCFP development from the viewpoints of scientists involved in both fields.

  12. Inertial-confinement-fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1982-01-01

    Much of the research in laser fusion has been done using simple ball on-stalk targets filled with a deuterium-tritium mixture. The targets operated in the exploding pusher mode in which the laser energy was delivered in a very short time (approx. 100 ps or less) and was absorbed by the glass wall of the target. The high energy density in the glass literally exploded the shell with the inward moving glass compressing the DT fuel to high temperatures and moderate densities. Temperatures achieved were high enough to produce DT reactions and accompanying thermonuclear neutrons and alpha particles. The primary criteria imposed on the target builders were: (1) wall thickness, (2) sphere diameter, and (3) fuel in the sphere

  13. Fusion engineering device design description

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-12-01

    The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein.

  14. Fusion Engineering Device design description

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-12-01

    The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein

  15. Fusion engineering device design description

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-12-01

    The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein

  16. Development of lithium target system in engineering validation and engineering design activity of the international fusion materials irradiation facility (IFMIF/EVEDA)

    International Nuclear Information System (INIS)

    Wakai, Eiichi; Kondo, Hiroo; Sugimoto, Masayoshi; Ida, Mizuho; Kanemura, Takuji; Watanabe, Kazuyoshi; Fujishiro, Kouji; Edao, Yuuki; Niitsuma, Shigeto; Kimura, Haruyuki; Fukada, Satoshi; Hiromoto, Tetsushi; Shigeharu, Satoshi; Yagi, Jyuro; Furukawa, Tomohiro; Hirakawa, Yasushi; Suzuki, Akihiro; Terai, Takayuki; Horiike, Hiroshi; Hoashi, Eiji; Suzuki, Sachiko; Yamaoka, Nobuo; Serizawa, Hisashi; Kawahito, Yosuke; Tsuji, Yoshiyuki; Furuya, Kazuyuki; Takeo, Fumio

    2012-01-01

    Engineering validation and engineering design activity (EVEDA) for the international fusion materials irradiation facility (IFMIF) has been conducted since 2007. Research and development of the Lithium target facility is an important part of this activity. We constructed a world largest liquid Lithium test loop with a capacity of 5000 L in 2010 and successfully completed the first stage validation tests (functional tests of components and Lithium flow test (flow velocity 15 m/s at the target). In the present article, recent results of the EVEDA activity for the Lithium target facility and related technologies on liquid Lithium are reviewed. (author)

  17. Electron Shock Ignition of Inertial Fusion Targets

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  18. Design of a distributed radiator target for inertial fusion driven from two sides with heavy ion beams

    International Nuclear Information System (INIS)

    Tabak, M.; Callahan-Miller, D.

    1997-01-01

    We describe the status of a distributed radiator heavy ion target design. In integrated calculations this target ignited and produced 390-430 MJ of yieldwhen driven with 5.8-6.5 MJ of 3-4 GeV Pb ions. The target has cylindrical symmetry with disk endplates. The ions uniformly illuminate these endplates in a 5mm radius spot. We discuss the considerations which led to this design together with some previously unused design features: low density hohlraum walls in approximate pressure balance with internal low-Z fill materials, radiationsymmetry determined by the position of the radiator materials and particle ranges, and early time pressure symmetry possibly influenced by radiation shims. We discuss how this target scales to lower input energy or to lower beam power. Variant designs with more realistic beam focusing strategies are also discussed. We show the tradeoffs required for targets which accept higher particle energies

  19. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

    Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  20. SOLASE: a conceptual laser fusion reactor design

    International Nuclear Information System (INIS)

    Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

    1977-12-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion ractors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

  1. Laser fusion target fabrication. Status report, 30 April 1974

    International Nuclear Information System (INIS)

    Fries, R.J.; Farnum, E.H.

    1974-11-01

    The laser fusion target fabrication effort at Los Alamos Scientific Laboratory has been successful in producing targets of the general design requested by, and with a range of parameters acceptable to, the theoretical designers and to the laser/target interaction physics experimentalists. Many novel techniques for handling and measuring the properties of various types of hollow microballoons were developed. (U.S.)

  2. Mass Producing Targets for Nuclear Fusion

    Science.gov (United States)

    Wang, T. G.; Elleman, D. D.; Kendall, J. M.

    1983-01-01

    Metal-encapsulating technique advances prospects of controlling nuclear fusion. Prefilled fusion targets form at nozzle as molten metal such as tin flows through outer channel and pressurized deuterium/tritium gas flows through inner channel. Molten metal completely encloses gas charge as it drops off nozzle.

  3. Automated laser fusion target production concept

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1977-01-01

    A target production concept is described for the production of multilayered cryogenic spherical inertial confinement fusion targets. The facility is to deliver targets to the reactor chamber at rates up to 10 per second and at costs consistent with economic production of power

  4. Ion-driver fast ignition: Reducing heavy-ion fusion driver energy and cost, simplifying chamber design, target fab, tritium fueling and power conversion

    Energy Technology Data Exchange (ETDEWEB)

    Logan, G.; Callahan-Miller, D.; Perkins, J.; Caporaso, G.; Tabak, M.; Moir, R.; Meier, W.; Bangerter, Roger; Lee, Ed

    1998-04-01

    Ion fast ignition, like laser fast ignition, can potentially reduce driver energy for high target gain by an order of magnitude, while reducing fuel capsule implosion velocity, convergence ratio, and required precisions in target fabrication and illumination symmetry, all of which should further improve and simplify IFE power plants. From fast-ignition target requirements, we determine requirements for ion beam acceleration, pulse-compression, and final focus for advanced accelerators that must be developed for much shorter pulses and higher voltage gradients than today's accelerators, to deliver the petawatt peak powers and small focal spots ({approx}100 {micro}m) required. Although such peak powers and small focal spots are available today with lasers, development of such advanced accelerators is motivated by the greater likely efficiency of deep ion penetration and deposition into pre-compressed 1000x liquid density DT cores. Ion ignitor beam parameters for acceleration, pulse compression, and final focus are estimated for two examples based on a Dielectric Wall Accelerator; (1) a small target with {rho}r {approx} 2 g/cm{sup 2} for a small demo/pilot plant producing {approx}40 MJ of fusion yield per target, and (2) a large target with {rho}r {approx} 10 g/cm{sup 2} producing {approx}1 GJ yield for multi-unit electricity/hydrogen plants, allowing internal T-breeding with low T/D ratios, >75 % of the total fusion yield captured for plasma direct conversion, and simple liquid-protected chambers with gravity clearing. Key enabling development needs for ion fast ignition are found to be (1) ''Close-coupled'' target designs for single-ended illumination of both compressor and ignitor beams; (2) Development of high gradient (>25 MV/m) linacs with high charge-state (q {approx} 26) ion sources for short ({approx}5 ns) accelerator output pulses; (3) Small mm-scale laser-driven plasma lens of {approx}10 MG fields to provide steep focusing angles

  5. Ion-driver fast ignition: Reducing heavy-ion fusion driver energy and cost, simplifying chamber design, target fab, tritium fueling and power conversion

    International Nuclear Information System (INIS)

    Logan, G.; Callahan-Miller, D.; Perkins, J.; Caporaso, G.; Tabak, M.; Moir, R.; Meier, W.; Bangerter, Roger; Lee, Ed

    1998-01-01

    Ion fast ignition, like laser fast ignition, can potentially reduce driver energy for high target gain by an order of magnitude, while reducing fuel capsule implosion velocity, convergence ratio, and required precisions in target fabrication and illumination symmetry, all of which should further improve and simplify IFE power plants. From fast-ignition target requirements, we determine requirements for ion beam acceleration, pulse-compression, and final focus for advanced accelerators that must be developed for much shorter pulses and higher voltage gradients than today's accelerators, to deliver the petawatt peak powers and small focal spots (∼100 (micro)m) required. Although such peak powers and small focal spots are available today with lasers, development of such advanced accelerators is motivated by the greater likely efficiency of deep ion penetration and deposition into pre-compressed 1000x liquid density DT cores. Ion ignitor beam parameters for acceleration, pulse compression, and final focus are estimated for two examples based on a Dielectric Wall Accelerator; (1) a small target with ρr ∼ 2 g/cm 2 for a small demo/pilot plant producing ∼40 MJ of fusion yield per target, and (2) a large target with ρr ∼ 10 g/cm 2 producing ∼1 GJ yield for multi-unit electricity/hydrogen plants, allowing internal T-breeding with low T/D ratios, >75 % of the total fusion yield captured for plasma direct conversion, and simple liquid-protected chambers with gravity clearing. Key enabling development needs for ion fast ignition are found to be (1) ''Close-coupled'' target designs for single-ended illumination of both compressor and ignitor beams; (2) Development of high gradient (>25 MV/m) linacs with high charge-state (q ∼ 26) ion sources for short (∼5 ns) accelerator output pulses; (3) Small mm-scale laser-driven plasma lens of ∼10 MG fields to provide steep focusing angles close-in to the target (built-in as part of each target); (4) beam space charge

  6. Environmental release targets for fusion power plants

    International Nuclear Information System (INIS)

    Gulden, W.; Raskob, W.

    2005-01-01

    Within the European fission community, so called European Utility Requirements were developed to define common targets, criteria and evaluation methods for, amongst others, safety, environmental protection and public health with respect to future nuclear fission power plant development. In the case of severe accidents, the objective is to restrict the radiological consequences to the vicinity of the plant, i.e., to avoid early and late countermeasures such as evacuation or relocation of the population, and to restrict food banning to small areas and the first year after the accident. Within the European Fusion Technology Programme, a methodology is being developed in compliance with these European Utility Requirements, to define design requirements for future fusion reactors. First results are presented. Concerning food banning, calculations revealed extremely conservative values for tritium in EU regulations and recommendations. This does not affect assessments for fission reactors, but is an overestimation of the tritium dose impact from ingestion. Therefore, in compliance with scientific justification, considerably higher maximum permissible activity levels for tritium should be considered

  7. Dim target detection method based on salient graph fusion

    Science.gov (United States)

    Hu, Ruo-lan; Shen, Yi-yan; Jiang, Jun

    2018-02-01

    Dim target detection is one key problem in digital image processing field. With development of multi-spectrum imaging sensor, it becomes a trend to improve the performance of dim target detection by fusing the information from different spectral images. In this paper, one dim target detection method based on salient graph fusion was proposed. In the method, Gabor filter with multi-direction and contrast filter with multi-scale were combined to construct salient graph from digital image. And then, the maximum salience fusion strategy was designed to fuse the salient graph from different spectral images. Top-hat filter was used to detect dim target from the fusion salient graph. Experimental results show that proposal method improved the probability of target detection and reduced the probability of false alarm on clutter background images.

  8. Spinning targets for laser fusion

    International Nuclear Information System (INIS)

    Baldwin, D.E.; Ryutov, D.D.

    1995-09-01

    Several techniques for spinning the ICF targets up prior to or in the course of their compression are suggested. Interference of the rotational shear flow with Rayleigh-Taylor instability is briefly discussed and possible consequences for the target performance are pointed out

  9. Implosion scaling and hydro dynamically equivalent target design - Strategy for proof of principle of high gain inertial fusion

    International Nuclear Information System (INIS)

    Murakami, M.; Nishihara, K.; Azechi, H.; Nakatsuka, M.; Kanabe, T.; Miyanaga, N.

    2003-01-01

    Scaling laws for hydro dynamically similar implosions are derived by applying Lie group analysis to the set of partial differential equations for the hydrodynamic system. Physically this implies that any fluid system belonging to a common similarity group evolves quite in the same manner including hydrodynamic instabilities. The scalings strongly depend on the description of the energy transport, i.e., whether the fluid system is heat conductive or adiabatic. Under a fully specified group transformation including prescriptions on the laser wavelength and the ionization state, the hydrodynamic similarity can still be preserved even when the system is cooperated with such other energy sources as classical laser absorption, hot electrons, local alpha heating, and bremsstrahlung loss. The results are expected to give the basis of target design and diagnostics for scaled high gain experiments in future. (author)

  10. Covariance descriptor fusion for target detection

    Science.gov (United States)

    Cukur, Huseyin; Binol, Hamidullah; Bal, Abdullah; Yavuz, Fatih

    2016-05-01

    Target detection is one of the most important topics for military or civilian applications. In order to address such detection tasks, hyperspectral imaging sensors provide useful images data containing both spatial and spectral information. Target detection has various challenging scenarios for hyperspectral images. To overcome these challenges, covariance descriptor presents many advantages. Detection capability of the conventional covariance descriptor technique can be improved by fusion methods. In this paper, hyperspectral bands are clustered according to inter-bands correlation. Target detection is then realized by fusion of covariance descriptor results based on the band clusters. The proposed combination technique is denoted Covariance Descriptor Fusion (CDF). The efficiency of the CDF is evaluated by applying to hyperspectral imagery to detect man-made objects. The obtained results show that the CDF presents better performance than the conventional covariance descriptor.

  11. General software design for multisensor data fusion

    Science.gov (United States)

    Zhang, Junliang; Zhao, Yuming

    1999-03-01

    In this paper a general method of software design for multisensor data fusion is discussed in detail, which adopts object-oriented technology under UNIX operation system. The software for multisensor data fusion is divided into six functional modules: data collection, database management, GIS, target display and alarming data simulation etc. Furthermore, the primary function, the components and some realization methods of each modular is given. The interfaces among these functional modular relations are discussed. The data exchange among each functional modular is performed by interprocess communication IPC, including message queue, semaphore and shared memory. Thus, each functional modular is executed independently, which reduces the dependence among functional modules and helps software programing and testing. This software for multisensor data fusion is designed as hierarchical structure by the inheritance character of classes. Each functional modular is abstracted and encapsulated through class structure, which avoids software redundancy and enhances readability.

  12. Magnetized Target Fusion At General Fusion: An Overview

    Science.gov (United States)

    Laberge, Michel; O'Shea, Peter; Donaldson, Mike; Delage, Michael; Fusion Team, General

    2017-10-01

    Magnetized Target Fusion (MTF) involves compressing an initial magnetically confined plasma on a timescale faster than the thermal confinement time of the plasma. If near adiabatic compression is achieved, volumetric compression of 350X or more of a 500 eV target plasma would achieve a final plasma temperature exceeding 10 keV. Interesting fusion gains could be achieved provided the compressed plasma has sufficient density and dwell time. General Fusion (GF) is developing a compression system using pneumatic pistons to collapse a cavity formed in liquid metal containing a magnetized plasma target. Low cost driver, straightforward heat extraction, good tritium breeding ratio and excellent neutron protection could lead to a practical power plant. GF (65 employees) has an active plasma R&D program including both full scale and reduced scale plasma experiments and simulation of both. Although pneumatic driven compression of full scale plasmas is the end goal, present compression studies use reduced scale plasmas and chemically accelerated aluminum liners. We will review results from our plasma target development, motivate and review the results of dynamic compression field tests and briefly describe the work to date on the pneumatic driver front.

  13. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  14. COST-EFFECTIVE TARGET FABRICATION FOR INERTIAL FUSION ENERGY

    International Nuclear Information System (INIS)

    GOODIN, D.T; NOBILE, A; SCHROEN, D.G; MAXWELL, J.L; RICKMAN, W.S

    2004-03-01

    A central feature of an Inertial Fusion Energy (IFE) power plant is a target that has been compressed and heated to fusion conditions by the energy input of the driver. The IFE target fabrication programs are focusing on methods that will scale to mass production, and working closely with target designers to make material selections that will satisfy a wide range of required and desirable characteristics. Targets produced for current inertial confinement fusion experiments are estimated to cost about $2500 each. Design studies of cost-effective power production from laser and heavy-ion driven IFE have found a cost requirement of about $0.25-0.30 each. While four orders of magnitude cost reduction may seem at first to be nearly impossible, there are many factors that suggest this is achievable. This paper summarizes the paradigm shifts in target fabrication methodologies that will be needed to economically supply targets and presents the results of ''nth-of-a-kind'' plant layouts and concepts for IFE power plant fueling. Our engineering studies estimate the cost of the target supply in a fusion economy, and show that costs are within the range of commercial feasibility for laser-driven and for heavy ion driven IFE

  15. Super liquid density target designs

    International Nuclear Information System (INIS)

    Pan, Y.L.; Bailey, D.S.

    1976-01-01

    The success of laser fusion depends on obtaining near isentropic compression of fuel to very high densities and igniting this fuel. To date, the results of laser fusion experiments have been based mainly on the exploding pusher implosion of fusion capsules consisting of thin glass microballoons (wall thickness of less than 1 micron) filled with low density DT gas (initial density of a few mg/cc). Maximum DT densities of a few tenths of g/cc and temperatures of a few keV have been achieved in these experiments. We will discuss the results of LASNEX target design calculations for targets which: (a) can compress fuel to much higher densities using the capabilities of existing Nd-glass systems at LLL; (b) allow experimental measurement of the peak fuel density achieved

  16. Design of the FMIT lithium target

    International Nuclear Information System (INIS)

    Hassberger, J.A.; Annese, C.E.; Greenwell, R.K.; Ingham, J.G.; Miles, R.R.; Miller, W.C.

    1981-01-01

    Development of the liquid lithium target for the Fusion Materials Irradiation Test (FMIT) Facility is described. The target concept, major design goals and design requirements are presented. Progress made in the research and development areas leading to detailed design of the target is discussed. This progress, including experimental and analytic results, demonstrates that the FMIT target design is capable of meeting its major design goals and requirements

  17. Magneized target fusion: An overview of the concept

    International Nuclear Information System (INIS)

    Kirkpatrick, R.C.

    1994-01-01

    Magnetized target fusion (MTF) seeks to take advantage of the reduction of thermal conductivity through the application of a strong magneticfield and thereby ease the requirements for reaching fusion conditions in a thermonuclear (TN) fusion fuel. A potentially important benefit of the strong field in the partial trapping of energetic charged particles to enhance energy deposition by the TN fusion reaction products. The essential physics is described. MTF appears to lead to fusion targets that require orders of magnitude less power and intensity for fusion ignition than currently proposed (unmagnetized) inertial confinement fusion (ICF) targets do, making some very energetic pulsed power drivers attractive for realizing controlled fusion

  18. Physical measurements of inertial-fusion targets

    International Nuclear Information System (INIS)

    Weinstein, B.W.

    1981-01-01

    Measurement of inertial-fusion targets has stimulated the development of many new techniques and instruments. This paper reviews the basis for selected target measurement requirements and the development of optical interferometry, optical scattering, microradiography and scanning electron microscopy as applied to target measurement. We summarize the resolution and speed which have been achieved to date, and describe several systems in which these are traded off to fill specific measurement applications. We point out the extent to which present capabilities meet the requirements for target measurement and the key problems which remain to be solved

  19. Realizing Technologies for Magnetized Target Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Wurden, Glen A. [Los Alamos National Laboratory

    2012-08-24

    Researchers are making progress with a range of magneto-inertial fusion (MIF) concepts. All of these approaches use the addition of a magnetic field to a target plasma, and then compress the plasma to fusion conditions. The beauty of MIF is that driver power requirements are reduced, compared to classical inertial fusion approaches, and simultaneously the compression timescales can be longer, and required implosion velocities are slower. The presence of a sufficiently large Bfield expands the accessibility to ignition, even at lower values of the density-radius product, and can confine fusion alphas. A key constraint is that the lifetime of the MIF target plasma has to be matched to the timescale of the driver technology (whether liners, heavy ions, or lasers). To achieve sufficient burn-up fraction, scaling suggests that larger yields are more effective. To handle the larger yields (GJ level), thick liquid wall chambers are certainly desired (no plasma/neutron damage materials problem) and probably required. With larger yields, slower repetition rates ({approx}0.1-1 Hz) for this intrinsically pulsed approach to fusion are possible, which means that chamber clearing between pulses can be accomplished on timescales that are compatible with simple clearing techniques (flowing liquid droplet curtains). However, demonstration of the required reliable delivery of hundreds of MJ of energy, for millions of pulses per year, is an ongoing pulsed power technical challenge.

  20. Acoustically Driven Magnetized Target Fusion At General Fusion: An Overview

    Science.gov (United States)

    O'Shea, Peter; Laberge, M.; Donaldson, M.; Delage, M.; the Fusion Team, General

    2016-10-01

    Magnetized Target Fusion (MTF) involves compressing an initial magnetically confined plasma of about 1e23 m-3, 100eV, 7 Tesla, 20 cm radius, >100 μsec life with a 1000x volume compression in 100 microseconds. If near adiabatic compression is achieved, the final plasma of 1e26 m-3, 10keV, 700 Tesla, 2 cm radius, confined for 10 μsec would produce interesting fusion energy gain. General Fusion (GF) is developing an acoustic compression system using pneumatic pistons focusing a shock wave on the CT plasma in the center of a 3 m diameter sphere filled with liquid lead-lithium. Low cost driver, straightforward heat extraction, good tritium breeding ratio and excellent neutron protection could lead to a practical power plant. GF (65 employees) has an active plasma R&D program including both full scale and reduced scale plasma experiments and simulation of both. Although acoustic driven compression of full scale plasmas is the end goal, present compression studies use reduced scale plasmas and chemically accelerated Aluminum liners. We will review results from our plasma target development, motivate and review the results of dynamic compression field tests and briefly describe the work to date on the acoustic driver front.

  1. Challenges Surrounding the Injection and Arrival of Targets at LIFE Fusion Chamber Center

    Energy Technology Data Exchange (ETDEWEB)

    Miles, R; Spaeth, M; Manes, K; Amendt, P; Tabak, M; Bond, T; Kucheyev, S; Latkowski, J; Loosmore, G; Bliss, E; Baker, K; Bhandarkar, S; Petzoldt, R; Alexander, N; Tillack, M; Holdener, D

    2010-12-01

    IFE target designers must consider several engineering requirements in addition to the physics requirements for successful target implosion. These considerations include low target cost, high manufacturing throughput, the ability of the target to survive the injection into the fusion chamber and arrive in a condition and physical position consistent with proper laser-target interaction and ease of post-implosion debris removal. This article briefly describes these considerations for the Laser Inertial Fusion-based Energy (LIFE) targets currently being designed.

  2. Advances in fusion reactor design

    International Nuclear Information System (INIS)

    Baker, C.C.

    1987-01-01

    The author addresses the tokamak as a power reactor. Contrary to popular opinion, there are still a few people that think a tokamak might make a good fusion power reactor. In thinking about advances in fusion reactor design, in the U.S., at least, that generally means advances relevant to the Starfire design. He reviews some of the features of Starfire. Starfire is the last major study done of the tokamak as a reactor in this country. It is now over eight years old in the sense that eight years ago was really the time in which major decisions were made as to its features. Starfire was a tokamak with a major radius of seven meters, about twice the linear dimensions of a machine like TIBER

  3. Externally guided target for inertial fusion

    International Nuclear Information System (INIS)

    Martinez-Val, J.M.; Piera, M.

    1996-01-01

    A totally new concept is proposed to reach fusion conditions by externally guided inertial confinement. The acceleration and compression of the fuel is guided by a cannon-like external duct with a conical section ending in a small-size cavity around the central point of the tube. The fuel pellets coming from each cannon mouth collide in the central cavity where the implosion and final compression of the fuel take place. Both the tube material density and its areal density must be much higher than the initial density and areal density of the fuel. The external tube will explode into pieces as a consequence of the inner pressures achieved after the fuel central collision. If the collision is suitably driven, a fusion burst can take place before the tube disassembly. because of the features of the central collision needed to trigger ignition, this concept could be considered as tamped impact fusion. Both the fusion products and the debris from the guide tube are caught by a liquid-lithium curtain surrounding the target. Only two driving beams are necessary. The system can be applied to any type of driver and could use a solid pellet at room temperature as the initial target. 54 refs., 24 figs., 1 tab

  4. Design activities of a fusion experimental breeder

    International Nuclear Information System (INIS)

    Huang, J.; Feng, K.; Sheng, G.

    1999-01-01

    The fusion reactor design studies in China are under the support of a fusion-fission hybrid reactor research Program. The purpose of this program is to explore the potential near-term application of fusion energy to support the long-term fusion energy on the one hand and the fission energy development on the other. During 1992-1996 a detailed consistent and integral conceptual design of a Fusion Experimental Breeder, FEB was completed. Beginning from 1996, a further design study towards an Engineering Outline Design of the FEB, FEB-E, has started. The design activities are briefly given. (author)

  5. Design activities of a fusion experimental breeder

    International Nuclear Information System (INIS)

    Huang, J.; Feng, K.; Sheng, G.

    2001-01-01

    The fusion reactor design studies in China are under the support of a fusion-fission hybrid reactor research Program. The purpose of this program is to explore the potential near-term application of fusion energy to support the long-term fusion energy on the one hand and the fission energy development on the other. During 1992-1996 a detailed consistent and integral conceptual design of a Fusion Experimental Breeder, FEB was completed. Beginning from 1996, a further design study towards an Engineering Outline Design of the FEB, FEB-E, has started. The design activities are briefly given. (author)

  6. Electrostatics, small particles, and laser fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1978-01-01

    The success of any Inertial Confinement Fusion system for the production of useful power depends critically on the production of suitable targets. This is true whether the arrangement is that proposed by Nuckolls et al. or some other arrangement. The target must have characteristics such as material composition, structure, and surface finish which are tailored to the laser pulse length, energy, peak and average power and pulse shape. To provide useful power on a continuous basis, it is likely that the repetition rate will be 1.0 to 10 per second. Thus, in a 24 hour running period 864,000 targets may be necessary and one must be placed at the focal point of the laser every tenth of a second. For economic operation it is necessary that the targets be produced at costs of less than $1.00 per target

  7. 21 CFR 886.1880 - Fusion and stereoscopic target.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Fusion and stereoscopic target. 886.1880 Section... (CONTINUED) MEDICAL DEVICES OPHTHALMIC DEVICES Diagnostic Devices § 886.1880 Fusion and stereoscopic target. (a) Identification. A fusion and stereoscopic target is a device intended for use as a viewing object...

  8. A design of steady state fusion burner

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Hatori, Tadatsugu; Itoh, Kimitaka; Ikuta, Takashi; Kodama, Yuji.

    1975-01-01

    We present a brief design of a steady state fusion burner in which a continuous burning of nuclear fuel may be achieved with output power of a gigawatt. The laser fusion is proposed to ignite the fuel. (auth.)

  9. Conceptual design of inertial confinement fusion power plant

    International Nuclear Information System (INIS)

    Mima, Kunioki; Yamanaka, Tatsuhiko; Nakai, Sadao

    1994-01-01

    Presented is the status of the conceptual design studies of inertial confinement fusion reactors. The recent achievements of the laser fusion research enable us to refine the conceptual design of the power plant. In the paper, main features of several new conceptual designs of ICF reactor; KOYO, SIRIUS-P, HYLIFE-II and so on are summarized. In particular, the target design and the reactor chamber design are described. Finally, the overview of the laser fusion reactor and the irradiation system is also described. (author)

  10. Magnetized Target Fusion Driven by Plasma Liners

    Science.gov (United States)

    Thio, Y. C. Francis; Cassibry, Jason; Eskridge, Richard; Kirkpatrick, Ronald C.; Knapp, Charles E.; Lee, Michael; Martin, Adam; Smith, James; Wu, S. T.; Rodgers, Stephen L. (Technical Monitor)

    2001-01-01

    For practical applications of magnetized target fusion, standoff drivers to deliver the imploding momentum flux to the target plasma remotely are required. Quasi-spherically converging plasma jets have been proposed as standoff drivers for this purpose. The concept involves the dynamic formation of a quasi-spherical plasma liner by the merging of plasma jets, and the use of the liner so formed to compress a spheromak or a field reversed configuration (FRC). Theoretical analysis and computer modeling of the concept are presented. It is shown that, with the appropriate choice of the flow parameters in the liner and the target, the impact between the liner and the target plasma can be made to be shockless in the liner or to generate at most a very weak shock in the liner. Additional information is contained in the original extended abstract.

  11. Micromachining of inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Gobby, P.L.; Salzer, L.J.; Day, R.D.

    1996-01-01

    Many experiments conducted on today's largest inertial confinement fusion drive lasers require target components with sub-millimeter dimensions, precisions of a micron or less and surface finishes measured in nanometers. For metal and plastic, techniques using direct machining with diamond tools have been developed that yield the desired parts. New techniques that will be discussed include the quick-flip locator, a magnetically held kinematic mount that has allowed the direct machining of millimeter-sized beryllium hemishells whose inside and outside surface are concentric to within 0.25 micron, and an electronic version of a tracer lathe which has produced precise azimuthal variations of less than a micron

  12. Advances in laser solenoid fusion reactor design

    International Nuclear Information System (INIS)

    Steinhauer, L.C.; Quimby, D.C.

    1978-01-01

    The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

  13. Alternative divertor target concepts for next step fusion devices

    Science.gov (United States)

    Mazul, I. V.

    2016-12-01

    The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

  14. Concepts for fabrication of inertial fusion energy targets

    Energy Technology Data Exchange (ETDEWEB)

    Nobile, A. (Arthur), Jr.; Hoffer, J. K. (James K.); Gobby, P. L. (Peter L.); Steckle, W. P. (Warren P.), Jr.; Goodin, D. T. (Daniel T.); Besenbruch, G. E. (Gottfried E.); Schultz, K. R. (Kenneth R.)

    2001-01-01

    Future inertial fusion energy (IFE) power plants will have a Target Fabrication Facility (TFF) that must produce approximately 500,000 targets per day. To achieve a relatively low cost of electricity, the cost to produce these targets will need to be less than approximately $0.25 per target. In this paper the status on the development of concepts for a TFF to produce targets for a heavy ion fusion (HIF) reactor, such as HYLIFE II, and a laser direct drive fusion reactor such as Sombrero, is discussed. The baseline target that is produced in the HIF TFF is similar to the close-coupled indirect drive target designed by Callahan-Miller and Tabak at Lawrence Livermore Laboratory. This target consists of a cryogenic hohlraum that is made of a metal case and a variety of metal foams and metal-doped organic foams. The target contains a DT-filled CH capsule. The baseline direct drive target is the design developed by Bodner and coworkers at Naval Research Laboratory. HIF targets can be filled with DT before or after assembly of the capsule into the hohlraum. Assembly of targets before filling allows assembly operations to be done at room temperature, but tritium inventories are much larger due to the large volume that the hohlraum occupies in the fill system. Assembly of targets cold after filling allows substantial reduction in tritium inventory, but this requires assembly of targets at cryogenic temperature. A model being developed to evaluate the tritium inventories associated with each of the assembly and fill options indicates that filling targets before assembling the capsule into the hohlraum, filling at temperatures as high as possible, and reducing dead-volumes in the fill system as much as possible offers the potential to reduce tritium inventories to acceptable levels. Use of enhanced DT ice layering techniques, such as infrared layering can reduce tritium inventories significantly by reducing the layering time and therefore the number of capsules being layered

  15. Final report on the Magnetized Target Fusion Collaboration

    Energy Technology Data Exchange (ETDEWEB)

    John Slough

    2009-09-08

    1 meter, and the time from the initiation of formation to the establishment of the final equilibrium was less than 10 microseconds. With some modification, each accelerator was made capable of producing FRCs suitable for the production of the target plasma for the MTF liner experiment. Based on the initial FRC merging/compression results, the design and methodology for an experimental realization of the target plasma for the MTF liner experiment can now be defined. A high density FRC plasmoid is to be formed and accelerated out of each IPA into a merging/compression chamber similar to the imploding liner at AFRL. The properties of the resultant FRC plasma (size, temperature, density, flux, lifetime) are obtained in the reevant regime of interest. The process still needs to be optimized, and a final design for implementation at AFRL must now be carried out. When implemented at AFRL it is anticipated that the colliding/merging FRCs will then be compressed by the liner. In this manner it is hoped that ultimately a plasma with ion temperatures reaching the 10 keV range and fusion gain near unity can be obtained.

  16. High temperature fusion reactor design

    International Nuclear Information System (INIS)

    Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

    1979-01-01

    Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

  17. Plasma processed coating of laser fusion targets

    International Nuclear Information System (INIS)

    Johnson, W.L.; Letts, S.A.; Myers, D.W.; Crane, J.K.; Illige, J.D.; Hatcher, C.W.

    1979-01-01

    Coatings for laser fusion targets have been deposited in an inductively coupled discharge device by plasma polymerization. Two feed gases were used: perfluoro-2-butene, which produced a fluorocarbon coating (CF 1 3 ) with a density of 1.8 g/cc, and trans-2-butene which produced a hydrocarbon coating (CH 1 3 ) with a density of 1.0 g/cc. Uniform pin-hole free films have been deposited to a thickness of up to 30 μm of fluorocarbon and up to 110 μm of hydrocarbon. The effect of process variables on surface smoothness has been investigated. The basic defect in the coating has been found to result from shadowing by a small surface irregularity in an anisotropic coating flux

  18. Premature activation of the paramyxovirus fusion protein before target cell attachment with corruption of the viral fusion machinery.

    Science.gov (United States)

    Farzan, Shohreh F; Palermo, Laura M; Yokoyama, Christine C; Orefice, Gianmarco; Fornabaio, Micaela; Sarkar, Aurijit; Kellogg, Glen E; Greengard, Olga; Porotto, Matteo; Moscona, Anne

    2011-11-04

    Paramyxoviruses, including the childhood pathogen human parainfluenza virus type 3, enter host cells by fusion of the viral and target cell membranes. This fusion results from the concerted action of its two envelope glycoproteins, the hemagglutinin-neuraminidase (HN) and the fusion protein (F). The receptor-bound HN triggers F to undergo conformational changes that render it competent to mediate fusion of the viral and cellular membranes. We proposed that, if the fusion process could be activated prematurely before the virion reaches the target host cell, infection could be prevented. We identified a small molecule that inhibits paramyxovirus entry into target cells and prevents infection. We show here that this compound works by an interaction with HN that results in F-activation prior to receptor binding. The fusion process is thereby prematurely activated, preventing fusion of the viral membrane with target cells and precluding viral entry. This first evidence that activation of a paramyxovirus F can be specifically induced before the virus contacts its target cell suggests a new strategy with broad implications for the design of antiviral agents.

  19. Method for mounting laser fusion targets for irradiation

    Science.gov (United States)

    Fries, R. Jay; Farnum, Eugene H.; McCall, Gene H.

    1977-07-26

    Methods for preparing laser fusion targets of the ball-and-disk type are disclosed. Such targets are suitable for irradiation with one or two laser beams to produce the requisite uniform compression of the fuel material.

  20. Fusion of a polarized projectile with a polarized target

    International Nuclear Information System (INIS)

    Christley, J.A.; Johnson, R.C.; Thompson, I.J.

    1995-01-01

    The fusion cross sections for a polarized target with both unpolarized and polarized projectiles are studied. Expressions for the observables are given for the case when both nuclei are polarized. Calculations for fusion of an aligned 165 Ho target with 16 O and polarized 7 Li beams are presented

  1. Advanced laser fusion target fabrication research and development proposal

    International Nuclear Information System (INIS)

    Stupin, D.M.; Fries, R.J.

    1979-05-01

    A research and development program is described that will enable the fabrication of 10 6 targets/day for a laser fusion prototype power reactor in 2007. We give personnel and cost estimates for a generalized laser fusion target that requires the development of several new technologies. The total cost of the program between 1979 and 2007 is $362 million in today's dollars

  2. Health physics in fusion reactor design

    International Nuclear Information System (INIS)

    Wong, K.Y.; Dinner, P.J.

    1984-06-01

    Experience in the control of tritium exposures to workers and the public gained through the design and operation of Ontario Hydro's nuclear stations has been applied to fusion projects and to design studies on emerging fusion reactor concepts. Ontario Hydro performance in occupational tritium exposure control and environmental impact is reviewed. Application of tritium control technologies and dose management methodology during facility design is highlighted

  3. An Updated Point Design for Heavy Ion Fusion

    International Nuclear Information System (INIS)

    Yu, S.S.; Meier, W.R.; Abbott, R.B.; Barnard, J.J.; Brown, T.; Callahan, D.A.; Heitzenroeder, P.; Latkowski, J.F.; Logan, B.G.; Pemberton, S.J.; Peterson, P.F.; Rose, D.V.; Sabbi, G.L.; Sharp, W.M.; Welch, D.R.

    2002-01-01

    An updated, self-consistent point design for a heavy ion fusion (HIF) power plant based on an induction linac driver, indirect-drive targets, and a thick liquid wall chamber has been completed. Conservative parameters were selected to allow each design area to meet its functional requirements in a robust manner, and thus this design is referred to as the Robust Point Design (RPD-2002). This paper provides a top-level summary of the major characteristics and design parameters for the target, driver, final focus magnet layout and shielding, chamber, beam propagation to the target, and overall power plant

  4. Magnetized Target Fusion Propulsion: Plasma Injectors for MTF Guns

    Science.gov (United States)

    Griffin, Steven T.

    2003-01-01

    To achieve increased payload size and decreased trip time for interplanetary travel, a low mass, high specific impulse, high thrust propulsion system is required. This suggests the need for research into fusion as a source of power and high temperature plasma. The plasma would be deflected by magnetic fields to provide thrust. Magnetized Target Fusion (MTF) research consists of several related investigations into these topics. These include the orientation and timing of the plasma guns and the convergence and interface development of the "pusher" plasma. Computer simulations of the gun as it relates to plasma initiation and repeatability are under investigation. One of the items under development is the plasma injector. This is a surface breakdown driven plasma generator designed to function at very low pressures. The performance, operating conditions and limitations of these injectors need to be determined.

  5. Development of target capsules for muon catalyzed fusion experiments

    International Nuclear Information System (INIS)

    Watts, K.D.; Jones, S.E.; Caffrey, A.J.

    1983-01-01

    A series of Muon Catalyzed Fusion experiments has been conducted at the Los Alamos Meson Physics Facility to determine how many fusion reactions one muon would catalyze under various temperature, pressure, contamination, and tritium concentration conditions. Target capsules to contain deuterium and tritium at elevated temperatures and pressures were engineered for a maximum temperature of 540 K (512 0 F) and a maximum pressure of 103 MPa (15,000 psig). Experimental data collected with these capsules indicated that the number of fusion reactions per muon continued to increase with temperature up to the 540-K design limit. Theory had indicated that the reaction rate should peak at approximately 540 K, but this was not confirmed during the experiments. A second generation of capsules which have a maximum design temperature of 800 K (980 0 F) and a maximum design pressure of 103 MPa (15,000 psig) has now been engineered. These new capsules will be used to further study the muon catalysis rate versus deuterium-tritium mixture temperature

  6. Integrated Chamber Design for the Laser Inertial Fusion Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Latkowski, J F; Kramer, K J; Abbott, R P; Morris, K R; DeMuth, J; Divol, L; El-Dasher, B; Lafuente, A; Loosmore, G; Reyes, S; Moses, G A; Fratoni, M; Flowers, D; Aceves, S; Rhodes, M; Kane, J; Scott, H; Kramer, R; Pantano, C; Scullard, C; Sawicki, R; Wilks, S; Mehl, M

    2010-12-07

    The Laser Inertial Fusion Energy (LIFE) concept is being designed to operate as either a pure fusion or hybrid fusion-fission system. A key component of a LIFE engine is the fusion chamber subsystem. The present work details the chamber design for the pure fusion option. The fusion chamber consists of the first wall and blanket. This integrated system must absorb the fusion energy, produce fusion fuel to replace that burned in previous targets, and enable both target and laser beam transport to the ignition point. The chamber system also must mitigate target emissions, including ions, x-rays and neutrons and reset itself to enable operation at 10-15 Hz. Finally, the chamber must offer a high level of availability, which implies both a reasonable lifetime and the ability to rapidly replace damaged components. An integrated LIFE design that meets all of these requirements is described herein.

  7. Integrated Chamber Design for the Laser Inertial Fusion Energy (LIFE) Engine

    International Nuclear Information System (INIS)

    Latkowski, J.F.; Kramer, K.J.; Abbott, R.P.; Morris, K.R.; DeMuth, J.; Divol, L.; El-Dasher, B.; Lafuente, A.; Loosmore, G.; Reyes, S.; Moses, G.A.; Fratoni, M.; Flowers, D.; Aceves, S.; Rhodes, M.; Kane, J.; Scott, H.; Kramer, R.; Pantano, C.; Scullard, C.; Sawicki, R.; Wilks, S.; Mehl, M.

    2010-01-01

    The Laser Inertial Fusion Energy (LIFE) concept is being designed to operate as either a pure fusion or hybrid fusion-fission system. A key component of a LIFE engine is the fusion chamber subsystem. The present work details the chamber design for the pure fusion option. The fusion chamber consists of the first wall and blanket. This integrated system must absorb the fusion energy, produce fusion fuel to replace that burned in previous targets, and enable both target and laser beam transport to the ignition point. The chamber system also must mitigate target emissions, including ions, x-rays and neutrons and reset itself to enable operation at 10-15 Hz. Finally, the chamber must offer a high level of availability, which implies both a reasonable lifetime and the ability to rapidly replace damaged components. An integrated LIFE design that meets all of these requirements is described herein.

  8. Fusion Engineering Device. Volume II. Design description

    International Nuclear Information System (INIS)

    1981-10-01

    This volume summarizes the design of the FED. It includes a description of the major systems and subsystems, the supporting plasma design analysis, a projected device cost and associated construction schedule, and a description of the facilities to house and support the device. This effort represents the culmination of the FY81 studies conducted at the Fusion Engineering Design Center (FEDC). Unique in these design activities has been the collaborative involvement of the Design Center personnel and numerous resource physicists from the fusion community who have made significant contributions in the physics design analysis as well as the physics support of the engineering design of the major FED systems and components

  9. Indirect drive targets for fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Amendt, Peter A.; Miles, Robin R.

    2016-10-11

    A hohlraum for an inertial confinement fusion power plant is disclosed. The hohlraum includes a generally cylindrical exterior surface, and an interior rugby ball-shaped surface. Windows over laser entrance holes at each end of the hohlraum enclose inert gas. Infrared reflectors on opposite sides of the central point reflect fusion chamber heat away from the capsule. P2 shields disposed on the infrared reflectors help assure an enhanced and more uniform x-ray bath for the fusion fuel capsule.

  10. Heavy ion fusion targets; issues for fast ignition

    International Nuclear Information System (INIS)

    Bangerter, Roger O.

    2014-01-01

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

  11. Implementing Target Value Design.

    Science.gov (United States)

    Alves, Thais da C L; Lichtig, Will; Rybkowski, Zofia K

    2017-04-01

    An alternative to the traditional way of designing projects is the process of target value design (TVD), which takes different departure points to start the design process. The TVD process starts with the client defining an allowable cost that needs to be met by the design and construction teams. An expected cost in the TVD process is defined through multiple interactions between multiple stakeholders who define wishes and others who define ways of achieving these wishes. Finally, a target cost is defined based on the expected profit the design and construction teams are expecting to make. TVD follows a series of continuous improvement efforts aimed at reaching the desired goals for the project and its associated target value cost. The process takes advantage of rapid cycles of suggestions, analyses, and implementation that starts with the definition of value for the client. In the traditional design process, the goal is to identify user preferences and find solutions that meet the needs of the client's expressed preferences. In the lean design process, the goal is to educate users about their values and advocate for a better facility over the long run; this way owners can help contractors and designers to identify better solutions. This article aims to inform the healthcare community about tools and techniques commonly used during the TVD process and how they can be used to educate and support project participants in developing better solutions to meet their needs now as well as in the future.

  12. Direct-driven target implosion in heavy ion fusion

    International Nuclear Information System (INIS)

    Noguchi, K.; Suzuki, T.; Kurosaki, T.; Barada, D.; Kawata, S.; Ma, Y. Y.; Ogoyski, A. I.

    2016-01-01

    In inertial confinement fusion, the driver beam illumination non-uniformity leads a degradation of fusion energy output. A fuel target alignment error would happen in a fusion reactor; the target alignment error induces heavy ion beam illumination non-uniformity on a target. On the other hand, heavy ion beam accelerator provides a capability to oscillate a beam axis with a high frequency. The wobbling beams may provide a new method to reduce or smooth the beam illumination non-uniformity. First we study the effect of driver irradiation non-uniformity induced by the target alignment error (dz) on the target implosion. We found that dz should be less than about 130 μm for a sufficient fusion energy output. We also optimize the wobbling scheme. The spiral wobbling heavy ion beams would provide a promissing scheme to the uniform beam illumination. (paper)

  13. Introduction to magnetic fusion reactor design

    International Nuclear Information System (INIS)

    Watanabe, Kenji

    1988-01-01

    Trend of the tokamak reactor design works so far carried out is reviewed, and method of conceptual design for commercial fusion reactor is critically considered concerning the black-box conpepts. System-framework of the engineering of magnetic fusion (commercial) reactor design is proposed as four steps. Based on it the next design studies are recommended in parallel approaches for making real-overcome of reactor material problem, from the view point of technological realization and not from the economical one. Real trials are involved. (author)

  14. Target injection and tracking for inertial fusion energy

    Energy Technology Data Exchange (ETDEWEB)

    Petzoldt, R.W. [Lawrence Livermore National Lab., CA (United States); Moir, R.W. [Lawrence Livermore National Lab., CA (United States)

    1996-11-01

    In an inertial fusion power plant, several cryogenic targets must be injected each second into a reaction chamber with speeds of about 100 m s{sup -1}. This speed can be achieved with an acceleration in the range from 1000 to 10 000 m s{sup -2}. The total accuracy of driver beam pointing and target position prediction must be less than {+-}0.6 mm for a 3 mm beam spot radius. A 0.1 {mu}m thick dual membrane supporting the capsule in the hohlraum will allow nearly 2000 m s{sup -2} acceleration. The strength of frozen DT in the capsule is calculated to allow acceleration in excess of 10 000 m s{sup -2} if the DT temperature is less than 17 K. A gas gun is the preferred device for injecting indirect drive targets owing to its simplicity and proven reliability. The amount of gas required for each target (about 10-100 mg) is acceptable. A revolver loading mechanism is recommendced with a cam-operated poppet valve to control the gas flow. Slots near the muzzle of the gun barrel are recommended to vent gas and thereby to improve accuracy and to aid gas pumping. Optical target tracking and electronic timing devices can predict target arrival time with sufficient accuracy. Target steering by electrostatic deflection of the in-flight target is shown to be feasible and would avoid the need to point the beams actively. Calculations show that induced tumble from electrostatically steering the target is not excessive. An experiment has been designed to develop target injection and to verify the predicted accuracy of sequential injection and tracking of multiple targets. (orig.)

  15. Fusion-product energy loss in inertial confinement fusion plasmas with applications to target burns

    International Nuclear Information System (INIS)

    Harris, D.B.; Miley, G.H.

    1984-01-01

    Inertial confinement fusion (ICF) has been proposed as a competitor to magnetic fusion in the drive towards energy production, but ICF target performance still contains many uncertainties. One such area is the energy-loss rate of fusion products. This situation is due in part to the unique plasma parameters encountered in ICF plasmas which are compressed to more than one-thousand times solid density. The work presented here investigates three aspects of this uncertainty

  16. OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 2: Designs, Assessments, and Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C. D.; Laybourne, P.; Shillito, K. R.

    1992-03-01

    This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability, economics, and technology development needs.

  17. Parametric study of a target factory for laser fusion

    International Nuclear Information System (INIS)

    Sherohman, J.W.; Meier, W.R.

    1980-01-01

    An analysis of a target factory leading to the derivation of production rate equations has provided the basis for a parametric study. Rate equations describing the production of laser fusion targets have been developed for the purpose of identifying key parameters, attractive production techniques and cost scaling relationships for a commercial target factory

  18. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-01-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-I through IV. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies. ARIES-1 investigated the use of SiC composite as the structural material to increase the blanket temperature and reduce the blanket activation. Li 2 ZrO 3 was used as the breeding material due to its high temperature stability and good tritium recovery characteristics. The ARIES-IV is a modification of ARIES-1. The plasma was in the second stability regime. Li 2 O was used as the breeding material to remove Zr. A gaseous divertor was used to replace the conventional divertor so that high Z divertor target is not required. The physics of ARIES-II was the same as ARIES-IV. The engineering design of the ARIES-II was based on a self-cooled lithium blanket with a V-alloy as the structural material. Even though it was assumed that the plasma was in the second stability regime, the plasma beta was still rather low (3.4%). The ARIES-III is an advanced fuel (D- 3 He) tokamak reactor. The reactor design assumed major advancement on the physics, with a plasma beta of 23.9%. A conventional structural material is acceptable due to the low neutron wall loading. From the radiation damage point of view, the first wall can last the life of the reactor, which is expected to be a major advantage from the engineering design and waste disposal point of view

  19. Current fusion power plant design concepts

    International Nuclear Information System (INIS)

    Gore, B.F.; Murphy, E.S.

    1976-09-01

    Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes

  20. Conceptual design of a laser fusion power plant

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Meier, W.R.; Monsler, M.J.

    1977-01-01

    A conceptual design of a laser fusion power plant is extensively discussed. Recent advances in high gain targets are exploited in the design. A smaller blanket structure is made possible by use of a thick falling region of liquid lithium for a first wall. Major design features of the plant, reactor, and laser systems are described. A parametric analysis of performance and cost vs. design parameters is presented to show feasible design points. A more definitive follow-on conceptual design study is planned

  1. Conceptual design of laser fusion reactor KOYO-fast

    International Nuclear Information System (INIS)

    Tomabechi, K.; Kozaki, Y.; Norimatsu, T.

    2006-01-01

    A conceptual design of the laser fusion reactor KOYO-F based on the fast ignition scheme is reported including the target design, the laser system and the design for chamber. A Yb-YAG ceramic laser operated at 200 K is the primary candidate for the compression laser and an OPCPA (optical parametric chirped pulse amplification) system is the one for the ignition laser. The chamber is basically a wet wall type but the fire position is vertically off-set to simplify the protection scheme of the ceiling. The target consists of foam insulated, cryogenic DT shells with a LiPb, reentrant guide-cone. (authors)

  2. Designing the Cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1987-01-01

    The primary goal in designing inertial confinement fusion (ICF) reactors is to produce electrical power as inexpensively as possible, with minimum activation and without compromising safety. This paper discusses a method for designing the Cascade rotating ceramic-granule-blanket reactor (Pitts, 1985) and its associated power plant (Pitts and Maya, 1985). Although focus is on the cascade reactor, the design method and issues presented are applicable to most other ICF reactors

  3. Materials design data for fusion reactors

    International Nuclear Information System (INIS)

    Tavassoli, A.A.F.

    1998-01-01

    Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.)

  4. Materials design data for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.A.F. [CEA Commissariat a l`Energie Atomique, Gif sur Yvette (France). CEREM

    1998-10-01

    Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.) 19 refs.

  5. Pellet design for a laser fusion reactor

    International Nuclear Information System (INIS)

    Thiessen, A.R.; Nuckolls, J.

    1974-01-01

    The requirements for laser fusion pellet design are discussed. Computer calculations are presented of a capsule consisting of a spherical solid drop of DT surrounded by a concentric shell of DT. Gains greater than 40 fold are achieved with laser energies of approximately 0.5 MJ, and peak powers of about 10 16 W. (U.S.)

  6. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Gerstl, S.A.; Houck, D.L.; Jalbert, R.A.; Krakowski, R.A.; Linford, R.K.; McDonald, T.E.; Rogers, J.D.; Thomassen, K.I.

    1975-01-01

    A general design of the system is given. The implosion heating and compression systems (METS) are described. Tritium handling, shielding and activation of the reactor, and safety and environmental aspects are discussed

  7. Compression of magnetized target in the magneto-inertial fusion

    Science.gov (United States)

    Kuzenov, V. V.

    2017-12-01

    This paper presents a mathematical model, numerical method and results of the computer analysis of the compression process and the energy transfer in the target plasma, used in magneto-inertial fusion. The computer simulation of the compression process of magnetized cylindrical target by high-power laser pulse is presented.

  8. Neutron penumbral imaging of laser-fusion targets

    International Nuclear Information System (INIS)

    Lerche, R.A.; Ress, D.B.

    1988-01-01

    Using a new technique, penumbral coded-aperture imaging, the first neutron images of laser-driven, inertial-confinement fusion targets were obtained. With these images the deuterium-tritium burn region within a compressed target can be measured directly. 4 references, 11 figures

  9. Magnetic compression/magnetized target fusion (MAGO/MTF)

    International Nuclear Information System (INIS)

    Kirkpatrick, R.C.; Lindemuth, I.R.

    1997-03-01

    Magnetized Target Fusion (MTF) was reported in two papers at the First Symposium on Current Trends in International Fusion Research. MTF is intermediate between two very different mainline approaches to fusion: Inertial Confinement Fusion (ICF) and magnetic confinement fusion (MCF). The only US MTF experiments in which a target plasma was compressed were the Sandia National Laboratory ''Phi targets''. Despite the very interesting results from that series of experiments, the research was not pursued, and other embodiments of MTF concept such as the Fast Liner were unable to attract the financial support needed for a firm proof of principle. A mapping of the parameter space for MTF showed the significant features of this approach. The All-Russian Scientific Research Institute of Experimental Physics (VNIIEF) has an on-going interest in this approach to thermonuclear fusion, and Los Alamos National Laboratory (LANL) and VNIIEF have done joint target plasma generation experiments relevant to MTF referred to as MAGO (transliteration of the Russian acronym for magnetic compression). The MAGO II experiment appears to have achieved on the order of 200 eV and over 100 KG, so that adiabatic compression with a relatively small convergence could bring the plasma to fusion temperatures. In addition, there are other experiments being pursued for target plasma generation and proof of principle. This paper summarizes the previous reports on MTF and MAGO and presents the progress that has been made over the past three years in creating a target plasma that is suitable for compression to provide a scientific proof of principle experiment for MAGO/MTF

  10. Status of fusion reactor blanket design

    International Nuclear Information System (INIS)

    Smith, D.L.; Sze, D.K.

    1986-02-01

    The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

  11. Conceptual design study of IFMIF target system

    International Nuclear Information System (INIS)

    Kato, Y.; Nakamura, H.; Ida, M.; Maekawa, H.; Katsuta, H.; Hua, T.; Cevolani, S.

    1997-01-01

    IFMIF-CDA (International Fusion Materials Irradiation Facility - Conceptual Design Activity) had been carried out during 1995 and 1996, under the auspices of the IEA. The mission of this facility is to provide an accelerator based deuterium-lithium (D-Li) neutron source to test the candidate materials of radiation - resistant and low - activation materials up to about a full lifetime of anticipated use in fusion energy reactors. The neutrons of about 14 MeV are obtained by the stripping reaction of the deuteron of Max. 40 MeV with target lithium. Total deuteron beam current is about 250 mA and beam footprint is 20 cm x 5 cm on the free surface of lithium jet. In this report general characteristics of the target lithium system and the results of thermal and flow analysis for the target lithium jet are described. (author)

  12. Physics and technology of inertial fusion energy targets chambers and drivers. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2005-09-01

    The third IAEA Technical Meeting on Physics and Technology of Inertial Fusion Energy Targets and Chambers took place 11-13 October 2004 in the Yousung Hotel Daejon, Republic of Korea. The first meeting was held in Madrid, Spain, 7-9 June 2000, and the second one in San Diego, California, 17-19 June 2002. Nuclear fusion has the promise of becoming an abundant energy source with good environmental compatibility. Excellent progress has been made in controlled nuclear fusion research on both magnetic and inertial approaches for plasma confinement. The IAEA plays a pro-active role to catalyze innovation and enhance worldwide commitment to fusion. This is done by creating awareness of the different concepts of magnetic as well as inertial confinement. The International Fusion Research Council (IFRC) supports the IAEA in the development of strategies to enhance fusion research in Member States. As part of the recommendations, a technical meeting on the physics and technology of inertial fusion energy (IFE) was proposed in one of the council meetings. The objective of the technical meeting was to contribute to advancing the understanding of targets and chambers for all proposed inertial fusion energy power plant designs. The topics to be covered were: Target design and physics, chamber design and physics, target fabrication injection and Tritium handling, assessment of safety, environment and economy aspect of IFE. It was recognized by the International Advisory Committee that the scope of the meeting should also include fusion drivers. The presentations of the meeting included target and chamber physics and technology for all proposed IFE plant concepts (laser driven, heavy-ion driven, Z-pinches, etc.). The final Research Coordination Meeting of the Coordinated Research Project on Elements of Power Plant Design for Inertial Fusion Energy, including further new results and achievements, followed the technical meeting. Twenty-nine participants from 12 countries participated

  13. LIBRA - a light ion beam fusion conceptual reactor design

    International Nuclear Information System (INIS)

    Badger, B.; Moses, G.A.; Engelstad, R.L.; Kulcinski, G.L.; Lovell, E.; MacFarlane, J.; Peterson, R.R.; Sawan, M.E.; Sviatovslavsky, I.N.; Wittenberg, L.J.; Cook, D.L.; Olson, R.E.; Stinnett, R.W.; Ehrhardt, J.; Kessler, G.; Stein, E.

    1990-08-01

    The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4 MJ shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbind tubes with Li 17 Pb 83 flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with Li 17 Pb 83 that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequently without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of a 331 MWe LIBRA design is estimated to be 2843 Dollar/kWe while a 1200 MWe LIBRA design will cost approximately 1300 Dollar/kWe. (orig.) [de

  14. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Linford, R.K.; Oliphant, T.A.; Thomassen, K.I.

    1976-01-01

    The SFTR is a proposed 80-m diameter D-T burning toroidal theta pinch. The system is designed to achieve Q = 1 where Q is the ratio of the total thermonuclear energy output to the maximum stored energy in the plasma. SFTR design studies [1] will provide valuable guidance to the Scyllac related research and to the needed technological development. The portion of the system directly related to the plasma confinement, stability, and heating, is described, and the approach used to obtain an operating point consistent with Q = 1, m = 1 stability, and technological limitations is outlined. (U.K.)

  15. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Linford, R.K.; Oliphant, T.A.; Thomassen, K.I.

    1975-01-01

    The SFTR is a proposed 80-m diameter D--T burning toroidal theta pinch. The system is designed to achieve Q = 1 where Q is the ratio of the total thermonuclear energy output to the maximum stored energy in the plasma. SFTR design studies will provide valuable guidance to the Scyllac related research and to the needed technological development. This paper describes the portion of the system directly related to the plasma confinement, stability, and heating, and outlines the approach used to obtain an operating point consistent with Q = 1, m = 1 stability, and technological limitations. (auth)

  16. Target injection methods for inertial fusion energy

    International Nuclear Information System (INIS)

    Petzoldt, R.W.; Moir, R.W.

    1994-06-01

    We have studied four methods to inject IFE targets: the gas gun, electrostatic accelerator, induction accelerator, and rail gun. We recommend a gas gun for indirect drive targets because they can support a gas pressure load on one end and can slide along the gun barrel without damage. With the gas gun, the amount of gas required for each target (about 10 to 100 mg) is acceptable; for other types of targets, a sabot would be necessary. A cam and poppet valve arrangement is recommended for gas flow control. An electrostatic accelerator is attractive for use with lightweight spherical direct drive targets. Since there is no physical contact between the target and the injector, there will be no wear of either component during the injection process. An induction accelerator has an advantage of no electrical contact between the target and the injector. Physical contact is not even necessary, so the wear should be minimal. It requires a cylindrical conductive target sleeve which is a substantial added mass. A rail gun is a simpler device than an electrostatic accelerator or induction accelerator. It requires electrical contact between the target and the rails and may have a significant wear rate. The wear in a vacuum could be reduced by use of a solid lubricant such as MoS 2 . The total required accuracy of target injection, tracking and beam pointing of ±0.4 mm appears achievable but will require development and experimental verification

  17. Conceptual design of Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Tone, T.; Fujisawa, N.

    1983-01-01

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been performed. The FER has an objective of achieving selfignition and demonstrating engineering feasibility as a next generation tokamak to JT-60. Various concepts of the FER have been considered. The reference design is based on a double-null divertor. Optional design studies with some attractive features based on advanced concepts such as pumped limiter and RF current drive have been carried out. Key design parameters are; fusion power of 440 MW, average neutron wall loading of 1MW/m 2 , major radius of 5.5m, plasma minor radius of 1.1m, plasma elongation of 1.5, plasma current of 5.3MA, toroidal beta of 4%, toroidal field on plasma axis of 5.7T and tritium breeding ratio of above unity

  18. Extrap conceptual fusion reactor design study

    International Nuclear Information System (INIS)

    Eninger, J.E; Lehnert, B.

    1987-12-01

    A study has recently been initiated to asses the fusion reactor potential of the Extrap concept. A reactor model is defined that fulfills certain economic and environmental criteria. This model is applied to Extrap and a reference reactor is outlined. The design is optimized by varying parameters subject to both physics and engineering constraints. Several design options are examined and key engineering issues are identified and addressed. Some preliminary results and conclusions of this work are summarized. (authors)

  19. Laser solenoid fusion--fission design

    International Nuclear Information System (INIS)

    Steinhauer, L.C.; Taussig, R.T.

    1976-01-01

    The dependence of breeding performance on system engineering parameters is examined for laser solenoid fusion-fission reactors. Reactor performance is found to be relatively insensitive to most of the engineering parameters, and compact designs can be built based on reasonable technologies. Point designs are described for the prototype series of reactors (mid-term technologies) and for second generation systems (advanced technologies). It is concluded that the laser solenoid has a good probability of timely application to fuel breeding needs

  20. SAR Data Fusion Imaging Method Oriented to Target Feature Extraction

    Directory of Open Access Journals (Sweden)

    Yang Wei

    2015-02-01

    Full Text Available To deal with the difficulty for target outlines extracting precisely due to neglect of target scattering characteristic variation during the processing of high-resolution space-borne SAR data, a novel fusion imaging method is proposed oriented to target feature extraction. Firstly, several important aspects that affect target feature extraction and SAR image quality are analyzed, including curved orbit, stop-and-go approximation, atmospheric delay, and high-order residual phase error. Furthermore, the corresponding compensation methods are addressed as well. Based on the analysis, the mathematical model of SAR echo combined with target space-time spectrum is established for explaining the space-time-frequency change rule of target scattering characteristic. Moreover, a fusion imaging strategy and method under high-resolution and ultra-large observation angle range conditions are put forward to improve SAR quality by fusion processing in range-doppler and image domain. Finally, simulations based on typical military targets are used to verify the effectiveness of the fusion imaging method.

  1. 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.)

  2. Economic comparison of fusion power plant designs

    International Nuclear Information System (INIS)

    O'Neill, J.E.

    1986-01-01

    Over the past 10 yr, a number of studies have been developed for fusion power plants of various types (tokamaks, mirrors, etc.) complete with figures of merit such as cost estimates and estimates of the cost of generating electricity (COE). Each of these designs involves unresolved physics and engineering problems which, it is assumed, will eventually be worked out. Because of such uncertainties the figures of merit associated with such designs are not to be compared as absolute measures of worth but as relative indicators of progress within a given concept type. As part of Grumman's involvement in fusion energy development, an effort has been undertaken to compare economic indicators from the referenced studies in order to determine the cost trend in recent reactor design activities

  3. An optimized hydrogen target for muon catalyzed fusion

    Energy Technology Data Exchange (ETDEWEB)

    Gheisari, R., E-mail: gheisari@pgu.ac.i [Physics Department, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of)

    2011-04-01

    This paper deals with the optimization of the processes involved in muon catalyzed fusion. Muon catalyzed fusion ({mu}CF) is studied in all layers of the solid hydrogen structure H/0.1%T+D{sub 2}+HD. The layer H/T acts as an emitter source of energetic t{mu} atoms, due to the so-called Ramsauer-Townsend effect. These t{mu} atoms are slowed down in the second layer (degrader) and are forced to take place nuclear fusion in HD. The degrader affects time evolution of t{mu} atomic beam. This effect has not been considered until now in {mu}CF-multilayered targets. Due to muon cycling and this effect, considerable reactions occur in the degrader. In our calculations, it is shown that the fusion yield equals 180{+-}1.5. It is possible to separate events that overlap in time.

  4. Enhancing potency of siRNA targeting fusion genes by optimization outside of target sequence.

    Science.gov (United States)

    Gavrilov, Kseniya; Seo, Young-Eun; Tietjen, Gregory T; Cui, Jiajia; Cheng, Christopher J; Saltzman, W Mark

    2015-12-01

    Canonical siRNA design algorithms have become remarkably effective at predicting favorable binding regions within a target mRNA, but in some cases (e.g., a fusion junction site) region choice is restricted. In these instances, alternative approaches are necessary to obtain a highly potent silencing molecule. Here we focus on strategies for rational optimization of two siRNAs that target the junction sites of fusion oncogenes BCR-ABL and TMPRSS2-ERG. We demonstrate that modifying the termini of these siRNAs with a terminal G-U wobble pair or a carefully selected pair of terminal asymmetry-enhancing mismatches can result in an increase in potency at low doses. Importantly, we observed that improvements in silencing at the mRNA level do not necessarily translate to reductions in protein level and/or cell death. Decline in protein level is also heavily influenced by targeted protein half-life, and delivery vehicle toxicity can confound measures of cell death due to silencing. Therefore, for BCR-ABL, which has a long protein half-life that is difficult to overcome using siRNA, we also developed a nontoxic transfection vector: poly(lactic-coglycolic acid) nanoparticles that release siRNA over many days. We show that this system can achieve effective killing of leukemic cells. These findings provide insights into the implications of siRNA sequence for potency and suggest strategies for the design of more effective therapeutic siRNA molecules. Furthermore, this work points to the importance of integrating studies of siRNA design and delivery, while heeding and addressing potential limitations such as restricted targetable mRNA regions, long protein half-lives, and nonspecific toxicities.

  5. Optical design considerations for laser fusion reactors

    International Nuclear Information System (INIS)

    Monsler, M.J.; Maniscalco, J.A.

    1977-09-01

    The plan for the development of commercial inertial confinement fusion (ICF) power plants is discussed, emphasizing the utilization of the unique features of laser fusion to arrive at conceptual designs for reactors and optical systems which minimize the need for advanced materials and techniques requiring expensive test facilities. A conceptual design for a liquid lithium fall reactor is described which successfully deals with the hostile x-ray and neutron environment and promises to last the 30 year plant lifetime. Schemes for protecting the final focusing optics are described which are both compatible with this reactor system and show promise of surviving a full year in order to minimize costly downtime. Damage mechanisms and protection techniques are discussed, and a recommendation is made for a high f-number metal mirror final focusing system

  6. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Palmer, B.J.F.

    1987-11-01

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  7. Recent designs for advanced fusion reactor blankets

    International Nuclear Information System (INIS)

    Sze, D.K.

    1994-06-01

    A series of reactor design studies based on the Tokamak configuration have been carried out under the direction of Professor Robert Conn of UCLA. They are called ARIES-1 through 4 and PULSAR 1 and 2. The key mission of these studies is to evaluate the attractiveness of fusion assuming different degrees of advancement in either physics or engineering development. Also, the requirements of engineering and physics systems for a pulsed reactor were evaluated by the PULSAR design studies. This paper discusses the directions and conclusions of the blanket and related engineering systems for those design studies

  8. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    Moir, R.W.

    1978-01-01

    This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

  9. Targeting the Conserved Fusion Loop of HAP2 Inhibits the Transmission of Plasmodium berghei and falciparum

    Directory of Open Access Journals (Sweden)

    Fiona Angrisano

    2017-12-01

    Full Text Available Summary: Inhibiting transmission of Plasmodium is a central strategy in malarial eradication, and the biological process of gamete fusion during fertilization is a proven target for this approach. The lack of a structure or known molecular function of current anti-malarial vaccine targets has previously been a hindrance in the development of transmission-blocking vaccines. Structure/function studies have indicated that the conserved gamete membrane fusion protein HAP2 is a class II viral fusion protein. Here, we demonstrate that targeting a function-critical site of the fusion/cd loop with species-specific antibodies reduces Plasmodium berghei transmission in vivo by 58.9% and in vitro fertilization by up to 89.9%. A corresponding reduction in P. falciparum transmission (75.5%/36.4% reductions in intensity/prevalence is observed in complimentary field studies. These results emphasize conserved mechanisms of fusion in Apicomplexa, while highlighting an approach to design future anti-malarial transmission-blocking vaccines. : Angrisano et al. find that the HAP2 cd-loop can be targeted as an anti-malarial intervention, is immunogenic across multiple plasmodial species, can induce antibodies that specifically recognize the sexual stages of the parasitic life cycle, and can mediate transmission-blocking immunity in the lab and the field. Keywords: HAP2, malaria, transmission, fusion, vaccine

  10. Designs of tandem-mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Barr, W.L.; Boghosian, B.M.

    1981-01-01

    We have completed a comparative evaluation of several end plug configurations for tandem mirror fusion reactors with thermal barriers. The axi-cell configuration has been selected for further study and will be the basis for a detailed conceptual design study to be carried out over the next two years. The axi-cell end plug has a simple mirror cell produced by two circular coils followed by a transition coil and a yin-yang pair, which provides for MHD stability

  11. Progress In Magnetized Target Fusion Driven by Plasma Liners

    Science.gov (United States)

    Thio, Francis Y. C.; Kirkpatrick, Ronald C.; Knapp, Charles E.; Cassibry, Jason; Eskridge, Richard; Lee, Michael; Smith, James; Martin, Adam; Wu, S. T.; Schmidt, George; hide

    2001-01-01

    Magnetized target fusion (MTF) attempts to combine the favorable attributes of magnetic confinement fusion (MCF) for energy confinement with the attributes of inertial confinement fusion (ICF) for efficient compression heating and wall-free containment of the fusing plasma. It uses a material liner to compress and contain a magnetized plasma. For practical applications, standoff drivers to deliver the imploding momentum flux to the target plasma remotely are required. Spherically converging plasma jets have been proposed as standoff drivers for this purpose. The concept involves the dynamic formation of a spherical plasma liner by the merging of plasma jets, and the use of the liner so formed to compress a spheromak or a field reversed configuration (FRC).

  12. Design of a fusion engineering test facility

    International Nuclear Information System (INIS)

    Sager, P.H.

    1980-01-01

    The fusion Engineering Test Facility (ETF) is being designed to provide for engineering testing capability in a program leading to the demonstration of fusion as a viable energy option. It will combine power-reactor-type components and subsystems into an integrated tokamak system and provide a test bed to test blanket modules in a fusion environment. Because of the uncertainties in impurity control two basic designs are being developed: a design with a bundle divertor (Design 1) and one with a poloidal divertor (Design 2). The two designs are similar where possible, the latter having somewhat larger toroidal field (TF) coils to accommodate removal of the larger torus sectors required for the single-null poloidal divertor. Both designs have a major radius of 5.4 m, a minor radius of 1.3 m, and a D-shaped plasma with an elongation of 1.6. Ten TF coils are incorporated in both designs, producing a toroidal field of 5.5 T on-axis. The ohmic heating and equilibrium field (EF) coils supply sufficient volt-seconds to produce a flat-top burn of 100 s and a duty cycle of 135 s, including a start of 12 s, a burn termination of 10 s, and a pumpdown of 13 s. The total fusion power during burn is 750 MW, giving a neutron wall loading of 1.5 MW/m 2 . In Design 1 of the poloidal field (PF) coils except the fast-response EF coils are located outside the FT coils and are superconducting. The fast-response coils are located inside the TF coil bore near the torus and are normal conducting so that they can be easily replaced.In Design 2 all of the PF coils are located outside the TF coils and are superconducting. Ignition is achieved with 60 MW of neutral beam injection at 150 keV. Five megawatts of radio frequency heating (electron cyclotron resonance heating) is used to assist in the startup and limit the breakdown requirement to 25 V

  13. Safety and environmental advantages of using tritium-lean targets for inertial fusion

    International Nuclear Information System (INIS)

    Arzeni, S.; Latkowski, J. F.; Logan, B. G.; Meier, W. R.; Moir, R. W.; Perkins, L. J.; Sanz, J.

    1999-01-01

    While traditional inertial fusion energy target designs typically use equimolar portions of deuterium and tritium and have areal densities (ρr) of ∼ 3 g/cm 2 , significant safety and environmental (S and E) advantages may be obtained through the use of high-density (ρr ∼ 10 g/cm 2 ) targets with tritium components as low as 0.5%. Such targets would absorb much of the neutron energy within the target and could be self-sufficient from a tritium breeding point of view. Tritium self-sufficiency within the target would free target chamber designers from the need to use lithium-bearing blanket materials, while low inventories within each target would translate into low inventories in target fabrication facilities. Absorption of much of the neutron energy within the target, the extremely low tritium inventories, and the greatly moderated neutron spectrum, make ''tritium-lean'' targets appear quite attractive from an S and E perspective

  14. Recent developments in laser-fusion target coatings

    International Nuclear Information System (INIS)

    Fries, R.J.; Catlett, D.S.; Fossey, D.; Mayer, A.; McCreary, W.J.; Powell, B.W.; Simonsic, G.A.

    1976-01-01

    Techniques to fabricate hollow, spherical, multilayered laser-fusion targets are described. The first is a glow discharge polymerization process for plastic coating. A chemical vapor deposition process for depositing Mo/Re alloys is also discussed along with some new techniques for electrodeless plating and for electroplating a wide variety of metals

  15. Electroless or autocatalytic coating of microparticles for laser fusion targets

    International Nuclear Information System (INIS)

    Mayer, A.; Catlett, D.S.

    1977-04-01

    Use of a novel device for applying uniform metallic coatings to spherical microparticles is described. The apparatus deposits electroless metal coatings on hollow, thin-walled metal or sensitized nonmetallic micromandrels. The apparatus and process were developed for fabrication of microsphere pressure vessels for use as targets in laser-initiated fusion research

  16. Electrolytic coating of microparticles for laser fusion targets

    International Nuclear Information System (INIS)

    Mayer, A.; Catlett, D.S.

    1977-04-01

    An electroplating apparatus for applying uniform metallic coatings that have excellent surface finishes to discrete microparticles is described. The device is used to electrodeposit metals onto thin-walled metal, metallized glass, or plastic mandrels. The apparatus and process were developed for fabrication of microsphere pressure vessels to be used as targets in laser fusion research

  17. Fusion blanket design and optimization techniques

    International Nuclear Information System (INIS)

    Gohar, Y.

    2005-01-01

    In fusion reactors, the blanket design and its characteristics have a major impact on the reactor performance, size, and economics. The selection and arrangement of the blanket materials, dimensions of the different blanket zones, and different requirements of the selected materials for a satisfactory performance are the main parameters, which define the blanket performance. These parameters translate to a large number of variables and design constraints, which need to be simultaneously considered in the blanket design process. This represents a major design challenge because of the lack of a comprehensive design tool capable of considering all these variables to define the optimum blanket design and satisfying all the design constraints for the adopted figure of merit and the blanket design criteria. The blanket design techniques of the First Wall/Blanket/Shield Design and Optimization System (BSDOS) have been developed to overcome this difficulty and to provide the state-of-the-art techniques and tools for performing blanket design and analysis. This report describes some of the BSDOS techniques and demonstrates its use. In addition, the use of the optimization technique of the BSDOS can result in a significant blanket performance enhancement and cost saving for the reactor design under consideration. In this report, examples are presented, which utilize an earlier version of the ITER solid breeder blanket design and a high power density self-cooled lithium blanket design for demonstrating some of the BSDOS blanket design techniques

  18. Electrostatic levitation and transport of laser fusion targets

    International Nuclear Information System (INIS)

    Johnson, W.L.; Hendricks, C.D.

    1980-01-01

    Several levitation concepts have been evaluated resulting in the electrostatic quadrupole being chosen as the most universal. A levitator has been constructed to handle laser fusion targets during and between the processing steps. The levitator is based on a quadrupole rail which is segmented to provide electrically controlled transport and confinement along the rail. This device has demonstrated transport both vertical and horizontal of targets with appropriate mass to size ratios and exhibits remarkably stable confinement at atmospheric pressure

  19. Ion tail filling in laser-fusion targets

    International Nuclear Information System (INIS)

    Henderson, D.B.

    1975-06-01

    Thermonuclear burn begins in laser-fusion targets with the collapse of the imploding fuel shell. At this instant the ion velocity distribution is non-Maxwellian, requiring correction to the commonly used computer simulation codes. This correction is computed and compared with that arising from the loss of fast ions in marginal (rho R less than 0.01 gm cm -2 ) targets. (U.S.)

  20. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

  1. Interaction physics for megajoule laser fusion targets

    International Nuclear Information System (INIS)

    Kruer, W.L.

    1992-02-01

    Some little-explored interaction phenomena for targets irradiated with megajoule lasers are considered. Simple estimates show that the laser plasma interaction then occurs in a hot (multi-keV) plasma with density much less than the critical density. In such plasmas, Raman and Brillouin scattering into the forward hemisphere are potentially significant. A simple model shows that Raman forward scattering can be saturated at low levels by ponderomotive detuning. Calculations also illustrate a suppression of ponderomotive filamentation by plasma-induced beam smoothing

  2. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-02-01

    This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

  3. Design of a target and moderator at the Los Alamos Spallation Radiation Effects Facility (LASREF) as a neutron source for fusion reactor materials development

    International Nuclear Information System (INIS)

    Ferguson, P.D.; Mueller, G.E.

    1993-01-01

    The LASREF facility is located in the beam stop area at LAMPF. The neutron spectrum is fission-like with the addition of a 3% to 5% component with E > 20 MeV. The present study evaluates the limits on geometry and material selection that will maximize the neutron flux. MCNP and LAHET were used to predict the neutron flux and energy spectrum for a variety of geometries. The problem considers 760 MeV protons incident on tungsten. The resulting neutrons are multiplied in uranium through (n,xn) reactions. Calculations show that a neutron flux greater than 10 19 n/m 2 /s is achievable. The helium to dpa ratio and the transmutation product generation are calculated. These results are compared to expectations for the proposed DEMO fusion reactor and to FFTF

  4. Osiris and SOMBRERO inertial confinement fusion power plant designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Bieri, R.L.; Monsler, M.J.

    1992-03-01

    The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs

  5. Thermal Studies of the Laser Inertial Fusion Energy (LIFE) Target during Injection into the Fusion Chamber

    Energy Technology Data Exchange (ETDEWEB)

    Miles, R. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Havstad, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); LeBlanc, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chang, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Golosker, I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rosso, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-09-09

    The tests of the external heat transfer coefficient suggests that the values used in the numerical analysis for the temperature distribution within the fusion fuel target following flight into the target chamber are probably valid. The tests of the heat transfer phenomena occurring within the target due the rapid heating of the LEH window for the hot gasses within the fusion chamber show that the heat does indeed convect via the internal helium environment of the target towards the capsule and that the pressure in the front compartment of the target adjacent to the LEH window increases such that t bypass venting of the internal helium into the second chamber adjacent to the capsule is needed to prevent rupture of the membranes. The bypass flow is cooled by the hohlraum during this venting. However, the experiments suggest that our internal heat flow calculations may be low by about a factor of 2. Further studies need to be conducted to investigate the differences between the experiment and the numerical analysis. Future studies could also possibly bring the test conditions closer to those expected in the fusion chamber to better validate the results. A sacrificial layer will probably be required on the LEH window of the target and this can be used to mitigate any unexpected target heating.

  6. Design windows of laser fusion power plants and conceptual design of laser-diode pumped slab laser

    International Nuclear Information System (INIS)

    Kozaki, Y.; Eguchi, T.; Izawa, Y.

    1999-01-01

    An analysis of the design space available to laser fusion power plants has been carried out, in terms of design key parameters such as target gain, laser energy and laser repetition rate, the number of fusion react ion chambers, and plant size. The design windows of economically attractive laser fusion plants is identified with the constraints of key design parameters and the cost conditions. Especially, for achieving high repetition rate lasers, we have proposed and designed a diode-pumped solid-state laser driver which consists of water-cooled zig-zag path slab amplifiers. (author)

  7. Non-LTE effects in inertial confinement fusion target chambers

    International Nuclear Information System (INIS)

    MacFarlane, J.J.; Moses, G.A.; Peterson, R.R.

    1989-01-01

    In previous studies of transport processes in inertial confinement fusion target chambers, the radiative properties of the background plasma were calculated under the assumption of local thermodynamic equilibrium (LTE). In this paper, the authors present a study of the equation of state and the radiative properties of high temperature, low-to-moderate density ( 21 cm -3 ) plasmas for the determination of the conditions under which non-LTE effects become important and for an assessment of the importance of non-LTE processes in target chambers during high yield inertial fusion target explosions. For this purpose, two-body (radiative and dielectronic) and three-body (collisional) recombination and de-excitation processes are considered in calculating the steady state ionization and excitation populations. The results of this study indicate that non-LTE processes generally become important at temperatures of > or approx. 1, 10 and 100 eV for plasma densities of 10 18 , 10 19 and 10 21 cm -3 , respectively. Radiation hydrodynamic simulations utilizing the equation of state and the opacities for a non-LTE argon plasma were performed to study the response of a background gas to an inertial fusion target explosion. These calculations indicate that non-LTE processes are often the dominant atomic processes in the background plasma and that they can strongly affect the radiative and shock properties as energy is transported away from the point of the target explosion. (author). 22 refs, 10 figs, 1 tab

  8. Laser-fusion targets for reactors

    International Nuclear Information System (INIS)

    Nuckolls, J.H.; Thiessen, A.R.

    1987-01-01

    This patent describes a target having a centrally located substantially spherically configured quantity of solid fuel for implosion by a pulse of laser energy and having no material therein with a Z of over about 13. The improvement consists of: means in spaced apart and non-contiguous relationship surrounding the fuel for at least providing an atmosphere about the fuel for ensuring electron transport around the fuel and enhancing subsequent implosion symmetry of the fuel, the fuel being configured as a hollow shell; the means consisting of at least one outer layer of substantially solid atmosphere forming material having a Z of 1-13. The atmosphere forms material comprising a shell positioned about the fuel defining a space therebetween, the space being filled with He, the fuel and the shell of atmosphere forming material being each composed of DT, the layer of atmosphere forming material being impacted and at least partially exploded by at least one separate and distinct laser prepulse to produce the atmosphere about the fuel prior to implosion of the fuel by the pulse of laser energy

  9. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    Energy Technology Data Exchange (ETDEWEB)

    Aleksandrova, I. V.; Koresheva, E. R., E-mail: elena.koresheva@gmail.com; Krokhin, O. N. [Russian Academy of Sciences, Lebedev Physical Institute (Russian Federation); Osipov, I. E. [Power Efficiency Centre, Inter RAO UES (Russian Federation)

    2016-12-15

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  10. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1985-01-01

    The Fusion Experimental Reactor (FER) being developed at JAERI as a next generation tokamak to JT-60 has a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. During FY82 and FY83 a comprehensive and intensive conceptual design study has been conducted for a pulsed operation FER as a reference option which employs a conventional inductive current drive and a double-null divertor. In parallel with the reference design, studies have been carried out to evaluate advanced reactor concepts such as quasi-steady state operation and steady state operation based on RF current drive and pumped limiter, and comparative studies for single-null divertor/pumped limiter. This report presents major results obtained primarily from FY83 design studies, while the results of FY82 design studies are described in previous references (JAERI-M 83-213--216). (author)

  11. Preliminary analysis of a target factory for laser fusion

    International Nuclear Information System (INIS)

    Sherohman, J.W.; Hendricks, C.D.

    1980-01-01

    An analysis of a target factory leading to the determination of production expressions has provided for the basis of a parametric study. Parameters involving the input and output rate of a process system, processing yield factors, and multiple processing steps and production lines have been used to develop an understanding of their dependence on the rate of target injection for laser fusion. Preliminary results have indicated that a parametric study of this type will be important in the selection of processing methods to be used in the final production scheme of a target factory

  12. Technological aspects of cryogenic laser-fusion targets

    International Nuclear Information System (INIS)

    Musinski, D.L.; Henderson, T.M.; Simms, R.J.; Pattinson, T.R.; Jacobs, R.B.

    1980-01-01

    Most current laser-fusion targets consist of hollow spherical glass shells which have been filled with a mixture of gaseous deuterium-tritium fuel. Theoretical considerations suggest that optimum yields can be obtained from these targets if the fuel is condensed as a uniform liquid or solid layer on the inner surface of the glass shell at the time it is irradiated. In principle, this can be accomplished in a straightforward way by cooling the target below the condensation or freezing point of the fuel. In practice, cryogenic targets can appear in routine laser experiments only when the necessary cryogenic technology is reliably integrated into experimental target chambers. Significant progress has been made recently in this field. The authors will discuss the scientific basis and the various technological features of a system which has allowed the successful irradiation of uniform solid-fuel-layer targets

  13. Low-density hydrocarbon foams for laser fusion targets: Progress report, 1986

    International Nuclear Information System (INIS)

    Chen, C.; Cook, R.C.; Haendler, B.L.; Hair, L.M.; Kong, F.M.; Letts, S.A.

    1987-06-01

    We describe progress made during 1986 in the development of direct-drive hydrocarbon foam targets for laser fusion. The foam materials are polystyrene and resorcinolformaldehyde. The processes for making the foams, their properties, characterization techniques, and the relationships of their properties to target specifications are presented. In the final section, we discuss statistical experimental design techniques that are being used to optimize the foams. 12 refs., 14 figs., 2 tabs

  14. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-01-01

    Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

  15. Direct drive target survival during injection in an inertial fusion energy power plant

    International Nuclear Information System (INIS)

    Petzoldt, R.W.; Goodin, D.T.; Nikroo, A.; Stephens, E.; Alexander, N.B.; Gallix, R.; Siegel, N.; Raffray, A.R.; Mau, T.K.; Tillack, M.; Najmabadi, F.; Krasheninnikov, S.I.

    2002-01-01

    In inertial fusion energy (IFE) power plant designs, the fuel is a spherical layer of frozen DT contained in a target that is injected at high velocity into the reaction chamber. For direct drive, typically laser beams converge at the centre of the chamber (CC) to compress and heat the target to fusion conditions. To obtain the maximum energy yield from the fusion reaction, the frozen DT layer must be at about 18.5 K and the target must maintain a high degree of spherical symmetry and surface smoothness when it reaches the CC. During its transit in the chamber the cryogenic target is heated by radiation from the hot chamber wall. The target is also heated by convection as it passes through the rarefied fill-gas used to control chamber wall damage by x-rays and debris from the target explosion. This article addresses the temperature limits at the target surface beyond which target uniformity may be damaged. It concentrates on direct drive targets because fuel warm up during injection is not currently thought to be an issue for present indirect drive designs and chamber concepts. Detailed results of parametric radiative and convective heating calculations are presented for direct-drive targets during injection into a dry-wall reaction chamber. The baseline approach to target survival utilizes highly reflective targets along with a substantially lower chamber wall temperature and fill-gas pressure than previously assumed. Recently developed high-Z material coatings with high heat reflectivity are discussed and characterized. The article also presents alternate target protection methods that could be developed if targets with inherent survival features cannot be obtained within a reasonable time span. (author)

  16. Ultrasmooth plasma polymerized coatings for laser-fusion targets

    International Nuclear Information System (INIS)

    Letts, S.A.; Myers, D.W.; Witt, L.A.

    1980-01-01

    Coatings for laser fusion targets were deposited up to 135 μm thick by plasma polymerization onto 140 μm diameter DT filled glass microspheres. Ultrasmooth surfaces (no defect higher than 0.1 μm) were achieved by eliminating particulate contamination. Process generated particles were eliminated by determining the optimum operating conditions of power, gas flow, and pressure, and maintaining these conditions through feedback control. From a study of coating defects grown over known surface irregularities, a quantitative relationship between irregularity size, film thickness, and defect size was determined. This relationship was used to set standards for the maximum microshell surface irregularity tolerable in the production of hydrocarbon or fluorocarbon coated laser fusion targets

  17. Metal coatings for laser fusion targets by electroplating

    International Nuclear Information System (INIS)

    Illige, J.D.; Yu, C.M.; Letts, S.A.

    1980-01-01

    Metal coated laser fusion targets must be dense, uniform spherically symmetric to within a few percent of their diameters and smooth to better than a few tenths of a micron. Electroplating offers some unique advantages including low temperature deposition, a wide choice of elements and substantial industrial plating technology. We have evaluatd electroless and electroplating systems for gold and copper, identified the factors responsible for small grain size, and plated glass microspheres with both metals to achieve smooth surfaces and highly symmetric coatings. We have developed plating cells which sustain the microspheres in continuous random motion during plating. We have established techniques for deposition of the initial conductive adherent layer on the glass microsphere surface. Coatings as thick as 15 μm have been made. The equipment is simple, relatively inexpensive and may be adopted for high volume production of laser fusion targets

  18. Method for nondestructive fuel assay of laser fusion targets

    Science.gov (United States)

    Farnum, Eugene H.; Fries, R. Jay

    1976-01-01

    A method for nondestructively determining the deuterium and tritium content of laser fusion targets by counting the x rays produced by the interaction of tritium beta particles with the walls of the microballoons used to contain the deuterium and tritium gas mixture under high pressure. The x rays provide a direct measure of the tritium content and a means for calculating the deuterium content using the initial known D-T ratio and the known deuterium and tritium diffusion rates.

  19. Method for nondestructive fuel assay of laser fusion targets

    International Nuclear Information System (INIS)

    Farnum, E.H.; Fries, R.J.

    1976-01-01

    A method is described for nondestructively determining the deuterium and tritium content of laser fusion targets by counting the x rays produced by the interaction of tritium beta particles with the walls of the microballoons used to contain the deuterium and tritium gas mixture under high pressure. The x rays provide a direct measure of the tritium content and a means for calculating the deuterium content using the initial known D-T ratio and the known deuterium and tritium diffusion rates

  20. Polyvinyl alcohol coating of polystyrene inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Annamalai, P.; Lee, M.C.; Crawley, R.L.; Downs, R.L.

    1985-01-01

    An inertial confinement fusion (ICF) target made of polystyrene is first levitated in an acoustic field. The surface of the target is then etched using an appropriate solution (e.g., cyclohexane) to enhance the wetting characteristics. A specially prepared polyvinyl alcohol solution is atomized using an acoustic atomizer and deposited on the surface of the target. The solution is air dried to form a thin coating (2 μm) on the target (outside diameter approx.350--850 μm). Thicker coatings are obtained by repeated applications of the coating solution. Preliminary results indicate that uniform coatings may be achievable on the targets with a background surface smoothness in the order of 1000 A

  1. Ceramic sphere-pac breeder design for fusion blankets

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Sullivan, J.D.

    1991-01-01

    Randomly packed beds of ceramic spheres are a practical approach to surrounding fusion plasmas with tritium-breeding material. This paper examines the general properties of sphere-pac beds for application in fusion breeder blankets. The design considerations and models are reviewed for packing, tritium breeding and recovery, thermal conductivity, purge-gas pressure drop, mechanical behavior and fabrication. The design correlations are compared against available fusion ceramic data. Specific conclusions are that ternary (three-size) beds are not attractive for fusion blankets, and that the fusion spheres should be as large as possible subject primarily to packing constraints. (orig.)

  2. Information management data base for fusion target fabrication processes

    International Nuclear Information System (INIS)

    Reynolds, J.

    1983-01-01

    A computer-based data management system has been developed to handle data associated with target fabrication processes including glass microballoon characterization, gas filling, materials coating, and storage locations. The system provides automatic data storage and computation, flexible data entry procedures, fast access, automated report generation, and secure data transfer. It resides on a CDC CYBER 175 computer and is compatible with the CDC data base language Query Update, but is based on custom fortran software interacting directly with the CYBER's file management system. The described data base maintains detailed, accurate, and readily available records of fusion targets information

  3. Evaluation of divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

    Ferrari, M.; Giancarli, L.; Kleefeldt, K.; Nardi, C.; Roedig, M.; Reimann, J.; Salavy, J.F.

    2001-01-01

    In the frame of the preliminary study of plants suitable for the energy production from the fusion power, particular emphasis has been given on the divertor studies. Since a significant percentage of the power generated from the fusion process is absorbed in the divertor, the thermal efficiency of the power conversion cycle requires a high coolant outlet temperature of the divertor, leading to solutions that are different from those adopted for the present experimental fusion plants. Therefore, copper alloys having extremely high thermal conductivity, cannot be used as structural material for this kind of devices. The most suitable coolants to be used in the divertor are water, helium and liquid metals. A conceptual design study has been developed for each of these three fluids, with the aim to evaluate the maximum allowable thermal flux at the divertor target plate and the R and D requirements for each solution. While a water-cooled divertor can be designed with a limited R and D effort, the development of helium or liquid metal cooled divertors requires a more engaging R and D program

  4. Fabrication and characterization of cryogenic targets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Rieger, H.; Kim, K.

    1979-08-01

    A new technique has been developed which is capable of fabricating uniform cryogenic targets for use in inertial confinement fusion. The essence of the technique is to directly wet a target with a cold helium gas jet, which results in freezing of the DT mixture contained in the target. A controlled amount of current is pulsed through a heater wire surrounding the target, giving rise to fast evaporation and refreezing of the DT-condensate into a uniform layer. Experiments, which have been performed with D 2 -filled glass microshells, successfully produce uniform layers of both liquid and solid D 2 inside the glass shells. A set of data illustrating the technique is presented and analyzed

  5. Henipavirus Mediated Membrane Fusion, Virus Entry and Targeted Therapeutics

    Directory of Open Access Journals (Sweden)

    Dimitar B. Nikolov

    2012-02-01

    Full Text Available The Paramyxoviridae genus Henipavirus is presently represented by the type species Hendra and Nipah viruses which are both recently emerged zoonotic viral pathogens responsible for repeated outbreaks associated with high morbidity and mortality in Australia, Southeast Asia, India and Bangladesh. These enveloped viruses bind and enter host target cells through the coordinated activities of their attachment (G and class I fusion (F envelope glycoproteins. The henipavirus G glycoprotein interacts with host cellular B class ephrins, triggering conformational alterations in G that lead to the activation of the F glycoprotein, which facilitates the membrane fusion process. Using the recently published structures of HeV-G and NiV-G and other paramyxovirus glycoproteins, we review the features of the henipavirus envelope glycoproteins that appear essential for mediating the viral fusion process, including receptor binding, G-F interaction, F activation, with an emphasis on G and the mutations that disrupt viral infectivity. Finally, recent candidate therapeutics for henipavirus-mediated disease are summarized in light of their ability to inhibit HeV and NiV entry by targeting their G and F glycoproteins.

  6. Power plant production of inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.; Johnson, W.L.

    1979-01-01

    Many of the current techniques for fabricating experimental targets appear to be directly extendable to the high-rate, low-cost production of reactor targets. This report describes several new techniques that, in conjunction with the expansion of existing techniques, can constitute a target factory. We have evaluated this concept on the basis of a generalized reactor target design and the projected specifications of reactor-grade targets

  7. Systems-design and energy-balance considerations for impact fusion

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Miller, R.L.

    1979-01-01

    Areas of concern and potential problems for impact fusion are qualitatively considered within an overall systems context. A parametric and qualitative description of the general energy balance and systems considerations for an Impact Fusion Reactor (IFR) design is discussed. Reactor systems design considerations for an IFR are presented. An attempt to assess the IFR viability is made based on highly simplified but limiting projectile-target energy balances and thermonuclear burn models

  8. Ultrasmooth plasma polymerized coatings for laser fusion targets

    International Nuclear Information System (INIS)

    Letts, S.A.; Myers, D.W.; Witt, L.A.

    1980-01-01

    Coatings for laser fusion were deposited up to 135μm thick by plasma polymerization onto 140 μm diameter DT filled glass microspheres. Ultrasmooth surfaces (no defect higher than 0.1 μm) were achieved by eliminating particulate contamination. Process generated particles were eliminated by determining the optimum operating conditions of power (20 watts), gas flow (0.3 sccm trans-2-butene, 10.0 sccm hydrogen), and pressure (75 millitorr), and maintaining these conditions through feedback control. From a study of coating defects grown over known surface irregularities, a quantitative relationship between irregularity size, film thickness, and defect size was determined. This relationship was used to set standards for the maximum microshell surface irregularity tolerable in the production of hydrocarbon or fluorocarbon coated laser fusion targets

  9. On the efficiency of conical targets for laser thermonuclear fusion

    International Nuclear Information System (INIS)

    Borovskij, A.V.; Korobkin, V.V.

    1981-01-01

    Advantages and drawbacks of conical targets (CT) for laser fusion (LF) are discussed. Possibility of the laser power reduction, laser pulse lengthening and neutron yield increase are analyzed for an ideal conical target with absolutely rigid and heat-proof walls as compared to a spherical target of the same mass. A simple theory is suggested which makes it possible to take into account an effect of walls on the fusion process in the conical target with gaseous fuel and heavy shell. Energy losses due to wall deformations and heat conduction are estimated. An influence of these effects on the neutron yield is estimated. CT used in the LF experiments are found to have serious drawbacks in comparison with spherical ones. These drawbacks are connected with the effect of walls on the processes taking place in CT. Analysis of CT, for which the effect of walls is not significant, points up some definite advantages of CT as compared with spherical one. These advantages are the possibility of laser pulse lengthening and laser power reduction in comparison with the irradiation of a sphere of an equal mass. These two positive qualities are connected with the fact that CT has large linear dimensions [ru

  10. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    Schulte, S.C.; Willke, T.L.; Young, J.R.

    1978-05-01

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  11. High pressure deuterium-tritium gas target vessels for muon-catalyzed fusion experiments

    International Nuclear Information System (INIS)

    Caffrey, A.J.; Spaletta, H.W.; Ware, A.G.; Zabriskie, J.M.; Hardwick, D.A.; Maltrud, H.R.; Paciotti, M.A.

    1989-01-01

    In experimental studies of muon-catalyzed fusion, the density of the hydrogen gas mixture is an important parameter. Catalysis of up to 150 fusions per muon has been observed in deuterium-tritium gas mixtures at liquid hydrogen density; at room temperature, such densities require a target gas pressure of the order of 1000 atmospheres (100 MPa, 15,000 psi). We report here the design considerations for hydrogen gas target vessels for muon-catalyzed fusion experiments that operate at 1000 and 10,000 atmospheres. The 1000 atmosphere high pressure target vessels are fabricated of Type A-286 stainless steel and lined with oxygen-free, high-conductivity (OFHC) copper to provide a barrier to hydrogen permeation of the stainless steel. The 10,000 atmosphere ultrahigh pressure target vessels are made from 18Ni (200 grade) maraging steel and are lined with OFHC copper, again to prevent hydrogen permeation of the steel. In addition to target design features, operating requirements, fabrication procedures, and secondary containment are discussed. 13 refs., 3 figs., 1 tab

  12. Design of the MYRRHA Spallation Target Assembly

    International Nuclear Information System (INIS)

    Keijers, S.; Fernandez, R.; Stankovskiy, A.; Kennedy, G.; Van Tichelen, K.

    2015-01-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is a multi-purpose research facility currently being developed at SCK.CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level. As a flexible irradiation facility, the MYRRHA research reactor will be able to work in both critical and subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material research for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by Lead Bismuth Eutectic (LBE) and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. This paper describes the evolution of the MYRRHA spallation target design. In the early phase of the MYRRHA project (XT-ADS), the target design was based on a dedicated spallation loop inside the primary reactor vessel. Within the core, the 3 central fuel assembly positions were occupied by the spallation target, which enabled a windowless design created by a free surface of LBE facing the proton beam. The windowless option was preferred because of high heat loads in combination with severe irradiation damage in the target region would result in unacceptably short lifetimes of a target window. The LBE in the loop served as spallation target and as target coolant, but was separated from the LBE cooling the reactor core. The loop was equipped with its own pump, heat exchanger and conditioning system. The change from cyclotron to linear accelerator allowed the increase in proton energy from 350 MeV to 600 MeV. This modification led to an important reduction of the specific heat load at the target level and an improvement of the neutronic performance. In addition to

  13. Generation and compression of a target plasma for magnetized target fusion

    International Nuclear Information System (INIS)

    Kirkpatrick, R.C.; Lindemuth, I.R.; Sheehey, P.T.

    1998-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Magnetized target fusion (MTF) is intermediate between the two very different approaches to fusion: inertial and magnetic confinement fusion (ICF and MCF). Results from collaboration with a Russian MTF team on their MAGO experiments suggest they have a target plasma suitable for compression to provide an MTF proof of principle. This LDRD project had tow main objectives: first, to provide a computational basis for experimental investigation of an alternative MTF plasma, and second to explore the physics and computational needs for a continuing program. Secondary objectives included analytic and computational support for MTF experiments. The first objective was fulfilled. The second main objective has several facets to be described in the body of this report. Finally, the authors have developed tools for analyzing data collected on the MAGO and LDRD experiments, and have tested them on limited MAGO data

  14. Target design for shock ignition

    International Nuclear Information System (INIS)

    Schurtz, G; Ribeyre, X; Lafon, M

    2010-01-01

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

  15. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

    The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

  16. Accelerator conceptual design of the international fusion materials irradiation facility

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, M.; Kinsho, M. [Japan Atomic Energy Res. Inst., Tokai, Ibaraki (Japan). Intense Neutron Source Lab.; Jameson, R.A.; Blind, B. [Los Alamos National Lab., NM (United States); Teplyakov, V. [Institute for High Energy Physics, Moscow (Russian Federation); Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J. [Northrop Grumman Corp., Bethpage, NY (United States); Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K. [Johann Wolfgang Goethe Univ., Frankfurt (Germany). Inst. fur Angewandte Phys.; Ferdinand, R.; Lagniel, J.-M. [CEA Saclay LNS, Gif-sur-Yvette (France); Miyahara, A. [Teikyo Univ., Tokyo (Japan); Olivier, M. [CEA DSM, Saclay, Gif-sur-Yvette (France); Piechowiak, E. [Northrop Grumman Corp., Baltimore, MD (United States); Tanabe, Y. [Toshiba Corp., Tsurumi-ku, Yokohama (Japan)

    1998-10-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.) 8 refs.

  17. Accelerator conceptual design of the international fusion materials irradiation facility

    International Nuclear Information System (INIS)

    Sugimoto, M.; Kinsho, M.; Teplyakov, V.; Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J.; Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K.; Miyahara, A.; Olivier, M.; Piechowiak, E.; Tanabe, Y.

    1998-01-01

    The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.)

  18. Radiological safety design considerations for fusion research experiments

    International Nuclear Information System (INIS)

    Crase, K.W.; Singh, M.S.

    1979-01-01

    A wide variety of fusion research experiments are in the planning or construction stages. Two such experiments, the Nova Laser Fusion Facility and the Mirror Fusion Test Facility (MFTF), are currently under construction at Lawrence Livermore Laboratory. Although the plasma chamber vault for MFTF and the Nova target room will have thick concrete walls and roofs, the radiation safety problems are made complex by the numerous requirements for shield wall penetrations. This paper addresses radiation safety considerations for the MFTF and Nova experiments, and the need for integrated safety considerations and safety technology development during the planning stages of fusion experiments

  19. Review of laser mega joule target area: Design and processes

    International Nuclear Information System (INIS)

    Geitzholz, M.; Lanternier, C.

    2006-01-01

    The Laser Mega Joule (LMJ) target area is currently designed to achieve ignition and significant fusion gain in laboratory. LMJ will be composed of 240 identical large 370 mm * 370 mm square laser beams. These beams will focus 2 mega-joules of energy at the wavelength of 351 nm on the center of an experiment chamber. Design studies for target equipment are well advanced, target chamber and target holder (concrete) works have already begun. A detailed overview of the target area equipment is presented: target chamber, frame, diagnostic inserter manipulator, final optic assembly, dual diagnostic and laser reference, non cryogenic target positioner. Recent technical and architectural choices are detailed including safety transfers and alignment processes (target, laser and diagnostic). All this target equipment allows us to optimize shot chrono-gram, from target metrology to the shot, including calibration process. (authors)

  20. Conceptual design of imploding liner fusion reactors

    International Nuclear Information System (INIS)

    Turchi, P.J.; Robson, A.E.

    1976-01-01

    The basic new ingredient is the concept of rotationally stabilized liquid metal liners accelerated with free pistons. The liner motion is constrained on its outer surface by the pistons, laterally by channel walls, during acceleration, and on its inner surface, where megagauss field levels are attained by the centrifugal motion of the liner material. In this way, stable, reversible motion of the liner should be possible, permitting repetitive, pulsed operation at interior pressures far greater than can be allowed in static conductor systems. Such higher operating pressures permit the use of simple plasma geometries, such as theta pinches, with greatly reduced dimensions. Furthermore, the implosion of thick, lithium-bearing liners with large radial compression ratios inherently provides the plasma with a surrounding blanket of neutron absorbing liquid metal, thereby substantially reducing the problems of induced radioactivity and first wall damage that haunt conventional fusion reactor designs. The following article discusses the basic operation of liner reactors and several important features influencing their design

  1. Fusion reactor blanket-main design aspects

    International Nuclear Information System (INIS)

    Strebkov, Yu.; Sidorov, A.; Danilov, I.

    1994-01-01

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

  2. Inertial Confinement Fusion Target Component Fabrication and Technology Development report

    International Nuclear Information System (INIS)

    Steinman, D.

    1994-03-01

    On December 30, 1990, the US Department of Energy entered into a contract with General Atomics (GA) to be the Inertial Confinement Fusion Target Component Fabrication and Technology Development Support contractor. This report documents the technical activities which took place under this contract during the period of October 1, 1992 through September 30, 1993. During this period, GA was assigned 18 tasks in support of the Inertial Confinement Fusion program and its laboratories. These tasks included ''Capabilities Activation'' and ''Capabilities Demonstration'' to enable us to begin production of glass and composite polymer capsules. Capsule delivery tasks included ''Small Glass Shell Deliveries'' and ''Composite Polymer Capsules'' for Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). We also were asked to provide direct ''Onsite Support'' at LLNL and LANL. We continued planning for the transfer of ''Micromachining Equipment from Rocky Flats'' and established ''Target Component Micromachining and Electroplating Facilities'' at GA. We fabricated over 1100 films and filters of 11 types for Sandia National Laboratory and provided full-time onsite engineering support for target fabrication and characterization. We initiated development of methods to make targets for the Naval Research Laboratory. We investigated spherical interferometry, built an automated capsule sorter, and developed an apparatus for calorimetric measurement of fuel fill for LLNL. We assisted LANL in the ''Characterization of Opaque b-Layered Targets.'' We developed deuterated and UV-opaque polymers for use by the University of Rochester's Laboratory for Laser Energetics (UR/LLE) and devised a triple-orifice droplet generator to demonstrate the controlled-mass nature of the microencapsulation process

  3. Determination of the pr of laser fusion targets using the α-particle TOF technique

    International Nuclear Information System (INIS)

    Slivinsky, V.W.; Lent, E.; Shay, H.D.; Manes, K.R.

    1975-01-01

    A computer code was written to describe the alpha particle energy loss. The problem of a symmetric compression of the DT gas by an exploding microsphere is analyzed. The code calculates the energy spectrum of a Gaussian distribution of alpha particles after passing through the compressed gas and the exploded glass. The calculations are being used to determine design parameters for diagnostic instruments for measuring charged particle energy distributions from laser fusion targets

  4. Resolving key heavy-ion fusion target issues with relativistic heavy-ion research accelerators

    International Nuclear Information System (INIS)

    Arnold, R.C.

    1988-01-01

    Heavy-ion accelerators designed for relativistic nuclear research experiments can also be adapted for target research in heavy-ion driver inertial fusion. Needle-shaped plasmas can be created that are adequate for studying basic properties of matter at high energy density. Although the ion range is very long, the specific deposited power nevertheless increases with kinetic energy, as the focus spot can be made smaller and more ions can be accumulated in larger rings

  5. Effect of experimentally observed hydrogenic fractionation on inertial confinement fusion ignition target performance

    International Nuclear Information System (INIS)

    McKenty, P. W.; Wittman, M. D.; Harding, D. R.

    2006-01-01

    The need of cryogenic hydrogenic fuels in inertial confinement fusion (ICF) ignition targets has been long been established. Efficient implosion of such targets has mandated keeping the adiabat of the main fuel layer at low levels to ensure drive energies are kept at reasonable minima. The use of cryogenic fuels helps meet this requirement and has therefore become the standard in most ICF ignition designs. To date most theoretical ICF ignition target designs have assumed a homogeneous layer of deuterium-tritium (DT) fuel kept slightly below the triple point. However, recent work has indicated that, as cryogenic fuel layers are formed inside an ICF capsule, isotopic dissociation of the tritium (T), deuterium (D), and DT takes place leading to a 'fractionation' of the final ice layer. This paper will numerically investigate the effects that various scenarios of fractionation have on hot-spot formation, ignition, and burn in ICF ignition target designs

  6. Fusion energy research with lasers, direct drive targets, and dry wall chambers

    International Nuclear Information System (INIS)

    Sethian, J.D.; Obenschain, S.P.; Myers, M.

    2003-01-01

    We are carrying out a coordinated, focused effort to develop Laser Inertial Fusion Energy. The key components are developed in concert with one another and the science and engineering issues are addressed concurrently. Significant progress has been made in this program: We are evaluating target designs that show it could be possible to achieve the high gains (>100) needed for a practical fusion system. These have a low density CH foam that is wicked with solid DT, and over coated with a thin high-Z layer. Significant advances have been made with the two types of laser are being developed: Krypton Fluoride (KrF) gas lasers and Diode Pumped Solid State Lasers (DPPSL). Both have the potential to meet the fusion energy requirements for rep-rate, efficiency, durability and cost. This paper also presents the advances in development of chamber operating windows (target survival plus no wall erosion), final optics (aluminum at grazing incidence has high reflectivity and exceeds required laser damage threshold), target fabrication (advanced foams and high Z overcoats), and target injection (new facility for target injection and tracking studies). (author)

  7. Development of position measurement unit for flying inertial fusion energy target

    International Nuclear Information System (INIS)

    Tsuji, R; Endo, T; Yoshida, H; Norimatsu, T

    2016-01-01

    We have reported the present status in the development of a position measurement unit (PMU) for a flying inertial fusion energy (IFE) target. The PMU, which uses Arago spot phenomena, is designed to have a measurement accuracy smaller than 1 μm. By employing divergent, pulsed orthogonal laser beam illumination, we can measure the time and the target position at the pulsed illumination. The two-dimensional Arago spot image is compressed into one-dimensional image by a cylindrical lens for real-time processing. The PMU are set along the injection path of the flying target. The local positions of the target in each PMU are transferred to the controller and analysed to calculate the target trajectory. Two methods are presented to calculate the arrival time and the arrival position of the target at the reactor centre. (paper)

  8. Development of position measurement unit for flying inertial fusion energy target

    Science.gov (United States)

    Tsuji, R.; Endo, T.; Yoshida, H.; Norimatsu, T.

    2016-03-01

    We have reported the present status in the development of a position measurement unit (PMU) for a flying inertial fusion energy (IFE) target. The PMU, which uses Arago spot phenomena, is designed to have a measurement accuracy smaller than 1 μm. By employing divergent, pulsed orthogonal laser beam illumination, we can measure the time and the target position at the pulsed illumination. The two-dimensional Arago spot image is compressed into one-dimensional image by a cylindrical lens for real-time processing. The PMU are set along the injection path of the flying target. The local positions of the target in each PMU are transferred to the controller and analysed to calculate the target trajectory. Two methods are presented to calculate the arrival time and the arrival position of the target at the reactor centre.

  9. Investigation of fusion gain in fast ignition with conical targets

    Directory of Open Access Journals (Sweden)

    MJ Tabatabaei

    2011-03-01

    Full Text Available Fast ignition is a new scheme for inertial confinement fusion (ICF. In this scheme, at first the interaction of ultraintense laser beam with the hohlraum wall surrounding a capsule containing deuterium-tritium (D-T fuel causes implosion and compression of fuel to high density and then laser produced protons penetrate in the compressed fuel and deposit their energy in it as the ignition hot spot is created. In this paper, following the energy gain of spherical target and considering relationship of the burn fraction to burn duration, we have obtained the energy gain of conical targets characterized by the angle β, and found a hemispherical capsule (β=π/2 has a gain as high as 96% of that of the whole spherical capsule. The results obtained in this study are qualitatively consistent with Atzeni et al.'s studies of simulations.

  10. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    1987-01-01

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  11. Bulk-shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Johnson, D.L.; Huang, S.T.

    1982-07-01

    The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

  12. Review of mirror fusion reactor designs

    International Nuclear Information System (INIS)

    Bender, D.J.

    1977-01-01

    Three magnetic confinement concepts, based on the mirror principle, are described. These mirror concepts are summarized as follows: (1) fusion-fission hybrid reactor, (2) tandem mirror reactor, and (3) reversed field mirror reactor

  13. Process for manufacture of inertial confinement fusion targets and resulting product

    International Nuclear Information System (INIS)

    Solomon, D.E.; Wise, K.D.; Wuttke, G.H.; Masnari, N.A.; Rensel, W.B.; Robinson, M.G.

    1980-01-01

    A method of manufacturing inertial confinement fusion targets is described which is adaptable for high volume production of low cost targets in a wide variety of sizes. The targets include a spherical pellet of fusion fuel surrounded by a protective concentric shell. (UK)

  14. System for automatic x-ray-image analysis, measurement, and sorting of laser fusion targets

    International Nuclear Information System (INIS)

    Singleton, R.M.; Perkins, D.E.; Willenborg, D.L.

    1980-01-01

    This paper describes the Automatic X-Ray Image Analysis and Sorting (AXIAS) system which is designed to analyze and measure x-ray images of opaque hollow microspheres used as laser fusion targets. The x-ray images are first recorded on a high resolution film plate. The AXIAS system then digitizes and processes the images to accurately measure the target parameters and defects. The primary goals of the AXIAS system are: to provide extremely accurate and rapid measurements, to engineer a practical system for a routine production environment and to furnish the capability of automatically measuring an array of images for sorting and selection

  15. Physics of laser fusion. Volume II. Diagnostics of experiments on laser fusion targets at LLNL

    Energy Technology Data Exchange (ETDEWEB)

    Ahlstrom, H.G.

    1982-01-01

    These notes present the experimental basis and status for laser fusion as developed at LLNL. There are two other volumes in this series: Vol. I, by C.E. Max, presents the theoretical laser-plasma interaction physics; Vol. III, by J.F. Holzrichter et al., presents the theory and design of high-power pulsed lasers. A fourth volume will present the theoretical implosion physics. The notes consist of six sections. The first, an introductory section, provides some of the history of inertial fusion and a simple explanation of the concepts involved. The second section presents an extensive discussion of diagnostic instrumentation used in the LLNL Laser Fusion Program. The third section is a presentation of laser facilities and capabilities at LLNL. The purpose here is to define capability, not to derive how it was obtained. The fourth and fifth sections present the experimental data on laser-plasma interaction and implosion physics. The last chapter is a short projection of the future.

  16. Physics of laser fusion. Volume II. Diagnostics of experiments on laser fusion targets at LLNL

    International Nuclear Information System (INIS)

    Ahlstrom, H.G.

    1982-01-01

    These notes present the experimental basis and status for laser fusion as developed at LLNL. There are two other volumes in this series: Vol. I, by C.E. Max, presents the theoretical laser-plasma interaction physics; Vol. III, by J.F. Holzrichter et al., presents the theory and design of high-power pulsed lasers. A fourth volume will present the theoretical implosion physics. The notes consist of six sections. The first, an introductory section, provides some of the history of inertial fusion and a simple explanation of the concepts involved. The second section presents an extensive discussion of diagnostic instrumentation used in the LLNL Laser Fusion Program. The third section is a presentation of laser facilities and capabilities at LLNL. The purpose here is to define capability, not to derive how it was obtained. The fourth and fifth sections present the experimental data on laser-plasma interaction and implosion physics. The last chapter is a short projection of the future

  17. Radiological design criteria for fusion power test facilities

    International Nuclear Information System (INIS)

    Singh, M.S.; Campbell, G.W.

    1982-01-01

    The quest for fusion power and understanding of plasma physics has resulted in planning, design, and construction of several major fusion power test facilities, based largely on magnetic and inertial confinement concepts. We have considered radiological design aspects of the Joint European Torus (JET), Livermore Mirror and Inertial Fusion projects, and Princeton Tokamak. Our analyses on radiological design criteria cover acceptable exposure levels at the site boundary, man-rem doses for plant personnel and population at large, based upon experience gained for the fission reactors, and on considerations of cost-benefit analyses

  18. Fusion reactor design: On the road to commercialization

    International Nuclear Information System (INIS)

    Kulcinski, G.L.

    1984-01-01

    The worldwide effort in fusion is now approximately 2 billion dollars per year and over 12 billion dollars has been invested since 1951 in developing this energy source for the 21st century. A vital component of the past efforts in fusion research has been the conceptual design activities performed by scientists and engineers around the world. Almost 80 such major designs of Tokamak, Mirror, Laser and Ion Beam Reactors have been published and this article discusses how recent conceptual designs have afftected our perception of future fusion reactor performance. (orig.) [de

  19. Optical and mechanical design of beam-target coupling sensor

    International Nuclear Information System (INIS)

    Wang Liquan; Li Tian'en; Feng Bin; Xiang Yong; Li Keyu; Zhong Wei; Liu Guodong

    2012-01-01

    A sensor based on conjugate principle has been designed for matching the light beams and the target in inertial confinement fusion. It can avoid the direct illumination of the simulation collimating light on the target under test in targeting processes. This paper introduces the optical and mechanical design of the sensor, according to its design functions and working principle. The resolution of the optical images obtained in experiments reaches 6 μm and the beam-target matching accuracy is 8.8 μm. The sensor has been successfully applied to the Shenguang-Ⅲ facility. Statistical analyses of the four-hole CH target images derived with pinhole camera shows that the targeting accuracy of the facility is better than 25 μm, satisfying the design requirements. (authors)

  20. Initial trade and design studies for the fusion engineering device

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Steiner, D.; Smith, G.E.

    1981-06-01

    The Magnetic Fusion Energy Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. The Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), initiated a program of trade and design studies in October 1980 to support the selection of the FED concept. This document presents the results of these initial trade and design studies. Based on these results, a baseline configuration has been identified and the Design Center effort for the remainder of the fiscal year will be devoted to the development of a self-consistent FED design description

  1. Safety considerations in next step fusion design and beyond

    International Nuclear Information System (INIS)

    Holland, D.F.

    1990-01-01

    Recent U.S. and international design studies provide insights into the potential safety and environmental advantages of fusion as well as the development needed to realize this potential. We in the Fusion Safety Program at EG ampersand G Idaho have analyzed the Compact Ignition Tokamak (CIT), the International Thermonuclear Engineering Reactor (ITER), and the Advanced Reactor Innovative Engineering Study (ARIES). I have reviewed these three designs to determine issues related to meeting the safety and the environmental goals that guide fusion development in the U.S. The paper lists safety and environmental issues that are generic to fusion and approaches to favorably resolve each issue. The technical developments that have the highest potential of contributing to improving the safety and environmental attractiveness of fusion are identified and discussed. These developments are in the areas of low-activation materials, plasma- facing components, and plasma physics relating to off-normal plasma events and tritium burn-up. 8 refs., 7 tabs

  2. Conceptual design of light ion beam inertia nuclear fusion reactors

    International Nuclear Information System (INIS)

    1983-07-01

    Light ion beam, inertia nuclear fusion system drew attention recently as one of the nuclear fusion systems for power reactors in the history of the research on nuclear fusion. Its beginning seemed to be the judgement that the implosion of fusion fuel pellets with light ions can be realized with the light ions which can be obtained in view of accelerator techniques. Of course, in order to generate practically usable nuclear fusion reaction by this system and maintain it, many technical difficulties must be overcome. This research was carried out for the purpose of discovering such technical problems and searching for their solution. At the time of doing the works, the following policy was adopted. Though their is the difference of fine and rough, the design of a whole reactor system is performed conformably. In order to make comparison with other reactor types and nuclear fusion systems, the design is carried out as the power plant of about one million kWe output. As the extent of the design, the works at conceptual design stage are performed to present the concept of design which satisfies the required function. Basically, the design is made from conservative standpoint. This research of design was started in 1981, and in fiscal 1982, the mutual adjustment among the design of respective parts was performed on the basis of the results in 1981, and the possible revision and new proposal were investigated. (Kako, I.)

  3. Charged--particle beam implosion of fusion targets

    International Nuclear Information System (INIS)

    Clauser, M.J.; Sweeney, M.A.

    1975-01-01

    This paper discusses the calculated behavior of fusion targets consisting of solid shells filled with DT gas, irradiated by high power electron or ion beams. The current required for breakeven with gold shells is 500 to 1000 MA, independent of target radius and nearly independent of beam voltage in the 1 / 2 to 1 MeV range. Above 1 MeV the breakeven current increases because of the increased bremsstrahlung production by the beam electrons. By using a diamond ablator and a gold pusher, the breakeven current is reduced to 220 MA. The ion current required for breakeven (about 10 MA of protons) is independent of proton voltage above 10 MeV with gold shell targets. Below 10 MeV the range of the proton becomes too short for efficient coupling, and the required current increases, but the power does not. Various aspects of the symmetry and stability of the implosion are discussed. One finds that the relatively long deposition lengths of electrons result in relatively small growths of the Rayleigh--Taylor instability during the acceleration of the pusher, resulting in a relatively stable implosion

  4. Microfabricated Ion Beam Drivers for Magnetized Target Fusion

    Science.gov (United States)

    Persaud, Arun; Seidl, Peter; Ji, Qing; Ardanuc, Serhan; Miller, Joseph; Lal, Amit; Schenkel, Thomas

    2015-11-01

    Efficient, low-cost drivers are important for Magnetized Target Fusion (MTF). Ion beams offer a high degree of control to deliver the required mega joules of driver energy for MTF and they can be matched to several types of magnetized fuel targets, including compact toroids and solid targets. We describe an ion beam driver approach based on the MEQALAC concept (Multiple Electrostatic Quadrupole Array Linear Accelerator) with many beamlets in an array of micro-fabricated channels. The channels consist of a lattice of electrostatic quadrupoles (ESQ) for focusing and of radio-frequency (RF) electrodes for ion acceleration. Simulations with particle-in-cell and beam envelope codes predict >10x higher current densities compared to state-of-the-art ion accelerators. This increase results from dividing the total ion beam current up into many beamlets to control space charge forces. Focusing elements can be biased taking advantage of high breakdown electric fields in sub-mm structures formed using MEMS techniques (Micro-Electro-Mechanical Systems). We will present results on ion beam transport and acceleration in MEMS based beamlets. Acknowledgments: This work is supported by the U.S. DOE under Contract No. DE-AC02-05CH11231.

  5. Conceptual design study of Fusion Experimental Reactor (FY87FER)

    International Nuclear Information System (INIS)

    1988-05-01

    The design study of Fusion Experimental Reactor(FER) which has been proposed to be the next step fusion device has been conducted by JAERI Reactor System Laboratory since 1982 and by FER design team since 1984. This is the final report of the FER design team program and describes the results obtained in FY1987 (partially in FY1986) activities. The contents of this report consist of the reference design which is based on the guideline in FY1986 by the Subcomitees set up in Nuclear Fusion Council of Atomic Energy Commission of Japan, the Low-Physics-Risk reactor design for achieving physics mission more reliably and the system study of FER design candidates including above two designs. (author)

  6. Inertial confinement fusion target insertion concepts for the National Ignition Facility

    International Nuclear Information System (INIS)

    Laughon, G.J.; Schultz, K.R.

    1996-01-01

    The National Ignition Facility (NIF) will be used to demonstrate fusion ignition in a laboratory environment in order to support development of inertial fusion as a potential fusion energy source for civilian use. However, target insertion must first be addressed before inertial fusion can become a practical energy source. Since target insertion systems currently utilized are not suitable for multiple shots in quick succession, insertion concepts involving free-falling and artificially accelerated targets are developed and evaluated against a set of predetermined guidelines. It is shown that a system involving a fast retraction positioner would be suitable. 5 refs., 4 figs

  7. An Approach to Automated Fusion System Design and Adaptation

    Directory of Open Access Journals (Sweden)

    Alexander Fritze

    2017-03-01

    Full Text Available Industrial applications are in transition towards modular and flexible architectures that are capable of self-configuration and -optimisation. This is due to the demand of mass customisation and the increasing complexity of industrial systems. The conversion to modular systems is related to challenges in all disciplines. Consequently, diverse tasks such as information processing, extensive networking, or system monitoring using sensor and information fusion systems need to be reconsidered. The focus of this contribution is on distributed sensor and information fusion systems for system monitoring, which must reflect the increasing flexibility of fusion systems. This contribution thus proposes an approach, which relies on a network of self-descriptive intelligent sensor nodes, for the automatic design and update of sensor and information fusion systems. This article encompasses the fusion system configuration and adaptation as well as communication aspects. Manual interaction with the flexibly changing system is reduced to a minimum.

  8. An Approach to Automated Fusion System Design and Adaptation.

    Science.gov (United States)

    Fritze, Alexander; Mönks, Uwe; Holst, Christoph-Alexander; Lohweg, Volker

    2017-03-16

    Industrial applications are in transition towards modular and flexible architectures that are capable of self-configuration and -optimisation. This is due to the demand of mass customisation and the increasing complexity of industrial systems. The conversion to modular systems is related to challenges in all disciplines. Consequently, diverse tasks such as information processing, extensive networking, or system monitoring using sensor and information fusion systems need to be reconsidered. The focus of this contribution is on distributed sensor and information fusion systems for system monitoring, which must reflect the increasing flexibility of fusion systems. This contribution thus proposes an approach, which relies on a network of self-descriptive intelligent sensor nodes, for the automatic design and update of sensor and information fusion systems. This article encompasses the fusion system configuration and adaptation as well as communication aspects. Manual interaction with the flexibly changing system is reduced to a minimum.

  9. An FPGA-based heterogeneous image fusion system design method

    Science.gov (United States)

    Song, Le; Lin, Yu-chi; Chen, Yan-hua; Zhao, Mei-rong

    2011-08-01

    Taking the advantages of FPGA's low cost and compact structure, an FPGA-based heterogeneous image fusion platform is established in this study. Altera's Cyclone IV series FPGA is adopted as the core processor of the platform, and the visible light CCD camera and infrared thermal imager are used as the image-capturing device in order to obtain dualchannel heterogeneous video images. Tailor-made image fusion algorithms such as gray-scale weighted averaging, maximum selection and minimum selection methods are analyzed and compared. VHDL language and the synchronous design method are utilized to perform a reliable RTL-level description. Altera's Quartus II 9.0 software is applied to simulate and implement the algorithm modules. The contrast experiments of various fusion algorithms show that, preferably image quality of the heterogeneous image fusion can be obtained on top of the proposed system. The applied range of the different fusion algorithms is also discussed.

  10. The Modular Point Design for Heavy Ion Fusion

    International Nuclear Information System (INIS)

    Yu, S.S.; Barnard, J.J.; Briggs, R.J.; Callahan, D.; Celata, C.M.; Chao, L.; Davidson, R.; Debonnel, C.S.; Eylon, S.; Friedman, A.; Henestroza, E.; Kaganovich, I.; Kwan, J.W.; Lee, E.P.; Leitner, M.; Logan, B.G.; Meier, W.; Peterson, P.F.; Reginato, L.; Rose, D.; Roy, P.; Waldron, W.; Welch, D.R.

    2004-01-01

    We report on an ongoing study on modular Heavy Ion Fusion drivers. The modular driver is characterized by 10 to 20 nearly identical induction linacs, each carrying a single high current beam. In this scheme, the Integrated Research Experiment (IRE) can be one of the full size induction linacs. Hence, this approach offers significant advantages in terms of driver development path. For beam transport, these modules use solenoids which are capable of carrying high line charge densities, even at low energies. A new injector concept allows compression of the beam to high line densities right at the source. The final drift compression is performed in a plasma, in which the large repulsive space charge effects are neutralized. Finally, the beam is transversely compressed onto the target, using either external solenoids or current-carrying channels (in the Assisted Pinch Mode of beam propagation). We will report on progress towards a self-consistent point design from injector to target. Considerations of driver architecture, chamber environment as well as the methodology for meeting target requirements of spot size, pulse shape and symmetry will also be described. Finally, some near-term experiments to address the key scientific issues will be discussed

  11. The Modular Point Design for Heavy Ion Fusion

    International Nuclear Information System (INIS)

    Yu, S.S.; Barnard, J.J.; Briggs, R.J.; Callahan-Miller, D.; Celata, C.M.; Chao, L.; Davidson, R.; Debonnel, C.S.; Eylon, S.; Friedman, A.; Henestroza, E.; Kaganovich, I.; Kwan, J.W.; Lee, E.P.; Leitner, M.; Logan, B.G.; Meier, W.; Peterson, P.F.; Reginato, L.; Rose, D.; Roy, P.; Waldron, W.; Welch, D.R.

    2005-01-01

    We report on an ongoing study on modular Heavy Ion Fusion drivers. The modular driver is characterized by tens (∼ 20) nearly identical induction linacs, each carrying a single high current beam. In this scheme, the Integrated Research Experiment (IRE) can be one of the full size induction linacs. Hence, this approach offers significant advantages in terms of driver development path. For beam transport, these modules use solenoids which are capable of carrying high line charge densities, even at low energies. A new injector concept allows compression of the beam to high line densities right at the source. The final drift compression is performed in a plasma, in which the large repulsive space charge effects are neutralized. Finally, the beam is transversely compressed onto the target, using either external solenoids or current-carrying channels (in the Assisted Pinch Mode of beam propagation). We will report on progress towards a self-consistent point design from injector to target. Considerations of driver architecture, chamber environment as well as the methodology for meeting target requirements of spot size, pulse shape and symmetry will also be described. Finally, some near-term experiments to address the key scientific issues will be discussed

  12. Electron and ion beam transport to fusion targets

    International Nuclear Information System (INIS)

    Freeman, J.R.; Baker, L.; Miller, P.A.; Mix, L.P.; Olsen, J.N.; Poukey, J.W.; Wright, T.P.

    1979-01-01

    ICF reactors have been proposed which incorporate a gas-filled chamber to reduce x-ray and debris loading of the first wall. Focused beams of either electrons or ions must be transported efficiently for 2-4 m to a centrally located fusion target. Laser-initiated current-carrying plasma discharge channels provide the guiding magnetic field and the charge- and current-neutralizing medium required for beam propagation. Computational studies of plasma channel formation in air using a 1-D MHD model with multigroup radiation diffusion have provided a good comparison with the expansions velocity and time dependent refractivity profile determined by holographic interferometry. Trajectory calculations have identified a beam expansion mechanism which combines with the usual ohmic dissipation to reduce somewhat the transported beam fluence for electrons. Additional trajectory calculations have been performed for both electrons and light ions to predict the limits on the particle current density which can be delivered to a central target by overlapping the many independently-generated beams. Critical features of the use of plasma channels for transport and overlap of charged particle beams are being tested experimentally with up to twelve electron beams from the Proto II accelerator

  13. Numerical simulation of performance of heavy ion inertial confinement fusion target with ellipsoidal chamber

    International Nuclear Information System (INIS)

    Basin, A.A.; Vatulin, V.V.; Vakhlamova, L.L.; Vinokurov, P.A.; Dement'ev, Yu.A.; Eliseev, G.M.; Ermolovich, V.F.; Morenko, L.Z.; Morenko, A.I.; Remizov, G.N.; Romanov, Yu.A.; Ryabikina, N.A.; Skrypnik, S.I.; Skidan, G.I.; Tikhomirov, B.P.; Shagaliev, R.M.

    1996-01-01

    To solve the design problem of an inertial thermonuclear fusion facility requires the united efforts of scientists in various countries. In the field of heavy ion fusion a collaboration between scientists in Germany and Russia is under successful development. VNIIEF possesses advanced software for numerical simulation of the processes in thermonuclear target operation. This paper describes a target design suggested and being studied by scientists of Frankfurt University and GSI which is based on 2D non-stationary calculation of the X-ray energy transport and capsule compression. The target consists of a spherical capsule with DT fuel and an ellipsoidal chamber containment. The ion beam energy is released in two fixed converters located on the chamber axis symmetricall with respect to the capsule. The X-ray field is formed on the capsule surface with a set of special shields. The basic aim of our research is to estimate the effect of gas dynamic expansion of the chamber walls, shields and capsule on the target operation. To increase the reliability of the obtained results and the assessment of probable errors in predicting radiation field parameters and the capsule state, the calculations were accomplished in a kinetic arrangement with various techniques. (orig.)

  14. Identifying heavy-ion-beam fusion design and system features with high economic leverage

    International Nuclear Information System (INIS)

    Meier, W.R.; Hogan, W.J.

    1985-01-01

    We have conducted parametric economic studies for heavy-ion-beam fusion electric power plants. We examined the effects on the cost of electricity of several design parameters: maximum achievable chamber pulse rate, driver cost, target gain, and electric conversion efficiency, and net electric power. We found with reasonable assumptions on driver cost, target gain, and electric conversion efficiency, a 2 to 3 GWe heavy-ion-beam fusion power plant, with a chamber pulse rate of 5 to 10 Hz, can be competitive with nuclear and coal power plants

  15. Heavy ion inertial fusion: interface between target gain, accelerator phase space and reactor beam transport revisited

    International Nuclear Information System (INIS)

    Barletta, W.A.; Fawley, W.M.; Judd, D.L.; Mark, J.W.K.; Yu, S.S.

    1984-01-01

    Recently revised estimates of target gain have added additional optimistic inputs to the interface between targets, accelerators and fusion chamber beam transport. But it remains valid that neutralization of the beams in the fusion chamber is useful if ion charge state Z > 1 or if > 1 kA per beamlet is to be propagated. Some engineering and economic considerations favor higher currents

  16. Space-time energy concentration and the design of DT fusion micro-explosions

    International Nuclear Information System (INIS)

    Sahlin, H.L.; Brandenburg, J.E.

    1980-03-01

    As part of the effort to employ a plasma focus as a driver for fusion micro-explosions, many target concepts were explored and extensive imposion calculations have been carried out. Some of the basic principles of micro-explosion design are presented

  17. Present status of inertial confinement fusion reactor design

    International Nuclear Information System (INIS)

    Mima, Kunioki; Ido, Shunji; Nakai, Sadao.

    1986-01-01

    Since inertial nuclear fusion reactors do not require high vacuum and high magnetic field, the structure of the reactor cavity becomes markedly simple as compared with tokamak type fusion reactors. In particular, since high vacuum is not necessary, liquid metals such as lithium and lead can be used for the first wall, and the damage of reactor structures by neutrons can be prevented. As for the core, the energy efficiency of lasers is not very high, accordingly it must be designed so that the pellet gain due to nuclear fusion becomes sufficiently high, and typically, the gain coefficient from 100 to 200 is necessary. In this paper, the perspective of pellet gain, the plan from the present status to the practical reactors, and the conceptual design of the practical reactors are discussed. The plan of fuel ignition, energy break-even and high gain by the implosion mode, of which the uncertain factor due to uneven irradiation and instability was limited to the minimum, was clarified. The scenario of the development of laser nuclear fusion reactors is presented, and the concept of the reactor system is shown. The various types of nuclear fusion-fission hybrid reactors are explained. As for the design of inertial fusion power reactors, the engineering characteristics of the core, the conceptual design, water fall type reactors and DD fuel reactors are discussed. (Kako, I.)

  18. Cost Modeling for Fabrication of Direct Drive Inertial Fusion Energy Targets

    International Nuclear Information System (INIS)

    Rickman, William Samuel; Goodin, Daniel T.

    2003-01-01

    Chemical engineering analyses are underway for a commercial-scale [1000-MW(electric)] divinyl benzene foam-based Inertial Fusion Energy (IFE) Target Fabrication Facility (TFF). This facility is designed to supply 500,000, 4-mm-outer diameter targets per day - coated via interfacial polycondensation, dried with supercritical CO 2 , sputter coated with Au and/or Pd, and filled with deuterium-tritium layered at cryogenic temperatures and injected into the fusion chamber. Such targets would be used in a direct-drive IFE power plant.The work uses manufacturing processes being developed in the laboratory, chemical engineering scaleup principles, and established cost-estimating methods. The plant conceptual design includes a process flow diagram, mass and energy balances, equipment sizing and sketches, storage tanks, and facility views.The cost estimate includes both capital and operating costs. Initial results for a TFF dedicated to one 1000-MW(electric) plant indicate that the costs per target are well within the commercially viable range. Larger TFF plants [3000 MW(electric)] are projected to lead to significantly reduced costs per injected target. Additional cost reductions are possible by producing dried, sputter-coated empty shells at a central facility that services multiple power plants.The results indicate that the installed capital cost is about $100 million and the annual operating costs will be about $20 million, for a cost per target of about $0.17 each. These design and cost projections assume that a significant process development and scaleup program is successfully completed for all of the basic unit operations included in the facility

  19. Nuclear design of a very-low-activation fusion reactor

    International Nuclear Information System (INIS)

    Cheng, E.T.; Hopkins, G.R.

    1983-06-01

    An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design

  20. A unified modeling approach for physical experiment design and optimization in laser driven inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Li, Haiyan [Mechatronics Engineering School of Guangdong University of Technology, Guangzhou 510006 (China); Huang, Yunbao, E-mail: Huangyblhy@gmail.com [Mechatronics Engineering School of Guangdong University of Technology, Guangzhou 510006 (China); Jiang, Shaoen, E-mail: Jiangshn@vip.sina.com [Laser Fusion Research Center, China Academy of Engineering Physics, Mianyang 621900 (China); Jing, Longfei, E-mail: scmyking_2008@163.com [Laser Fusion Research Center, China Academy of Engineering Physics, Mianyang 621900 (China); Tianxuan, Huang; Ding, Yongkun [Laser Fusion Research Center, China Academy of Engineering Physics, Mianyang 621900 (China)

    2015-11-15

    Highlights: • A unified modeling approach for physical experiment design is presented. • Any laser facility can be flexibly defined and included with two scripts. • Complex targets and laser beams can be parametrically modeled for optimization. • Automatically mapping of laser beam energy facilitates targets shape optimization. - Abstract: Physical experiment design and optimization is very essential for laser driven inertial confinement fusion due to the high cost of each shot. However, only limited experiments with simple structure or shape on several laser facilities can be designed and evaluated in available codes, and targets are usually defined by programming, which may lead to it difficult for complex shape target design and optimization on arbitrary laser facilities. A unified modeling approach for physical experiment design and optimization on any laser facilities is presented in this paper. Its core idea includes: (1) any laser facility can be flexibly defined and included with two scripts, (2) complex shape targets and laser beams can be parametrically modeled based on features, (3) an automatically mapping scheme of laser beam energy onto discrete mesh elements of targets enable targets or laser beams be optimized without any additional interactive modeling or programming, and (4) significant computation algorithms are additionally presented to efficiently evaluate radiation symmetry on the target. Finally, examples are demonstrated to validate the significance of such unified modeling approach for physical experiments design and optimization in laser driven inertial confinement fusion.

  1. Design study of an accelerator for heavy ion fusion

    International Nuclear Information System (INIS)

    Katayama, T.; Noda, A.; Tokuda, N.; Hirao, Y.

    1980-01-01

    Design of a demonstration accelerator for heavy ion fusion based on a synchrotron system is briefly described. The proposed complex system of injector linac, rapid cycling synchrotron and five accumulation rings can produce a peak current 1.6 kA, peak power 32 TW and total energy 0.3 MJ. Investigations of the intrabeam scattering give a lifetime of the beam longer than the fusion cycle time of 1 sec

  2. Definition and conceptual design of a small fusion reactor

    International Nuclear Information System (INIS)

    1979-04-01

    The objective of this project is to evaluate various mirror fusion reactor concepts that might result in small systems for the effective production of electrical power or stored energy (e.g., nuclear and chemical fuels). The basic two-year program goal is to select a particular concept and develop the conceptual design of a pilot plant that could provide a useful output from fusion. The pilot plant would be built and operated in the late 1980s

  3. Double-shell target designs for the Los Alamos Scientific Laboratory eight-beam laser system

    International Nuclear Information System (INIS)

    Kindel, J.M.; Stroscio, M.A.

    1978-03-01

    We investigate two double-pusher laser fusion targets, one that incorporates an outer exploding pusher shell and another that uses velocity multiplication. Specific designs are presented for the Los Alamos Scientific Laboratory Eight-Beam Laser System

  4. Design of neutron streak camera for fusion diagnostics

    International Nuclear Information System (INIS)

    Wang, C.L.; Kalibjian, R.; Singh, M.S.

    1982-06-01

    The D-T burn time for advanced laser-fusion targets is calculated to be very short, 2 . Each fission fragment leaving the cathode generates 400 secondary electrons that are all < 20 eV. These electrons are focussed to a point with an extractor and an anode, and are then purified with an electrostatic deflector. The electron beam is streaked and detected with the standard streak camera techniques. Careful shielding is needed for x-rays from the fusion target and general background. It appears that the neutron streak camera can be a viable and unique tool for studying temporal history of fusion burns in D-T plasmas of a few keV ion temperature

  5. Design aspects of low activation fusion ignition experiments

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. Investigation of IFMIF target assembly structure design

    International Nuclear Information System (INIS)

    Ida, Mizuho; Nakamura, Hiroo; Sugimoto, Masayoshi; Yamamura, Toshio

    2006-10-01

    In the International Fusion Materials Irradiation Facility (IFMIF), the back-wall of target assembly is the part suffered the highest neutron-flux. The back-wall and the assembly are designed to have lips for cutting/welding at the back-wall replacement. To reduce thermal stress and deformation of the back-wall under neutron irradiation, contact pressure between the back-wall and the assembly is one of dominant factors. Therefore, an investigation was performed for feasible clamping pressure of a mechanical clamp set in limited space around the back-wall. It was clarified that the clamp can give a pressure difference up to 0.4 MPa between the contact pressure and atmosphere pressure in the test cell room. Also a research was performed for the dissimilar metal welding in the back-wall. Use of 309 steel was found adequate as the intermediate filler metal through the research of previous welding. Maintaining a temperature of the target assembly so as to avoid a freezing of liquid lithium is needed at the lithium charge into the loop before the beam injection. The assembly is covered with thermal insulation. Therefore, a research and an investigation were performed for compact and light thermal-insulation effective even under helium (i.e. high heat-conduction) condition of the test cell room. The result was as follows; in the case that a thermal conductivity 0.008 W/m·K of one of found insulation materials is available in the temperature range up to 300degC of the IFMIF target assembly, needed thickness and weight of the insulation were respectively only 8.2 mm and 32 kg. Also a research was performed for high-heat-density heaters to maintain temperature of the back-wall which can not be cover with insulation due to limited space. A heater made of silicon-nitride was found to be adequate. Total heat of 8.4 kW on the back-wall was found to be achievable through an investigations of heater arrange. Also an investigation was performed for remote-handling device to

  7. International Fusion Materials Irradiation Facility conceptual design activity. Present status and perspective

    International Nuclear Information System (INIS)

    Kondo, Tatsuo; Noda, Kenji; Oyama, Yukio

    1998-01-01

    For developing the materials for nuclear fusion reactors, it is indispensable to study on the neutron irradiation behavior under fusion reactor conditions, but there is not any high energy neutron irradiation facility that can simulate fusion reactor conditions at present. Therefore, the investigation of the IFMIF was begun jointly by Japan, USA, Europe and Russia following the initiative of IEA. The conceptual design activities were completed in 1997. As to the background and the course, the present status of the research on heavy irradiation and the testing means for fusion materials, the requirement and the technical basis of high energy neutron irradiation, and the international joint design activities are reported. The materials for fusion reactors are exposed to the neutron irradiation with the energy spectra up to 14 MeV. The requirements from the users that the IFMIF should satisfy, the demand of the tests for the materials of prototype and demonstration fusion reactors and the evaluation of the neutron field characteristics of the IFMIF are discussed. As to the conceptual design of the IFMIF, the whole constitution, the operational mode, accelerator system and target system are described. (K.I.)

  8. Radiological safety design considerations for a laser-fusion facility

    International Nuclear Information System (INIS)

    Singh, M.S.

    1977-01-01

    Detailed neutronics and photonics calculations have been performed for analyzing prompt and residual radiations and required shielding associated with the design of a laser-fusion facility with a nominal yield of 10 19 neutrons per D--T burn pulse. The standard Livermore Monte Carlo codes and nuclear data cross section libraries were used in calculations. The Bateman equation was used to calculate the accumulation and decay of radionuclide chain products. A number of activation sensitivity experiments were conducted and the results were found to be in very good agreement within 10 percent of those calculated. It has been found that neutron yields of 2 x 10 19 per day can be conducted continuously if the reactor chamber is Kevlar-epoxy or silica, the primary shield is 0.60-m of water immediately on the chamber, and the building concrete is 1.80 m thick. These precautions result in dose equivalents below the primary protection limits inside the target room after a few hours of cool-down per each 10 19 pulse, 10 percent of the primary protection limits immediately outside the target room, and 1 percent of the natural background level at the nearest site boundary

  9. Repetitive laser fusion experiment and operation using a target injection system

    International Nuclear Information System (INIS)

    Nishimura, Yasuhiko; Komeda, Osamu; Mori, Yoshitaka

    2017-01-01

    Since 2008, a collaborative research project on laser fusion development based on a high-speed ignition method using repetitive laser has been carried out with several collaborative research institutes. This paper reports the current state of operation of high repetition laser fusion experiments, such as target introduction and control based on a target injection system that allows free falling under 1 Hz, using a high repetition laser driver that has been under research and development, as well as the measurement of targets that freely fall. The HAMA laser driver that enabled high repetition fusion experiments is a titanium sapphire laser using a diode-pumped solid-state laser KURE-I of green light output as a driver pump light source. In order to carry out high repetition laser fusion experiments, the target injection device allows free falling of deuterated polystyrene solid sphere targets of 1 mm in diameter under 1 Hz. The authors integrated the developed laser and injection system, and succeeded first in the world in making the nuclear fusion reaction continuously by hitting the target to be injected with laser, which is essential technology for future laser nuclear fusion reactor. In order to realize repetition laser fusion experiments, stable laser, target synchronization control, and target position measurement technologies are indispensable. (A.O.)

  10. Conceptual design study for a laser fusion hybrid

    International Nuclear Information System (INIS)

    Maniscalco, J.A.

    1976-01-01

    Lawrence Livermore Laboratory and Bechtel Corporation have been involved in a joint effort to conceptually design a laser fusion hybrid reactor. The design which has evolved is a depleted-uranium fueled fast-fission blanket which produces fissile plutonium and electricity. A major objective of the design study was to evaluate the feasibility of producing fissile fuel with laser fusion. This feasibility evaluation was carried out by analyzing the integrated engineering performance of the complete conceptual design and by identifying the required laser/pellet performance. The performance of the laser fusion hybrid has also been compared to a typical fast breeder reactor. The results show that the laser fusion hybrid produces enough fissile material to fuel more than six light water reactors (LWRs) of equivalent thermal power while operating in a regime which requires an order of magnitude less laser and pellet performance than pure laser fusion. In comparison to a fast breeder reactor the hybrid produces 10 times more fissile fuel. An economic analysis of the design shows that the cost of electricity in a combined hybrid-LWR scenario increases by only 20 to 40 percent when the capital cost of the hybrid ranges from 2 to 3 times more than an LWR

  11. Conceptual design study for a laser fusion hybrid

    International Nuclear Information System (INIS)

    Maniscalco, J.A.

    1976-09-01

    Lawrence Livermore Laboratory and Bechtel Corporation have been involved in a joint effort to conceptually design a laser fusion hybrid reactor. The design which has evolved is a depleted-uranium fueled fast-fission blanket which produces fissile plutonium and electricity. A major objective of the design study was to evaluate the feasibility of producing fissile fuel with laser fusion. This feasibility evaluation was carried out by analyzing the integrated engineering performance of the complete conceptual design and by identifying the required laser/pellet performance. The performance of the laser fusion hybrid has also been compared to a typical fast breeder reactor. The results show that the laser fusion hybrid produces enough fissile material to fuel more than six light water reactors (LWR's) of equivalent thermal power while operating in a regime which requires an order of magnitude less laser and pellet performance than pure laser fusion. In comparison to a fast breeder reactor the hybrid produces 10 times more fissile fuel. An economic analysis of the design shows that the cost of electricity in a combined hybrid-LWR scenario is insensitive to the capital cost of the hybrid, increasing by only 20 to 40 percent when the capital cost of the hybrid ranges from 2 to 3 times more than an LWR

  12. Osiris and SOMBRERO inertial fusion power plant designs - summary, conclusions, and recommendations

    International Nuclear Information System (INIS)

    Meier, Wayne R.

    1994-01-01

    An 18 month study to evaluate the potential of inertial fusion energy (IFE) for electric power production has been completed. The primary objective of the study was to provide the US Department of Energy with an evaluation of the potential of inertial fusion for electric power production. The study included the conceptual design of two inertial fusion power plants. Osiris uses an induction linac heavy ion beam driver, and SOMBRERO uses a krypton fluoride laser driver. Conceptual designs were completed for the reactors, power conversion and plant facilities, and drivers. Environmental and safety aspects, technical issues, technology development needs, and economics of the final point designs were assessed and compared. This paper summarizes the results and conclusions of the conceptual designs and results of the assessment studies. We conclude that IFE has the potential of producing technically credible designs with environmental, safety, and economics characteristics that are just as attractive as magnetic fusion. Realizing this potential will require additional research and development on target physics, chamber design, target production and injection systems, and drivers. ((orig.))

  13. Research and Application of Autodesk Fusion360 in Industrial Design

    Science.gov (United States)

    Song, P. P.; Qi, Y. M.; Cai, D. C.

    2018-05-01

    In 2016, Fusion 360, a productintroduced byAutodesk and integrating industrial design, structural design, mechanical simulation, and CAM, turns out a design platform supportingcollaboration and sharing both cross-platform and via the cloud. In previous products, design and manufacturing use to be isolated. In the course of design, research and development, the communication between designers and engineers used to go on through different software products, tool commands, and even industry terms. Moreover, difficulty also lies with the communication between design thoughts and machining strategies. Naturally, a difficult product design and R & D process would trigger a noticeable gap between the design model and the actual product. A complete product development process tends to cover several major areas, such as industrial design, mechanical design, rendering and animation, computer aided emulation (CAE), and computer aided manufacturing (CAM). Fusion 360, a perfect design solving the technical problems of cross-platform data exchange, realizes the effective control of cross-regional collaboration and presents an overview of collaboration and breaks the barriers between art and manufacturing, andblocks between design and processing. The “Eco-development of Fusion360 Industrial Chain” is both a significant means to and an inevitable trend forthe manufacturers and industrial designers to carry out innovation in China.

  14. FY-2013 FES (Fusion Energy Sciences) Joint Research Target Report

    Energy Technology Data Exchange (ETDEWEB)

    Fenstermacher, M. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Garofalo, A. M. [General Atomics, San Diego, CA (United States); Gerhardt, S. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Hubbard, A. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Maingi, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Whyte, D. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    The H-mode confinement regime is characterized by a region of good thermal and particle confinement at the edge of the confined plasma, and has generally been envisioned as the operating regime for ITER and other next step devices. This good confinement is often interrupted, however, by edge-localized instabilities, known as ELMs. On the one hand, these ELMs provide particle and impurity flushing from the plasma core, a beneficial effect facilitating density control and stationary operation. On the other hand, the ELMs result in a substantial fraction of the edge stored energy flowing in bursts to the divertor and first wall; this impulsive thermal loading would result in unacceptable erosion of these material surfaces if it is not arrested. Hence, developing and understanding operating regimes that have the energy confinement of standard H-mode and the stationarity that is provided by ELMs, while at the same time eliminating the impulsive thermal loading of large ELMs, is the focus of the 2013 FES Joint Research Target (JRT): Annual Target: Conduct experiments and analysis on major fusion facilities, to evaluate stationary enhanced confinement regimes without large Edge Localized Modes (ELMs), and to improve understanding of the underlying physical mechanisms that allow acceptable edge particle transport while maintaining a strong thermal transport barrier. Mechanisms to be investigated can include intrinsic continuous edge plasma modes and externally applied 3D fields. Candidate regimes and techniques have been pioneered by each of the three major US facilities (C-Mod, D3D and NSTX). Coordinated experiments, measurements, and analysis will be carried out to assess and understand the operational space for the regimes. Exploiting the complementary parameters and tools of the devices, joint teams will aim to more closely approach key dimensionless parameters of ITER, and to identify correlations between edge fluctuations and transport. The role of rotation will be

  15. Knudsen and inverse Knudsen layer effect on tail ion distribution and fusion reactivity in inertial confinement fusion targets

    Science.gov (United States)

    McDevitt, C. J.; Tang, X.-Z.; Guo, Z.; Berk, H. L.

    2014-10-01

    A series of reduced models are used to study the fast ion tail in the vicinity of a transition layer between plasmas at disparate temperatures and densities, which is typical of the gas-pusher interface in inertial confinement fusion targets. Emphasis is placed on utilizing progressively more comprehensive models in order to identify the essential physics for computing the fast ion tail at energies comparable to the Gamow peak. The resulting fast ion tail distribution is subsequently used to compute the fusion reactivity as a function of collisionality and temperature. It is found that while the fast ion distribution can be significantly depleted in the hot spot, leading to a reduction of the fusion reactivity in this region, a surplus of fast ions is present in the neighboring cold region. The presence of this fast ion surplus in the neighboring cold region is shown to lead to a partial recovery of the fusion yield lost in the hot spot.

  16. Safety issues relating to the design of fusion power facilities

    International Nuclear Information System (INIS)

    Stasko, R.R.; Wong, K.Y.; Russell, S.B.

    1986-06-01

    In order to make fusion power a viable future source of energy, it will be necessary to ensure that the cost of power for fusion electric generation is competitive with advanced fission concepts. In addition, fusion power will have to live up to its original promise of being a more radiologically benign technology than fission, and be able to demonstrate excellent operational safety performance. These two requirements are interrelated, since the selection of an appropriate safety philosophy early in the design phase could greatly reduce or eliminate the capital costs of elaborate safety related and protective sytems. This paper will briefly overview a few of the key safety issues presently recognized as critical to the ultimate achievement of licensable, environmentally safe and socially acceptable fusion power facilities. 12 refs

  17. ELMO Bumpy Torus fusion-reactor design study

    International Nuclear Information System (INIS)

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

    1981-01-01

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is described that emphasizes those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are generic to magnetic fusion being adopted from past, more extensive tokamak reactor designs

  18. HIV-1 gp41 Fusion Intermediate: A Target for HIV Therapeutics

    Directory of Open Access Journals (Sweden)

    Chungen Pan

    2010-02-01

    Full Text Available Human immunodeficiency virus (HIV-1 infection is initiated by the binding of gp120 envelope glyco-protein to its cell receptor (CD4 and a coreceptor (CXCR4 or CCR5, followed by a series of conformational changes in the gp41 transmembrane subunit. These changes include insertion of fusion peptide into the target cell membrane and association of C-heptad repeat (CHR peptide with the N-heptad repeat (NHR trimer, a pre-hairpin fusion intermediate. A stable six-helix bundle core is then formed, bringing the viral envelope and target cell membrane into close proximity for fusion. Peptides derived from the CHR region, such as T20 and C34, inhibit HIV-1 fusion by interacting with the gp41 fusion intermediate. A number of anti-HIV-1 peptides and small molecule compounds targeting the gp41 NHR-trimer have been identified. By combining HIV fusion/entry inhibitors targeting different sites in the gp41 fusion intermediate, a potent synergistic effect takes place, resulting in a potential new therapeutic strategy for the HIV infection/AIDS. Here, we present an overview of the current development of anti-HIV drugs, particularly those targeting the gp41 fusion intermediate.

  19. Challenges of designing fusion reactors for remote maintainability

    International Nuclear Information System (INIS)

    Mason, L.S.

    1981-01-01

    One of the major problems faced by the fusion community is the development of the high level of reliability required to assure that fusion will be a viable commercial power source. Much of the responsibility for solving this problem falls directly on the designer in developing concepts that have a high level of maintainability. The problems are both near-term, in developing maintainability for next generation engineering oriented reactors; and long range, in developing full maintainability for the more commercial concepts with their required high level of on-line time. The near-time challenge will include development of unqiue design concepts to perform inspection, maintenance, replacement, and testing under the stringent conditions imposed by the next generation engineering oriented machines. The long range challenge will focus on basic design concepts that will enable the full mainatability required by commerical fusion

  20. Conceptual design study of a scyllac fusion test reactor

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements

  1. Challenges of designing fusion reactors for remote maintainability

    International Nuclear Information System (INIS)

    Masson, L.S.

    1981-01-01

    One of the major problems faced by the fusion community is the development of the high level of reliability required to assure that fusion will be a viable commercial power source. Much of the responsibility for solving this problem falls directly on the designer in developing concepts that have a high level of maintainability for the next generation engineering oriented reactors; and long range, in developing full maintainability for the more complicated commercial concepts with their required high level of on-line time. The near-term challenge will include development of unique design concepts to perform inspection, maintenance, replacement, and testing under the stringent conditions imposed by the next generation engineering oriented machines. The long range challenge will focus on basic design concepts that will enable the full maintainability required by commercial fusion. In addition to the purely technical challenges, the fusion community is also faced with the problem of developing programmatic means to assure that reactor maintenance issues are given proper and timely emphasis as the nuclear phase of fusion is approached

  2. Production of direct drive cylindrical targets for inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Elliott, N.E.; Day, R.D.; Hatch, D.J.; Sandoval, D.L.; Gomez, V.M.; Pierce, T.H.; Elliott, J.E.; Manzanares, R.

    2002-01-01

    We have made targets with cylindrical geometry for Inertial Confinement Fusion (ICF) experiments. These targets are used in hydrodynamic experiments on the OMEGA laser at the University of Rochester. The cylindrical design allows the study of three dimensional hydrodynamic effects in a pseudo 2D mode, simplifying data gathering and analysis. Direct drive refers to the fact that the target is illuminated directly by approximately 50 laser beams and is imploded by the material pressure generated from ablation of the outside of the target. The production of cylindrical targets involves numerous steps. These steps are shared in common with many other types of ICF targets but no other single target type encompasses such a wide range of fabrication techniques. These targets consist of a large number of individual parts, all fabricated from commercially purchased raw material, requiring many machining, assembly, electroplating and chemical process steps. Virtually every manufacturing and assembly process we currently possess is involved in the production of these targets. The generic target consists of a plastic cylinder (ablator) that is roughly lmm in diameter by 2.25mm long. The wall of the cylinder is roughly 0.07mm thick. There is an aluminum cylinder 0.5mm wide and O.Olmm thick centered on the inside of the plastic cylinder and coaxial with the outside plastic cylinder. The outside of this aluminum band has surface finishes of differing random average roughness. The required average surface roughness is determined in advance by experimental design based on the amount of turbulent mix to be observed. The interior of the cylinder is filled with low density polystyrene foam that is made in house. To produce a finished target additional features are added to each target. X-ray backlighters are cantilevered off the target that allow time resolved x-ray images of the imploding target to be recorded during the experiment. The x-ray backlighters are driven by additional

  3. Early and late HIV-1 membrane fusion events are impaired by sphinganine lipidated peptides that target the fusion site.

    Science.gov (United States)

    Klug, Yoel A; Ashkenazi, Avraham; Viard, Mathias; Porat, Ziv; Blumenthal, Robert; Shai, Yechiel

    2014-07-15

    Lipid-conjugated peptides have advanced the understanding of membrane protein functions and the roles of lipids in the membrane milieu. These lipopeptides modulate various biological systems such as viral fusion. A single function has been suggested for the lipid, binding to the membrane and thus elevating the local concentration of the peptide at the target site. In the present paper, we challenged this argument by exploring in-depth the antiviral mechanism of lipopeptides, which comprise sphinganine, the lipid backbone of DHSM (dihydrosphingomyelin), and an HIV-1 envelope-derived peptide. Surprisingly, we discovered a partnership between the lipid and the peptide that impaired early membrane fusion events by reducing CD4 receptor lateral diffusion and HIV-1 fusion peptide-mediated lipid mixing. Moreover, only the joint function of sphinganine and its conjugate peptide disrupted HIV-1 fusion protein assembly and folding at the later fusion steps. Via imaging techniques we revealed for the first time the direct localization of these lipopeptides to the virus-cell and cell-cell contact sites. Overall, the findings of the present study may suggest lipid-protein interactions in various biological systems and may help uncover a role for elevated DHSM in HIV-1 and its target cell membranes.

  4. Conceptual design of the SlimCS fusion DEMO reactor

    International Nuclear Information System (INIS)

    Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Utoh, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; Sakurai, Shinji; Kurita, Genichi; Hayashi, Takao; Oyama, Naoyuki; Liu Changle; Hamamatsu, Kiyotaka; Inoue, Takashi; Ozeki, Takahisa; Sato, Masayasu; Suzuki, Satoshi; Kawashima, Hisato; Ezato, Koichiro; Tsuru, Daigo; Koizumi, Norikiyo; Sakamoto, Keiji; Ando, Masami; Sakamoto, Yoshiteru; Shibama, Yusuke; Suzuki, Takahiro; Takechi, Manabu; Takahashi, Koji; Hirose, Takanori; Sato, Satoru; Nozawa, Takashi; Tanigawa, Hisashi; Kakudate, Satoshi; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Ochiai, Kentaro; Ide, Shunsuke; Aiba, Nobuyuki; Shimizu, Katsuhiro; Honda, Mitsuru; Nakamichi, Masaru; Nishi, Hiroshi; Seki, Yoji; Nakamura, Yukiharu; Tsuchiya, Kunihiko; Yoshida, Tohru; Song Yuntao

    2010-08-01

    This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). Owing to low aspect ratio, the reactor will be capable of having comparatively high beta limit and high elongation (which can elevate the Greenwald density limit), having potential for high power density. The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m 2 . This report covers various aspects of design study including systematic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept. (author)

  5. Storage and Containment of Nuclear Targets for Pulsed Fission-Fusion Testing

    Data.gov (United States)

    National Aeronautics and Space Administration — The combined fission-fusion fuel target is the heart of an engine concept that can open the solar system to fast and efficient human exploration. This is a unique...

  6. High gain direct drive target designs and supporting experiments with KrF

    International Nuclear Information System (INIS)

    Karasik, Max; Bates, Jason W.; Aglitskiy, Yefim

    2013-01-01

    Krypton-fluoride laser is an attractive inertial fusion energy driver from the standpoint of target physics. Target designs taking advantage of zooming, shock ignition, and favorable physics with KrF reach energy gains of 200 with sub-MJ laser energy. The designs are robust under 2D simulations. Experiments on the Nike KrF laser support the physics basis. (author)

  7. Numerical studies of deuterium-tritium ignition in impact-fusion targets

    International Nuclear Information System (INIS)

    Zubrin, R.M.; Ribe, F.L.

    1989-01-01

    A numerical one-dimensional solution of the Euler equations for an imploding spherical tungsten shell with internal deuterium-tritium gas is applied to study impact-fusion dynamics with parameters of fusion reactor relevance. Thermal conduction and radiative energy loss by the plasma are taken into account, as is heating by fusion generated alpha particles. A variety of target sizes and impact velocities are examined, and scaling laws for fusion yields are deduced which define possible parameters for conceptual commercial impact-fusion power reactors. It is found that shell energies and velocities of about 30 MJ and 110 km/s would be satisfactory. A commercial impact-fusion reactor based on such parameters is discussed

  8. Introduction of fusion driven subcritical system plasma design

    International Nuclear Information System (INIS)

    Bin Wu

    2003-01-01

    Fusion driven subcritical nuclear system (FDS) is a multifunctional hybrid reactor, which could breed nuclear fuel, transmute long-lived wastes, producing tritium and so on. This paper presents an introduction of FDS plasma design. Several different advance equilibrium configurations have been proposed and a 1.5-D discharge simulation of FDS was also present

  9. Conceptual design of a laser fusion power plant. Part I. An integrated facility

    International Nuclear Information System (INIS)

    1981-07-01

    This study is a new preliminary conceptual design and economic analysis of an inertial confinement fusion (ICF) power plant performed by Bechtel under the direction of Lawrence Livermore National Laboratory (LLNL). The purpose of a new conceptual design is to examine alternatives to the LLNL HYLIFE power plant and to incorporate information from the recent liquid metal cooled power plant conceptual design study (CDS) into the reactor system and balance of plant design. A key issue in the design of a laser fusion power plant is the degree of symmetry in the illumination of the target that will be required for a proper burn. Because this matter is expected to remain unresolved for some time, another purpose of this study is to determine the effect of symmetry requirements on the total plant size, layout, and cost

  10. Maximum entropy restoration of laser fusion target x-ray photographs

    International Nuclear Information System (INIS)

    Brolley, J.E.; Lazarus, R.B.; Suydam, B.R.

    1976-01-01

    Maximum entropy principles were used to analyze the microdensitometer traces of a laser-fusion target photograph. The object is a glowing laser-fusion target microsphere 0.95 cm from a pinhole of radius 2 x 10 -4 cm, the image is 7.2 cm from the pinhole and the photon wavelength is likely to be 6.2 x 10 -8 cm. Some computational aspects of the problem are also considered

  11. Reactor design considerations for inertial confinement fusion

    International Nuclear Information System (INIS)

    Booth, L.A.

    1979-01-01

    The most challenging reactor design consideration is protection of the cavity wall from the various energy forms as released by the pellet and as affected by the reaction-chamber phenomena. These phenomena depend on both the design and the yield of the pellet, as well as on ambient conditions in the chamber at the time of the pellet microexplosion. The effects on pellet energy-release mechanisms of various reaction chamber atmosphere options are summarized

  12. Electrostatic levitation, control and transport in high rate, low cost production of inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Hendricks, C.D.; Johnson, W.L.

    1979-01-01

    Inertial confinement fusion requires production of power plant grade targets at high rates and process yield. A review of present project specifications and techniques to produce targets is discussed with special emphasis on automating the processes and combining them with an electrostatic transport and suspension system through the power plant target factory

  13. Multiple Target Laser Designator (MTLD)

    Science.gov (United States)

    2007-03-01

    Optimized Liquid Crystal Scanning Element Optimize the Nonimaging Predictive Algorithm for Target Ranging, Tracking, and Position Estimation...commercial potential. 3.0 PROGRESS THIS QUARTER 3.1 Optimization of Nonimaging Holographic Antenna for Target Tracking and Position Estimation (Task 6) In

  14. Present status of the conceptual design of IFMIF target facility

    International Nuclear Information System (INIS)

    Katsuta, H.; Kato, Y.; Konishi, S.; Miyauchi, Y.; Smith, D.; Hua, T.; Green, L.; Benamati, G.; Cevolani, S.; Roehrig, H.; Schutz, W.

    1998-01-01

    The conceptual design activity (CDA) for the international fusion materials irradiation facility (IFMIF) has been conducted. For the IFMIF target facility, the conceptual designs of the following two main components have been performed. The design concept of IFMIF utilizes a high energy deuteron beam of 30-40 MeV and total current of 250 mA, impinging on a flowing lithium jet to produce high energy neutrons for irradiation of candidate fusion materials. (1) The target assembly: The kinetic energy of the deuteron beam is deposited on a Li-jet target and neutrons are produced through the d-Li stripping reaction in this target. The assembly is designed to get a stable lithium jet and to prevent the onset of lithium boiling. For 40-MeV deuteron beam (total current of 250 mA) and a beam footprint of 5 x 20 cm 2 lithium jet dimensions are designed to be 2.5 cm thick and 26 cm wide. The lithium jet parameters are given. (2) Lithium loop: The loop circulates the lithium to and from the target assembly and removes the heat deposited by the deuteron beam containing systems for maintaining the-high purity of the lithium required for radiological safety and to minimize corrosion. The maximum lithium flow rate is 130 l/s and the total lithium inventory is about 21 m 3 . The IFMIF policy requires that the lithium loop system be designed to guarantee no combustion of lithium in the event of a lithium leak. This can be achieved by use of multiple confinement of the lithium carrying components. The radioactive waste generated by the target facilities is estimated. (orig.)

  15. Biochemistry and biophysics of HIV-1 gp41 - membrane interactions and implications for HIV-1 envelope protein mediated viral-cell fusion and fusion inhibitor design.

    Science.gov (United States)

    Cai, Lifeng; Gochin, Miriam; Liu, Keliang

    2011-12-01

    Human immunodeficiency virus type 1 (HIV-1), the pathogen of acquired immunodeficiency syndrome (AIDS), causes ~2 millions death every year and still defies an effective vaccine. HIV-1 infects host cells through envelope protein - mediated virus-cell fusion. The transmembrane subunit of envelope protein, gp41, is the molecular machinery which facilitates fusion. Its ectodomain contains several distinguishing functional domains, fusion peptide (FP), Nterminal heptad repeat (NHR), C-terminal heptad repeat (CHR) and membrane proximal extracellular region (MPER). During the fusion process, FP inserts into the host cell membrane, and an extended gp41 prehairpin conformation bridges the viral and cell membranes through MPER and FP respectively. Subsequent conformational change of the unstable prehairpin results in a coiled-coil 6-helix bundle (6HB) structure formed between NHR and CHR. The energetics of 6HB formation drives membrane apposition and fusion. Drugs targeting gp41 functional domains to prevent 6HB formation inhibit HIV-1 infection. T20 (enfuvirtide, Fuzeon) was approved by the US FDA in 2003 as the first fusion inhibitor. It is a 36-residue peptide from the gp41 CHR, and it inhibits 6HB formation by targeting NHR and lipids. Development of new fusion inhibitors, especially small molecule drugs, is encouraged to overcome the shortcomings of T20 as a peptide drug. Hydrophobic characteristics and membrane association are critical for gp41 function and mechanism of action. Research in gp41-membrane interactions, using peptides corresponding to specific functional domains, or constructs including several interactive domains, are reviewed here to get a better understanding of gp41 mediated virus-cell fusion that can inform or guide the design of new HIV-1 fusion inhibitors.

  16. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

    A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  17. 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)

  18. Osiris and SOMBRERO inertial confinement fusion power plant designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Bieri, R.L.; Monsler, M.J.

    1992-03-01

    Conceptual designs and assessments have been completed for two inertial fusion energy (IFE) electric power plants. The detailed designs and results of the assessment studies are presented in this report. Osiris is a heavy-ion-beam (HIB) driven power plant and SOMBRERO is a Krypton-Fluoride (KrF) laser-driven power plant. Both plants are sized for a net electric power of 1000 MWe

  19. Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

    International Nuclear Information System (INIS)

    Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

    1987-01-01

    The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

  20. Conceptual design of the JAERI demonstration fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Tone, T.; Seki, Y.

    1976-01-01

    Conceptual design of a tokamak demonstration fusion reactor is carried out. This design is an extended and improved version of the previous design which was presented at the 5th IAEA Conference. The main design parameters are as follows: the reactor thermal power 2000 MW, torus radius 10.5 m, plasma radius 2.7 m, first wall radius 3.0 m, toroidal magnetic field on axis 6T, blanket fertile material Li 2 O, coolant He, structural material Mo-alloy and tritium breeding ratio 1.2

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  2. Qualitative Reliability Issues for Solid and Liquid Wall Fusion Design

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles

    2001-01-01

    This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features.

  3. Qualitative Reliability Issues for Solid and Liquid Wall Fusion Designs

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, L.C.

    2001-01-31

    This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features.

  4. Qualitative Reliability Issues for Solid and Liquid Wall Fusion Designs

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    2001-01-01

    This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features

  5. MRI/TRUS fusion software-based targeted biopsy: the new standard of care?

    Science.gov (United States)

    Manfredi, M; Costa Moretti, T B; Emberton, M; Villers, A; Valerio, M

    2015-09-01

    The advent of multiparametric MRI has made it possible to change the way in which prostate biopsy is done, allowing to direct biopsies to suspicious lesions rather than randomly. The subject of this review relates to a computer-assisted strategy, the MRI/US fusion software-based targeted biopsy, and to its performance compared to the other sampling methods. Different devices with different methods to register MR images to live TRUS are currently in use to allow software-based targeted biopsy. Main clinical indications of MRI/US fusion software-based targeted biopsy are re-biopsy in men with persistent suspicious of prostate cancer after first negative standard biopsy and the follow-up of patients under active surveillance. Some studies have compared MRI/US fusion software-based targeted versus standard biopsy. In men at risk with MRI-suspicious lesion, targeted biopsy consistently detects more men with clinically significant disease as compared to standard biopsy; some studies have also shown decreased detection of insignificant disease. Only two studies directly compared MRI/US fusion software-based targeted biopsy with MRI/US fusion visual targeted biopsy, and the diagnostic ability seems to be in favor of the software approach. To date, no study comparing software-based targeted biopsy against in-bore MRI biopsy is available. The new software-based targeted approach seems to have the characteristics to be added in the standard pathway for achieving accurate risk stratification. Once reproducibility and cost-effectiveness will be verified, the actual issue will be to determine whether MRI/TRUS fusion software-based targeted biopsy represents anadd-on test or a replacement to standard TRUS biopsy.

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

    International Nuclear Information System (INIS)

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

    2008-10-01

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

  7. Robust Ground Target Detection by SAR and IR Sensor Fusion Using Adaboost-Based Feature Selection

    Directory of Open Access Journals (Sweden)

    Sungho Kim

    2016-07-01

    Full Text Available Long-range ground targets are difficult to detect in a noisy cluttered environment using either synthetic aperture radar (SAR images or infrared (IR images. SAR-based detectors can provide a high detection rate with a high false alarm rate to background scatter noise. IR-based approaches can detect hot targets but are affected strongly by the weather conditions. This paper proposes a novel target detection method by decision-level SAR and IR fusion using an Adaboost-based machine learning scheme to achieve a high detection rate and low false alarm rate. The proposed method consists of individual detection, registration, and fusion architecture. This paper presents a single framework of a SAR and IR target detection method using modified Boolean map visual theory (modBMVT and feature-selection based fusion. Previous methods applied different algorithms to detect SAR and IR targets because of the different physical image characteristics. One method that is optimized for IR target detection produces unsuccessful results in SAR target detection. This study examined the image characteristics and proposed a unified SAR and IR target detection method by inserting a median local average filter (MLAF, pre-filter and an asymmetric morphological closing filter (AMCF, post-filter into the BMVT. The original BMVT was optimized to detect small infrared targets. The proposed modBMVT can remove the thermal and scatter noise by the MLAF and detect extended targets by attaching the AMCF after the BMVT. Heterogeneous SAR and IR images were registered automatically using the proposed RANdom SAmple Region Consensus (RANSARC-based homography optimization after a brute-force correspondence search using the detected target centers and regions. The final targets were detected by feature-selection based sensor fusion using Adaboost. The proposed method showed good SAR and IR target detection performance through feature selection-based decision fusion on a synthetic

  8. Robust Ground Target Detection by SAR and IR Sensor Fusion Using Adaboost-Based Feature Selection

    Science.gov (United States)

    Kim, Sungho; Song, Woo-Jin; Kim, So-Hyun

    2016-01-01

    Long-range ground targets are difficult to detect in a noisy cluttered environment using either synthetic aperture radar (SAR) images or infrared (IR) images. SAR-based detectors can provide a high detection rate with a high false alarm rate to background scatter noise. IR-based approaches can detect hot targets but are affected strongly by the weather conditions. This paper proposes a novel target detection method by decision-level SAR and IR fusion using an Adaboost-based machine learning scheme to achieve a high detection rate and low false alarm rate. The proposed method consists of individual detection, registration, and fusion architecture. This paper presents a single framework of a SAR and IR target detection method using modified Boolean map visual theory (modBMVT) and feature-selection based fusion. Previous methods applied different algorithms to detect SAR and IR targets because of the different physical image characteristics. One method that is optimized for IR target detection produces unsuccessful results in SAR target detection. This study examined the image characteristics and proposed a unified SAR and IR target detection method by inserting a median local average filter (MLAF, pre-filter) and an asymmetric morphological closing filter (AMCF, post-filter) into the BMVT. The original BMVT was optimized to detect small infrared targets. The proposed modBMVT can remove the thermal and scatter noise by the MLAF and detect extended targets by attaching the AMCF after the BMVT. Heterogeneous SAR and IR images were registered automatically using the proposed RANdom SAmple Region Consensus (RANSARC)-based homography optimization after a brute-force correspondence search using the detected target centers and regions. The final targets were detected by feature-selection based sensor fusion using Adaboost. The proposed method showed good SAR and IR target detection performance through feature selection-based decision fusion on a synthetic database generated

  9. Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2011-01-01

    Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

  10. Tritium management in fusion synfuel designs

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1980-01-01

    Two blanket types are being studied: a lithium-sodium pool boiler and a lithium-oxide- or lithium-sodium pool boiler and a lithium-oxide- or aluminate-microsphere moving bed. For each, a wide variety of current technology was considered in handling the tritium. Here, we show the pool boiler with the sulfur-iodine thermochemical cycle first developed and now being piloted by the General Atomic Company. The tritium (T 2 ) will be generated in the lithium-sodium mixture where the concentration is approx. 10 ppM and held constant by a scavenging system consisting mainly of permeators. An intermediate sodium loop carries the blanket heat to the thermochemical cycle, and the T 2 in this loop is held to 1 ppM by a similar scavenging system. With this design, we have maintained blanket inventory at 1 kg of tritium, kept thermochemical cycle losses to 5 Ci/d and environmental loss to 10 Ci/d, and held total plant risk inventory at 7 kg tritium

  11. Substantial reductions of input energy and peak power requirements in targets for heavy ion fusion

    International Nuclear Information System (INIS)

    Mark, J.W.K.; Pan, Y.L.

    1986-01-01

    Two ways of reducing the requirements of the heavy ion driver for inertial confinement fusion (ICF) target implosion are described. Compared to estimates of target gain not using these methods, the target input energy and peak power may be reduced by about a factor of two with the use of the hybrid-implosion concept. Another factor of two reduction in input energy may be obtained with the use of spin-polarized DT fuel in the ICF target

  12. Fusion reactor design and technology program in China

    International Nuclear Information System (INIS)

    Huang, J.H.

    1994-01-01

    A fusion-fission hybrid reactor program was launched in 1987. The purpose of development of the hybrid reactor is twofold: to solve the problem of nuclear fuel supply for an expected large-scale development of fission reactor plants, and to maintain the momentum of fusion research. The program is described and the activities and progress of the program are presented. Two conceptual designs of an engineering test reactor with tokamak configuration were developed at the Southwestern Institute of Physics and the Institute of Plasma Physics. The results are a tokamak engineering test breeder (TETB) series design and a fusion-fission hybrid reactor design (SSEHR), characterized by a liquid-Li self-cooled blanket and an He-cooled solid tritium breeder blanket respectively. In parallel with the design studies, relevant technological experiments on a small or medium scale have been supported by this program. These include LHCD, ICRH and pellet injection in the area of plasma engineering; neutronics integral experiments with U, Pu, Fe and Be; various irradiation tests of austenitic and ferritic steels, magnetohydrodynamic (MHD) pressure drop experiments using a liquid metal loop; research into permeation barriers for tritium and hydrogen isotopes; solid tritium breeder tests using an in-situ loop in a fission reactor. All these experiments have proceeded successfully. The second step of this program is now starting. It seems reasonable that most of the research carried out in the first step will continue. ((orig.))

  13. LIBRA - a light ion beam inertial confinement fusion reactor conceptual design

    International Nuclear Information System (INIS)

    Moses, G.A.; Kulcinski, G.L.; Bruggink, D.

    1989-01-01

    The LIBRA light ion beam fusion commercial reactor study is a self-consistent conceptual design of a 330 MWe power plant with an accompanying economic analysis. Fusion targets are imploded by 4 MJ shaped pulses of 30 MeV Li ions at a rate of 3 Hz. The target gain is 80, leading to a yield of 320 MJ. The high intensity part of the ion pulse is delivered by 16 diodes through 16 separate z-pinch plasma channels formed in 100 torr of helium with trace amounts of lithium. The blanket is an array of porous flexible silicon carbide tubes with Li 17 Pb 83 flowing downward through them. These tubes (INPORT units) shield the target chamber wall from both neutron damage and the shock overpressure of the target explosion. The target chamber is a right circular cylinder, 8.7 meters in diameter. The target chamber is 'self-pumped' by the target explosion generated overpressure into a surge tank partially filled with liquid that surrounds the target chamber. This scheme refreshes the chamber at the desired 3 Hz frequency without excessive pumping demands. The blanket multiplication is 1.2 and the tritium breeding ratio is 1.4. The direct capital cost of LIBRA is estimated to be $2200/kWe. (author)

  14. Relativistic electron transport in a solid target: study of heating in the framework of inertial fusion

    International Nuclear Information System (INIS)

    Martinolli, E.

    2003-04-01

    This work is dedicated to the study of the energy deposition of fast electrons in matter. This topic is of prime importance for inertial fusion driven by laser since relativistic electrons are produced in laser-matter interaction for a laser operating in ultra-intense regime. This thesis is made up of: a theoretical chapter dealing with the generation and transport of fast electrons, of 2 chapters reporting experimental data obtained with optical and X-rays diagnostics at the laser facilities of LULI in France and RAL in U.K., and of a chapter dedicated to the simulation of electron transport by using a Monte-Carlo code combined to a hybrid collisional-electromagnetic PIC code. A new spectrometer has been designed: the detection of Kα rays coming from a fluorescent layer embedded in the target has allowed us to assess the size of the electron beam and the level of ionisation. (A.C.)

  15. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua

    2003-01-01

    A summary of the tritium system design studies for the engineering outline design of a fusion experimental breeder (FEB-E) is presented. This paper is divided into three sections. In first section, the geometry, loading features and tritium concentrations in liquid lithium of tritium breeding zones of blanket are described. The tritium flow chart corresponding to the tritium fuel cycle system has been constructed, and the inventories in ten subsystems are calculated using SWITRIM code in section 2. Results show that the necessary initial tritium storage to start up FEB-E with fusion power of 143 MW is about 319 g. In final section, the tritium leakage issues under different operation circumstances have been analyzed. It was found that the potential danger of tritium leakage could be resulted from the exhausted gas of the diverter system. It is important to elevate the tritium burnup fraction and reduce the tritium throughput. (authors)

  16. Second preliminary design of JAERI experimental fusion reactor (JXFR)

    International Nuclear Information System (INIS)

    Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

    1979-06-01

    Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

  17. Mechanical design of a magnetic fusion production reactor

    International Nuclear Information System (INIS)

    Neef, W.S.; Jassby, D.L.

    1986-01-01

    The mechanical aspects of a tandem mirror and tokamak concepts for the tritium production mission are compared, and a proposed breeding blanket configuration for each type of reactor is presented in detail, along with a design outline of the complete fusion reaction system. In both cases, the reactor design is developed sufficiently to permit preliminary cost estimates of all components. A qualitative comparison is drawn between both concepts from the view of mechanical design and serviceability, and suggestions are made for technology proof tests on unique mechanical features. Detailed cost breakdowns indicate less than 10% difference in the overall costs of the two reactors

  18. Magnetized target fusion: An ultra high energy approach in an unexplored parameter space

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1994-01-01

    Magnetized target fusion is a concept that may lead to practical fusion applications in a variety of settings. However, the crucial first step is to demonstrate that it works as advertised. Among the possibilities for doing this is an ultrahigh energy approach to magnetized target fusion, one powered by explosive pulsed power generators that have become available for application to thermonuclear fusion research. In a collaborative effort between Los Alamos and the All-Russian Scientific Institute for Experimental Physics (VNIIEF) a very powerful helical generator with explosive power switching has been used to produce an energetic magnetized plasma. Several diagnostics have been fielded to ascertain the properties of this plasma. We are intensively studying the results of the experiments and calculationally analyzing the performance of this experiment

  19. Progress on the reference mirror fusion reactor design

    International Nuclear Information System (INIS)

    Carlson, G.A.; Doggett, J.N.; Moir, R.W.

    1976-01-01

    The design of a reference mirror fusion reactor is underway at Lawrence Livermore Laboratory. The reactor, rated at about 900 MWe, features steady-state operation, an absence of plasma impurity problems, and good accessibility for blanket maintenance. It is concluded that a mirror reactor appears workable, but its dollar/kWe cost will be considerably higher than present-day nuclear costs. The cost would be reduced most markedly by an increase in plasma Q

  20. Design issues for a laboratory high gain fusion facility

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1987-01-01

    In an inertial fusion laboratory high gain facility, experiments will be carried out with up to 1000 MJ of thermonuclear yield. The experiment area of such a facility will include many systems and structures that will have to operate successfully in the difficult environment created by the sudden large energy release. This paper estimates many of the nuclear effects that will occur, discusses the implied design issues and suggests possible solutions so that a useful experimental facility can be built. 4 figs

  1. An Efficient Method for Identifying Gene Fusions by Targeted RNA Sequencing from Fresh Frozen and FFPE Samples.

    Directory of Open Access Journals (Sweden)

    Jonathan A Scolnick

    Full Text Available Fusion genes are known to be key drivers of tumor growth in several types of cancer. Traditionally, detecting fusion genes has been a difficult task based on fluorescent in situ hybridization to detect chromosomal abnormalities. More recently, RNA sequencing has enabled an increased pace of fusion gene identification. However, RNA-Seq is inefficient for the identification of fusion genes due to the high number of sequencing reads needed to detect the small number of fusion transcripts present in cells of interest. Here we describe a method, Single Primer Enrichment Technology (SPET, for targeted RNA sequencing that is customizable to any target genes, is simple to use, and efficiently detects gene fusions. Using SPET to target 5701 exons of 401 known cancer fusion genes for sequencing, we were able to identify known and previously unreported gene fusions from both fresh-frozen and formalin-fixed paraffin-embedded (FFPE tissue RNA in both normal tissue and cancer cells.

  2. Inertial fusion energy target injection, tracking, and beam pointing

    International Nuclear Information System (INIS)

    Petzoldt, R.W.

    1995-01-01

    Several cryogenic targets must be injected each second into a reaction chamber. Required target speed is about 100 m/s. Required accuracy of the driver beams on target is a few hundred micrometers. Fuel strength is calculated to allow acceleration in excess of 10,000 m/s 2 if the fuel temperature is less than 17 K. A 0.1 μm thick dual membrane will allow nearly 2,000 m/s 2 acceleration. Acceleration is gradually increased and decreased over a few membrane oscillation periods (a few ms), to avoid added stress from vibrations which could otherwise cause a factor of two decrease in allowed acceleration. Movable shielding allows multiple targets to be in flight toward the reaction chamber at once while minimizing neutron heating of subsequent targets. The use of multiple injectors is recommended for redundancy which increases availability and allows a higher pulse rate. Gas gun, rail gun, induction accelerator, and electrostatic accelerator target injection devices are studied, and compared. A gas gun is the preferred device for indirect-drive targets due to its simplicity and proven reliability. With the gas gun, the amount of gas required for each target (about 10 to 100 mg) is acceptable. A revolver loading mechanism is recommended with a cam operated poppet valve to control the gas flow. Cutting vents near the muzzle of the gas gun barrel is recommended to improve accuracy and aid gas pumping. If a railgun is used, we recommend an externally applied magnetic field to reduce required current by an order of magnitude. Optical target tracking is recommended. Up/down counters are suggested to predict target arrival time. Target steering is shown to be feasible and would avoid the need to actively point the beams. Calculations show that induced tumble from electrostatically steering the target is not excessive

  3. Review of target studies for heavy ion fusion

    International Nuclear Information System (INIS)

    Lindl, J.D.; Bangerter, R.D.; Mark, J.W.K.; Pan, Y.L.

    1986-01-01

    We present an updated set of gain curves for radiation driven ion beam targets. The improved target performance calculated with nuclear spin polarized fuel will also be discussed. We discuss the conditions required for efficient conversion to x-rays of ion beam energy. These requirements are compared with those obtained for lasers. Recent results on symmetry requirements for direct drive ion beam targets are presented

  4. Fusion neutron generation by high-repetitive target injection

    International Nuclear Information System (INIS)

    Kitagawa, Yoneyoshi

    2015-01-01

    Pellet injection and repetitive laser illumination are key technologies for realizing inertial fusion energy. The Graduate School for the Creation of New Photonics Industries, Hamamatsu Photonics K. K. and Toyota Motor Corporation demonstrate the pellet injection, counter laser beams' engagement and neutron generation. Deuterated polystyrene (CD) bead pellets, after free-falling for a distance of 18 cm at 1 Hz, are successfully engaged by two counter laser beams from a diode-pumped, ultra-intense laser HAMA. The laser energy, pulse duration, wavelength and the intensity are 0.63 J per beam, 104 fs, 811 nm and 4.7 x 10 18 W/cm 2 , respectively. The irradiated pellets produce D (D, n) 3 He-reacted neutrons with a maximum yield of 9.5 x 10 4 /4π sr/shot. A straight channel with 10 μm-diameter is found through the beads. The pellet size is 1 mm. The results indicate potentially useful technologies for the next step in realizing inertial fusion energy. The results are reviewed as well as some oversea activities. (author)

  5. Ignition energy scaling of inertial confinement fusion targets

    International Nuclear Information System (INIS)

    Johner, J.

    1998-01-01

    Scaling of the ignition energy threshold Ε ig with the implosion velocity v im and isentrope parameter α of imploding spherical deuterium-tritium shells is investigated by performing one-dimensional hydrodynamic simulations of the implosion and hot spot formation dynamics. We find that the a and b exponents in the power-law approximation Ε ig ∝ α a v im -b depend crucially on the subset of initial configurations chosen to establish the scaling law. When we generate the initial states in the same way as in the Livermore study [W.K. Levedahl and J. D. Lindl, Nucl. Fusion 37 (1997) 165 ], we recover the same scaling, Ε ig ∝ α 1.7 v im -5.5 . If, however, the initial states are generated by rescaling the parent configuration according to the hydrodynamic similarity laws, we obtain a different scaling, Ε ig ∝ α 3 v im -9 , which is very close to the αv im -10 dependence predicted by the simple isobaric model for assembled fuel states. The latter is more favourable that the Livermore scaling when rescaling the fusion capsules to higher implosion velocities, but requires the peak drive pressure to be increased as P ∝ v im 5 . (authors)

  6. Target technologies for laser inertial confinement fusion: state-of-the-art and future perspective

    International Nuclear Information System (INIS)

    Zhang Lin; Du Kai

    2013-01-01

    Targets are physical base of the laser inertial confinement fusion (ICF) researches. The quality of the targets has extremely important influences on the reliabilities and degree of precision of the ICF experimental results. The characteristics of the ICF targets, such as complexity and microscale, high precision, determine that the target fabrication process must be a system engineering. This paper presents progresses on the fabrication technologies of ICF targets. The existing problem and the future needs of ICF target fabrication technologies are also discussed. (authors)

  7. Fabrication of cryogenic inertial-confinement-fusion targets using target free-fall technique. Report No. 2-82

    International Nuclear Information System (INIS)

    Kim, K.; Murphy, M.J.

    1982-04-01

    Techniques for fabricating cryogenic inertial confinement fusion targets (i.e., spherical shells containing a uniform layer of DT ice) are investigated using target free-fall concept. Detection and characterization of the moving targets are effected by optoelectronic means, of which the principal is an RF ac-interferometer. This interferometer system demonstrates, for the first time, the speed capabilities of the phase-modulation ac-interferometry. New techiques developed for handling, holding, launching, and transporting targets are also described. Results obtained at both room and cryogenic temperatures are presented

  8. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

  9. Magnet Design Considerations for Fusion Nuclear Science Facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kessel, C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States) Fusion Technology Institute; Titus, P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    2016-06-01

    The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5 T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb3Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.

  10. New design of cable-in-conduit conductor for application in future fusion reactors

    Science.gov (United States)

    Qin, Jinggang; Wu, Yu; Li, Jiangang; Liu, Fang; Dai, Chao; Shi, Yi; Liu, Huajun; Mao, Zhehua; Nijhuis, Arend; Zhou, Chao; Yagotintsev, Konstantin A.; Lubkemann, Ruben; Anvar, V. A.; Devred, Arnaud

    2017-11-01

    The China Fusion Engineering Test Reactor (CFETR) is a new tokamak device whose magnet system includes toroidal field, central solenoid (CS) and poloidal field coils. The main goal is to build a fusion engineering tokamak reactor with about 1 GW fusion power and self-sufficiency by blanket. In order to reach this high performance, the magnet field target is 15 T. However, the huge electromagnetic load caused by high field and current is a threat for conductor degradation under cycling. The conductor with a short-twist-pitch (STP) design has large stiffness, which enables a significant performance improvement in view of load and thermal cycling. But the conductor with STP design has a remarkable disadvantage: it can easily cause severe strand indentation during cabling. The indentation can reduce the strand performance, especially under high load cycling. In order to overcome this disadvantage, a new design is proposed. The main characteristic of this new design is an updated layout in the triplet. The triplet is made of two Nb3Sn strands and one soft copper strand. The twist pitch of the two Nb3Sn strands is large and cabled first. The copper strand is then wound around the two superconducting strands (CWS) with a shorter twist pitch. The following cable stages layout and twist pitches are similar to the ITER CS conductor with STP design. One short conductor sample with a similar scale to the ITER CS was manufactured and tested with the Twente Cable Press to investigate the mechanical properties, AC loss and internal inspection by destructive examination. The results are compared to the STP conductor (ITER CS and CFETR CSMC) tests. The results show that the new conductor design has similar stiffness, but much lower strand indentation than the STP design. The new design shows potential for application in future fusion reactors.

  11. Design of force-cooled conductors for large fusion magnets

    Energy Technology Data Exchange (ETDEWEB)

    Dresner, L.; Lue, J.W.

    1977-01-01

    Conductors cooled by supercritical helium in forced convection are under active consideration for large toroidal fusion magnets. One of the central problems in designing such force cooled conductors is to maintain an adequate stability margin while keeping the pumping power tolerably low. A method has been developed for minimizing the pumping power for fixed stability by optimally choosing the matrix-to-superconductor and the metal-to-helium ratios. Such optimized conductors reduce pumping power requirements for fusion size magnets to acceptable limits. Furthermore, the mass flow and hence pumping losses can be varied through a magnet according to the local magnetic field and magnitude of desired stability margin. Force cooled conductors give flexibility in operation, permitting, for example, higher fields to be obtained than originally intended by lowering the bath temperature or increasing the pumping power or both. This flexibility is only available if the pumping power is low to begin with. Scaling laws for the pumping requirement and stability margin as functions of operating current density, number of strands and such physical parameters as stabilizer resistivity and critical current density, have been proved. Numerical examples will be given for design of conductors intended for use in large toroidal fusion magnet systems.

  12. Conceptual design study for a mirror fusion breeder

    International Nuclear Information System (INIS)

    Huang Jinhua; Deng Boquan; Li Guiqing

    1986-01-01

    A mirror fusion breeder, CHD, has been designed for providing plenty of nuclear fuel for light water reactors to meet the needs for rapid development of nuclear power in the first half of next century. The breeder is able to support the nuclear fuel needs for more than 10 LWRs of equal scale in power with fuel enriched directly in CHD without reprocessing. Measures are taken to flatten the power density distribution in the blanket so that fission is suppressed in the region close to the plasma, and by this way fuel production is enhanced for this direct enriched fusion breeder. In order to reduce the MHD pressure drop, LiPb flows in the blanket axially. Though the tritium inventory in the reactor is very low, special material and design have to be developed to reduce the permeation of tritium through the coolant pipes. The cost of electricity from the system, consisting of 11 LWR plants and one fusion breeder is predicted to be 1.05 times of that from a traditional LWR plant. This figure is insensitive both to the cost of CHD and its support ratio

  13. Design of force-cooled conductors for large fusion magnets

    International Nuclear Information System (INIS)

    Dresner, L.; Lue, J.W.

    1977-01-01

    Conductors cooled by supercritical helium in forced convection are under active consideration for large toroidal fusion magnets. One of the central problems in designing such force cooled conductors is to maintain an adequate stability margin while keeping the pumping power tolerably low. A method has been developed for minimizing the pumping power for fixed stability by optimally choosing the matrix-to-superconductor and the metal-to-helium ratios. Such optimized conductors reduce pumping power requirements for fusion size magnets to acceptable limits. Furthermore, the mass flow and hence pumping losses can be varied through a magnet according to the local magnetic field and magnitude of desired stability margin. Force cooled conductors give flexibility in operation, permitting, for example, higher fields to be obtained than originally intended by lowering the bath temperature or increasing the pumping power or both. This flexibility is only available if the pumping power is low to begin with. Scaling laws for the pumping requirement and stability margin as functions of operating current density, number of strands and such physical parameters as stabilizer resistivity and critical current density, have been proved. Numerical examples will be given for design of conductors intended for use in large toroidal fusion magnet systems

  14. Critical plasma-materials issues for fusion reactor designs

    International Nuclear Information System (INIS)

    Wilson, K.L.; Bauer, W.

    1983-01-01

    Plasma-materials interactions are a dominant driving force in the design of fusion power reactors. This paper presents a summary of plasma-materials interactions research. Emphasis is placed on critical aspects related to reactor design. Particular issues to be addressed are plasma edge characterization, hydrogen recycle, impurity introduction, and coating development. Typical wall fluxes in operating magnetically confined devices are summarized. Recent calculations of tritium inventory and first wall permeation, based on laboratory measurements of hydrogen recycling, are given for various reactor operating scenarios. Impurity introduction/wall erosion mechanisms considered include sputtering, chemical erosion, and evaporation (melting). Finally, the advanced material development for in-vessel components is discussed. (author)

  15. Design study of fusion Demo plant at JAERI

    International Nuclear Information System (INIS)

    Tobita, K.; Nishio, S.; Enoeda, M.

    2006-01-01

    Three options of fusion Demo plant are proposed characterized by functions of the center solenoid (Cs). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3 GW with a major radius of 5.5 m, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4 T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb 3 Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper

  16. Overview of the Fusion Engineering Device (FED) design

    International Nuclear Information System (INIS)

    Steiner, D.; Flanagan, C.A.

    1981-01-01

    The device has a major radius of 5.0 m with a plasma minor radius of 1.3 m elongated by 1.6. Capability is provided for operating the toroidal field coils up to 10 T, but the bulk of the operations are designed for 8 T. At 8-T conditions the fusion power is approx. 180 MW (neutron wall loading approx. 0.4 MW/m 2 ) and a plasma Q of approx. 5 is expected. At 10-T conditions, which are expected to be limited to about 10% of the total operations, the fusion power is approx. 450 MW (approx. 1.0 MW/m 2 ) and ignition is expected

  17. Overview of the fusion engineering device (FED) design

    International Nuclear Information System (INIS)

    Steiner, D.; Flanagan, C.A.

    1981-10-01

    The device has a major radius of 5.0 m with a plasma minor radius of 1.3 m elongated by 1.6. Capability is provided for operating the toroidal field coils up to 10 T, but the bulk of the operations are designed for 8 T. At 8-T conditions, the fusion power is approx. 180 MW (neutron wall loading approx. 0.4 MW/m 2 ) and a plasma Q of approx. 5 is expected. At 10-T conditions, which are expected to be limited to about 10% of the total operations, the fusion power is approx. 450 MW (approx. 1.0 MW/m 2 ) and ignition is expected

  18. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

  19. 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

  20. Osiris and SOMBRERO inertial confinement fusion power plant designs. Volume 2, Designs, assessments, and comparisons, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W.R.; Bieri, R.L.; Monsler, M.J.

    1992-03-01

    The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs.

  1. Low-density hydrocarbon foams for laser fusion targets: Progress report, 1987

    International Nuclear Information System (INIS)

    Haendler, B.L.; Buckley, S.R.; Chen, C.

    1988-06-01

    This report describes progress made in the development of direct-drive hydrocarbon foam targets for laser inertial confinement fusion during 1987. The foam materials are polystyrene, resorcinol-formaldehyde, carbonized resorcinol-formaldehyde, and cellulose acetate. The processes for making the foams, their properties, characterization techniques, and the relationship of their properties to target specifications are presented. Progress in the creation and testing of prototype targets is also described

  2. Block Ignition Inertial Confinement Fusion (ICF) with Condensed Matter Cluster Type Targets for p-B11 Powered Space Propulsion

    International Nuclear Information System (INIS)

    Miley, George H.; Hora, H.; Badziak, J.; Wolowski, J.; Sheng Zhengming; Zhang Jie; Osman, F.; Zhang Weiyan; Tuhe Xia

    2009-01-01

    The use of laser-driven Inertial Confinement Fusion (ICF) for space propulsion has been the subject of several earlier conceptual design studies, (see: Orth, 1998; and other references therein). However, these studies were based on older ICF technology using either 'direct' or 'in-direct x-ray driven' type target irradiation. Important new directions have opened for laser ICF in recent years following the development of 'chirped' lasers capable of ultra short pulses with powers of TW up to few PW which leads to the concept of 'fast ignition (FI)' to achieve higher energy gains from target implosions. In a recent publication the authors showed that use of a modified type of FI, termed 'block ignition' (Miley et al., 2008), could meet many of the requirements anticipated (but not then available) by the designs of the Vehicle for Interplanetary Space Transport Applications (VISTA) ICF fusion propulsion ship (Orth, 2008) for deep space missions. Subsequently the first author devised and presented concepts for imbedding high density condensed matter 'clusters' of deuterium into the target to obtain ultra high local fusion reaction rates (Miley, 2008). Such rates are possible due to the high density of the clusters (over an order of magnitude above cryogenic deuterium). Once compressed by the implosion, the yet higher density gives an ultra high reaction rate over the cluster volume since the fusion rate is proportional to the square of the fuel density. Most recently, a new discovery discussed here indicates that the target matrix could be composed of B 11 with proton clusters imbedded. This then makes p-B 11 fusion practical, assuming all of the physics issues such as stability of the clusters during compression are resolved. Indeed, p-B 11 power is ideal for fusion propulsion since it has a minimum of unwanted side products while giving most of the reaction energy to energetic alpha particles which can be directed into an exhaust (propulsion) nozzle. Power plants

  3. Block Ignition Inertial Confinement Fusion (ICF) with Condensed Matter Cluster Type Targets for p-B11 Powered Space Propulsion

    Science.gov (United States)

    Miley, George H.; Hora, H.; Badziak, J.; Wolowski, J.; Sheng, Zheng-Ming; Zhang, Jie; Osman, F.; Zhang, Weiyan; tu He, Xia

    2009-03-01

    The use of laser-driven Inertial Confinement Fusion (ICF) for space propulsion has been the subject of several earlier conceptual design studies, (see: Orth, 1998; and other references therein). However, these studies were based on older ICF technology using either "direct "or "in-direct x-ray driven" type target irradiation. Important new directions have opened for laser ICF in recent years following the development of "chirped" lasers capable of ultra short pulses with powers of TW up to few PW which leads to the concept of "fast ignition (FI)" to achieve higher energy gains from target implosions. In a recent publication the authors showed that use of a modified type of FI, termed "block ignition" (Miley et al., 2008), could meet many of the requirements anticipated (but not then available) by the designs of the Vehicle for Interplanetary Space Transport Applications (VISTA) ICF fusion propulsion ship (Orth, 2008) for deep space missions. Subsequently the first author devised and presented concepts for imbedding high density condensed matter "clusters" of deuterium into the target to obtain ultra high local fusion reaction rates (Miley, 2008). Such rates are possible due to the high density of the clusters (over an order of magnitude above cryogenic deuterium). Once compressed by the implosion, the yet higher density gives an ultra high reaction rate over the cluster volume since the fusion rate is proportional to the square of the fuel density. Most recently, a new discovery discussed here indicates that the target matrix could be composed of B11 with proton clusters imbedded. This then makes p-B11 fusion practical, assuming all of the physics issues such as stability of the clusters during compression are resolved. Indeed, p-B11 power is ideal for fusion propulsion since it has a minimum of unwanted side products while giving most of the reaction energy to energetic alpha particles which can be directed into an exhaust (propulsion) nozzle. Power plants using p

  4. Magnetized Target Fusion (MTF): A Low-Cost Fusion Development Path

    International Nuclear Information System (INIS)

    Lindemuth, I.R.; Siemon, R.E.; Kirkpatrick, R.C.; Reinovsky, R.E.

    1998-01-01

    Simple transport-based scaling laws are derived to show that a density and time regime intermediate between conventional magnetic confinement and conventional inertial confinement offers attractive reductions in system size and energy when compared to magnetic confinement and attractive reductions in heating power and intensity when compared to inertial confinement. This intermediate parameter space appears to be readily accessible by existing and near term pulsed power technologies. Hence, the technology of the Megagauss conference opens up an attractive path to controlled thermonuclear fusion

  5. Study on conceptual design system of tritium production fusion reactor

    International Nuclear Information System (INIS)

    He Kaihui

    2004-11-01

    Conceptual design of an advanced tritium production reactor based on spherical torus, which is intermediate application of fusion energy, was presented. Different from traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can within vacuum vessel in order to produce 1 kg excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented. Besides systematical analyses; design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (author)

  6. Conceptual design study of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1986-11-01

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. During two years from 1984 to 1985 FER concept was reviewed and redesigned. This report is the summary of the results obtained in the review and redesign activities in 1984 and 85. In the first year FER concept was discussed again and its frame work was reestablished. According to the new frame work the major reactor components of FER were designed. In the second year the whole plant system design including plant layout plan was conducted as well as the more detailed design analysis of the reactor conponents. The newly established frame for FER design is as follows: 1) Plasma : Self-ignition. 2) Operation scenario : Quasi-steady state operation with long burn pulse. 3) Neutron fluence on the first wall : 0.3 MWY/M 2 . 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

  7. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

    1987-09-01

    This report describes the results of the reactor configuration/structure design for the fusion experimental reactor (FER) performed in FY 1986. The design was intended to meet the physical and engineering mission of the next step device which was decided by the subcommittee on the next step device of the nuclear fusion council. The objectives of the design study in FY 1986 are to advance and optimize the design concept of the last year because the recommendation of the subcommittee was basically the same as the design philosophy of the last year. Six candidate reactor configurations which correspond to options C ∼ D presented by the subcommittee were extensively examined. Consequently, ACS reactor (Advanced Option-C with Single Null Divertor) was selected as the reference configuration from viewpoints of technical risks and cost performance. Regarding the reactor structure, the following items were investigated intensively: minimization of reactor size, protection of first wall against plasma disruption, simplification of shield structure, reactor configuration which enables optimum arrangement of poloidal field coils. (author)

  8. Conceptual design considerations and neutronics of lithium fall laser target chambers

    International Nuclear Information System (INIS)

    Meier, W.R.; Thomson, W.B.

    1978-01-01

    Atomics International and Lawrence Livermore Laboratory are involved in the conceptual design of a laser fusion power plant incorporating the lithium fall target chamber. In this paper we discuss some of the more important design considerations for the target chamber and evaluate its nuclear performance. Sizing and configuration of the fall, hydraulic effects, and mechanical design considerations are addressed. The nuclear aspects examined include tritium breeding, energy deposition, and radiation damage

  9. 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

  10. Method for selecting hollow microspheres for use in laser fusion targets

    Science.gov (United States)

    Farnum, Eugene H.; Fries, R. Jay; Havenhill, Jerry W.; Smith, Maurice Lee; Stoltz, Daniel L.

    1976-01-01

    Hollow microspheres having thin and very uniform wall thickness are useful as containers for the deuterium and tritium gas mixture used as a fuel in laser fusion targets. Hollow microspheres are commercially available; however, in commercial lots only a very small number meet the rigid requirements for use in laser fusion targets. Those meeting these requirements may be separated from the unsuitable ones by subjecting the commercial lot to size and density separations and then by subjecting those hollow microspheres thus separated to an external pressurization at which those which are aspherical or which have nonuniform walls are broken and separating the sound hollow microspheres from the broken ones.

  11. Calculation of fusion gain in fast ignition with magnetic target by relativistic electrons and protons

    International Nuclear Information System (INIS)

    Parvazian, A.; Javani, A.

    2010-01-01

    Fast ignition is a new method for inertial confinement fusion in which the compression and ignition steps are separated. In the first stage, fuel is compressed by laser or ion beams. In the second phase, relativistic electrons are generated by pettawat laser in the fuel. Also, in the second phase 5-35 MeV protons can be generated in the fuel. Electrons or protons can penetrate in to the ultra-dense fuel and deposit their energy in the fuel. More recently, cylindrical rather than spherical fuel chambers with magnetic control in the plasma domain have been also considered. This is called magnetized target fusion. Magnetic field has effects on relativistic electrons energy deposition rate in fuel. In this work, fast ignition method in cylindrical fuel chambers is investigated and transportation of the relativistic electrons and protons is calculated using MCNPX and FLUKA codes with 0.25 and 0.5 tesla magnetic field in single and dual hot spot. Furthermore, the transfer rate of relativistic electrons and high energy protons to the fuel and fusion gain are calculated. The results show that the presence of external magnetic field guarantees higher fusion gain, and relativistic electrons are much more appropriate objects for ignition. Magnetized target fusion in dual hot spot can be considered as an appropriate substitution for the current inertial confinement fusion techniques.

  12. Elements of power plant design for inertial fusion energy. Final report of a coordinated research project 2000-2004

    International Nuclear Information System (INIS)

    2005-06-01

    There are two major approaches in fusion energy research: magnetic fusion energy (MFE) and inertial fusion energy (IFE). The basic physics of IFE (compression and ignition of small fuel pellets containing deuterium and tritium) is being increasingly understood. Based on recent advances by individual countries, IFE has reached a stage at which benefits could be obtained from a coordinated approach in the form of an IAEA Coordinated Research Project (CRP) on Elements of Power Plant Design for Inertial Fusion Energy. This CRP helped Member States to promote the development of plasma/fusion technology transfer and to emphasize safety and environmental advantages of fusion energy. The CRP was focused on interface issues including those related to, - the driver/target interface (e.g. focusing and beam uniformity required by the target), - the driver/chamber interface (e.g. final optics and magnets protection and shielding), - and the target/chamber interface (e.g. target survival during injection, target positioning and tracking in the chamber). The final report includes an assessment of the state of the art of the technologies required for an IFE power plant (drivers, chambers, targets) and systems integration as presented and evaluated by members of the CRP. Additional contributions by cost free invited experts to the final RCM are included. The overall objective of this CRP was to foster the inertial fusion energy development by improving international cooperation. The variety of contributions compiled in this TECDOC reflects, that the goal of stimulating the exchange of knowledge was well achieved. Further the CRP led to the creation of a network, which not only exchanged their scientific results, but also developed healthy professional relations and strong mutual interest in the work of the group members

  13. 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.

  14. Mechanical design aspects of a tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.

    1977-01-01

    Two ''plugs'' of dense plasma at either end of a central solenoid cell form the basis of a new mirror fusion power plant concept. A central cell blanket design is presented. Modules on crawler tracks serviced by remote welding and handling machines of very simple design are important features resulting from linear axisymmetric geometry. Three blanket designs are considered and the best one presented in some detail. It has lithium as the breeder material, helium cooled. ''Plug'' magnet field strengths must be high. A novel magnet is presented to satisfy the physics of the end plugs. Beam sources at 1,200 KV present special problems. Methods of voltage standoff, arc damage control, and neutralization are discussed. New secondary containment ideas are presented to allow removable roof sections of balanced design

  15. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    Saito, Ryusei; Kashihara, Shin-ichiro; Itoh, Shin-ichi

    1987-08-01

    This report describes the results of conceptual design study on plant systems for the Fusion Experimental Reactor (FY86 FER). Design studies for FER plant systems have been continued from FY85, especially for design modifications made in accordance with revisions of plasma scaling parameters and system improvements. This report describes 1) system construction, 2) site and reactor building plan, 3) repaire and maintenance system, 4) tritium circulation system, 5) heating, ventilation and air conditioning system, 6) tritium clean-up system, 7) cooling and baking system, 8) waste treatment and storage system, 9) control system, 10) electric power system, 11) site factory plan, all of which are a part of FY86 design work. The plant systems described in this report generally have been based on the FY86 FER (ACS Reactor) which is an one of the six candidates for FER. (author)

  16. 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

  17. Conceptual design report for a Fusion Engineering Device sector-handling machine and movable manipulator system

    International Nuclear Information System (INIS)

    Watts, K.D.; Masson, L.S.; McPherson, R.S.

    1982-10-01

    Design requirements, trade studies, design descriptions, conceptual designs, and cost estimates have been completed for the Fusion Engineering Device sector handling machine, movable manipulator system, subcomponent handling machine, and limiter blade handling machine. This information will be used by the Fusion Engineering Design Center to begin to determine the cost and magnitude of the effort required to perform remote maintenance on the Fusion Engineering Device. The designs presented are by no means optimum, and the costs estimates are rough-order-of-magnitude

  18. Binding and Fusion of Extracellular Vesicles to the Plasma Membrane of Their Cell Targets.

    Science.gov (United States)

    Prada, Ilaria; Meldolesi, Jacopo

    2016-08-09

    Exosomes and ectosomes, extracellular vesicles of two types generated by all cells at multivesicular bodies and the plasma membrane, respectively, play critical roles in physiology and pathology. A key mechanism of their function, analogous for both types of vesicles, is the fusion of their membrane to the plasma membrane of specific target cells, followed by discharge to the cytoplasm of their luminal cargo containing proteins, RNAs, and DNA. Here we summarize the present knowledge about the interactions, binding and fusions of vesicles with the cell plasma membrane. The sequence initiates with dynamic interactions, during which vesicles roll over the plasma membrane, followed by the binding of specific membrane proteins to their cell receptors. Membrane binding is then converted rapidly into fusion by mechanisms analogous to those of retroviruses. Specifically, proteins of the extracellular vesicle membranes are structurally rearranged, and their hydrophobic sequences insert into the target cell plasma membrane which undergoes lipid reorganization, protein restructuring and membrane dimpling. Single fusions are not the only process of vesicle/cell interactions. Upon intracellular reassembly of their luminal cargoes, vesicles can be regenerated, released and fused horizontally to other target cells. Fusions of extracellular vesicles are relevant also for specific therapy processes, now intensely investigated.

  19. Technical Design Report, Second Target Station

    International Nuclear Information System (INIS)

    Galambos, John D.; Anderson, David E.; Bethea, Katie L.; Carden, W. F.; Chae, Steven M.; Bechtol, D.; Brown, N.; Clark, A.

    2015-01-01

    The Second Target Station (STS) is a proposed upgrade for SNS. It includes a doubling of the accelerator power and an additional instrument hall. The new instrument hall will receive a 467 kW 10 Hz beam. The parameters and preliminary design aspects of the STS are presented for the accelerator, target systems, instrument hall, instruments and civil construction aspects.

  20. Application of uncertainty analysis in conceptual fusion reactor design

    International Nuclear Information System (INIS)

    Wu, T.; Maynard, C.W.

    1979-01-01

    The theories of sensitivity and uncertainty analysis are described and applied to a new conceptual tokamak fusion reactor design--NUWMAK. The responses investigated in this study include the tritium breeding ratio, first wall Ti dpa and gas productions, nuclear heating in the blanket, energy leakage to the magnet, and the dpa rate in the superconducting magnet aluminum stabilizer. The sensitivities and uncertainties of these responses are calculated. The cost/benefit feature of proposed integral measurements is also studied through the uncertainty reductions of these responses

  1. Shield design for the Fusion Materials Irradiation Test facility

    International Nuclear Information System (INIS)

    Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

    1983-03-01

    The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

  2. Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, Kevin James [Univ. of California, Berkeley, CA (United States)

    2010-04-08

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 μm of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles

  3. Fusion-based multi-target tracking and localization for intelligent surveillance systems

    Science.gov (United States)

    Rababaah, Haroun; Shirkhodaie, Amir

    2008-04-01

    In this paper, we have presented two approaches addressing visual target tracking and localization in complex urban environment. The two techniques presented in this paper are: fusion-based multi-target visual tracking, and multi-target localization via camera calibration. For multi-target tracking, the data fusion concepts of hypothesis generation/evaluation/selection, target-to-target registration, and association are employed. An association matrix is implemented using RGB histograms for associated tracking of multi-targets of interests. Motion segmentation of targets of interest (TOI) from the background was achieved by a Gaussian Mixture Model. Foreground segmentation, on other hand, was achieved by the Connected Components Analysis (CCA) technique. The tracking of individual targets was estimated by fusing two sources of information, the centroid with the spatial gating, and the RGB histogram association matrix. The localization problem is addressed through an effective camera calibration technique using edge modeling for grid mapping (EMGM). A two-stage image pixel to world coordinates mapping technique is introduced that performs coarse and fine location estimation of moving TOIs. In coarse estimation, an approximate neighborhood of the target position is estimated based on nearest 4-neighbor method, and in fine estimation, we use Euclidean interpolation to localize the position within the estimated four neighbors. Both techniques were tested and shown reliable results for tracking and localization of Targets of interests in complex urban environment.

  4. Rational design of an EGF-IL18 fusion protein: Implication for developing tumor therapeutics

    International Nuclear Information System (INIS)

    Lu Jianxin; Peng Ying; Meng Zhefeng; Jin Liqin; Lu Yongsui; Guan Minxin

    2005-01-01

    Interleukin-18 (IL-18) is a proinflammatory cytokine. This protein has a role in regulating immune responses and exhibits significant anti-tumor activities. Epidermal growth factor (EGF) is an important growth factor that plays a central role in the regulation of cell cycle and differentiation. It was proposed that a targeted delivery of IL-18 by generation of IL-18-EGF fusion protein might decrease adverse effects and result in enhancing cytotoxic and antitumor activities. In the present study, a fusion protein, consisting of EGFR binding domain fused to human IL-18 mature peptide via a linker peptide of (Gly 4 Ser) 3, was constructed and expressed in the insect cell line Sf9 using Bac-to-Bac baculovirus expression system. We showed that the purified recombinant fusion protein induced similar levels of IFN-γ to that of native IL-18 protein in human PBMC in the presence of ConA. Furthermore, EGF receptor competitive test in human epithelial cancer A431 cell line showed that EGF-IL18 fusion protein can specifically bind with EGFR by competing with native EGF protein. These suggest that this rationally designed protein can be further developed as novel tumor therapeutics

  5. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  6. A Perspective on Equipment Design for Fusion Remote Handling

    International Nuclear Information System (INIS)

    Mills, S.; Haist, B.; Hamilton, D.

    2006-01-01

    For 8 years, JET remote operations have become more capable and confident. Many tasks have been successfully completed, even those never intended to be remote maintenance activities. The general approach to the provision of remote handling equipment at JET has been the preferred use of commercially-off-the-shelf equipment. In the areas of electrical, electronic, software and control this approach has been generally achievable. However, in the area of mechanical equipment it has been more difficult. In particular the RH tooling has been almost entirely bespoke as its requirements are highly sensitive to the design of the JET component being handled and there are many design variations. Hence, JET has required the design and manufacture of over 700 types of bespoke RH equipment. This paper will discuss the experience of introducing and developing remote handling mechanical equipment for JET. The paper will cover the relationship between the remote handling equipment and the JET component design and the potential for improving the design function. A major lesson from the introduction of remote handling to JET has been demonstration of the very close interdependency of the design of JET components with design of remote handling tooling. The JET remote handling manual was originally introduced as the vehicle to ensure remote handling compatibility by the introduction of standards. Experience has shown that in general the remote handling manual approach has been insufficient. Future fusion machines will be much more complex than JET and will demand even greater remote handling compatibility. This paper will discuss possible methods for improving this process. Equipment operating in a high radiation environment must be dependable It may spend part of its time in areas that would be extremely difficult to recover from in the case of failure. The equipment may also have a high duty cycle to minimise shutdown times and probably cannot be manually inspected on a frequent

  7. A preliminary conceptual design study for Korean fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

    2013-10-15

    Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

  8. Proposed design criteria for a fusion facility electrical ground system

    International Nuclear Information System (INIS)

    Armellino, C.A.

    1983-01-01

    Ground grid design considerations for a nuclear fusion reactor facility are no different than any other facility in that the basis for design must be safety first and foremost. Unlike a conventional industrial facility the available fault energy comes not only from the utility source and in-house rotating machinery, but also from energy storage capacitor banks, collapsing magnetic fields and D.C. transmission lines. It is not inconceivable for a fault condition occurrence where all available energy can be discharged. The ground grid must adequately shunt this sudden energy discharge in a way that personnel will not be exposed by step and/or touch to hazardous energy levels that are in excess of maximum tolerable levels for humans. Fault energy discharge rate is a function of the ground grid surge impedance characteristic. Closed loop paths must be avoided in the ground grid design so that during energy discharge no stray magnetic fields or large voltage potentials between remote points can be created by circulating currents. Single point connection of equipment to the ground grid will afford protection to personnel and sensitive equipment by reducing the probability of circulating currents. The overall ground grid system design is best illustrated as a wagon wheel concept with the fusion machine at the center. Radial branches or spokes reach out to the perimeter limits designated by step-and-touch high risk areas based on soil resistivity criteria considerations. Conventional methods for the design of a ground grid with all of its radial branches are still pertinent. The center of the grid could include a deep well single ground rod element the length of which is at least equivalent to the radius of an imaginary sphere that enshrouds the immediate machine area. Special facilities such as screen rooms or other shielded areas are part of the ground grid system by way of connection to radial branches

  9. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    Science.gov (United States)

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  10. Automation of fusion first wall design using artificial intelligence technique

    International Nuclear Information System (INIS)

    Yoshimura, Shinobu; Yagawa, Genki; Mochizuki, Yoshihiko

    1990-01-01

    This paper describes the application of artificial intelligence techniques to a design automation of the fusion first wall to be operated in the complex environment where huge electromagnetic and thermal loading as well as heavy neutron irradiation occur. As a basic strategy of designing structure shape considering many coupled phenomena, an ordinary design procedure based on the generate and test strategy is adopted because of its simplicity and broad applicability. To automate the design procedure with maintaining its flexibility, extensibility and efficiency, artificial intelligence techniques are utilized in the following. An object-oriented knowledge representation technique is adopted to store knowledge modules, that is, objects, related to the first wall design, while a data-flow processing technique is utilized as an inference mechanism among the knowledge modules. These techniques realize the flexibility and extensibility of the system. Moreover, as an efficient design modification mechanism, which is essential in a design process, an empirical approach based on experts' empirical knowledge and a mathematical approach based on a kind of numerical sensitivity analysis are introduced. The developed system is applied to a simple example of the design of a two-dimensional model of the first wall with a cooling channel, and its fundamental performance is clearly demonstrated. (author)

  11. Recent developments in the design of conceptual fusion reactors

    International Nuclear Information System (INIS)

    Ribe, F.L.

    1977-01-01

    Since the first round of conceptual fusion reactor designs in 1973 - 1974, there has been considerable progress in design improvement. Two recent tokamak designs of the Wisconsin and Culham groups, with increased plasma beta and wall loading (power density), lead to more compact reactors with easier maintenance. The Reference Theta-Pinch Reactor has undergone considerable upgrading in the design of the first wall insulator and blanket. In addition, a conceptual homopolar energy storage and transfer system has been designed. In the case of the mirror reactor, there are design changes toward improved modular construction and ease of handling, as well as improved direct converters. Conceptual designs of toroidal-multiple-mirror, liner-compression, and reverse-field pinch reactors are also discussed. A design is presented of a toroidal multiple-mirror reactor that combines the advantages of steady-state operation and high-aspect ratio. The liner-compression reactor eliminates a major problem of radiation damage by using a liquid-metal first wall that also serves as a neutron-thermalizing blanket. The reverse-field pinch reactor operates at higher beta, larger current density and larger aspect ratio than a tokamak reactor. These properties allow the possibility of ignition by ohmic heating alone and greater ease of maintenance

  12. SAR Target Recognition Based on Multi-feature Multiple Representation Classifier Fusion

    Directory of Open Access Journals (Sweden)

    Zhang Xinzheng

    2017-10-01

    Full Text Available In this paper, we present a Synthetic Aperture Radar (SAR image target recognition algorithm based on multi-feature multiple representation learning classifier fusion. First, it extracts three features from the SAR images, namely principal component analysis, wavelet transform, and Two-Dimensional Slice Zernike Moments (2DSZM features. Second, we harness the sparse representation classifier and the cooperative representation classifier with the above-mentioned features to get six predictive labels. Finally, we adopt classifier fusion to obtain the final recognition decision. We researched three different classifier fusion algorithms in our experiments, and the results demonstrate thatusing Bayesian decision fusion gives thebest recognition performance. The method based on multi-feature multiple representation learning classifier fusion integrates the discrimination of multi-features and combines the sparse and cooperative representation classification performance to gain complementary advantages and to improve recognition accuracy. The experiments are based on the Moving and Stationary Target Acquisition and Recognition (MSTAR database,and they demonstrate the effectiveness of the proposed approach.

  13. Distributed ISAR Subimage Fusion of Nonuniform Rotating Target Based on Matching Fourier Transform.

    Science.gov (United States)

    Li, Yuanyuan; Fu, Yaowen; Zhang, Wenpeng

    2018-06-04

    In real applications, the image quality of the conventional monostatic Inverse Synthetic Aperture Radar (ISAR) for the maneuvering target is subject to the strong fluctuation of Radar Cross Section (RCS), as the target aspect varies enormously. Meanwhile, the maneuvering target introduces nonuniform rotation after translation motion compensation which degrades the imaging performance of the conventional Fourier Transform (FT)-based method in the cross-range dimension. In this paper, a method which combines the distributed ISAR technique and the Matching Fourier Transform (MFT) is proposed to overcome these problems. Firstly, according to the characteristics of the distributed ISAR, the multiple channel echoes of the nonuniform rotation target from different observation angles can be acquired. Then, by applying the MFT to the echo of each channel, the defocused problem of nonuniform rotation target which is inevitable by using the FT-based imaging method can be avoided. Finally, after preprocessing, scaling and rotation of all subimages, the noncoherent fusion image containing all the RCS information in all channels can be obtained. The accumulation coefficients of all subimages are calculated adaptively according to the their image qualities. Simulation and experimental data are used to validate the effectiveness of the proposed approach, and fusion image with improved recognizability can be obtained. Therefore, by using the distributed ISAR technique and MFT, subimages of high-maneuvering target from different observation angles can be obtained. Meanwhile, by employing the adaptive subimage fusion method, the RCS fluctuation can be alleviated and more recognizable final image can be obtained.

  14. Design study of electrostatically plugged cusp fusion reactor

    International Nuclear Information System (INIS)

    Dolan, T.J.

    1976-01-01

    This study concentrates on the following aspects of an electrostatically plugged cusp reactor that will be different from other fusion reactor designs: the coil geometry and structural supports, high voltage electrodes, plasma parameters, power balance, and operating cycle. Assuming the electron density distribution in the anodes to have a characteristic width of two electron Larmor radii, which is consistent with present experimental results, the theory predicts that a device with a magnetic field strength, B = 8 T sustained solely by electron beam injection at 300 kV will have a power gain ratio, Q, of about 5. A toroidal multipole cusp configuration with six cusps was selected for the present design, based on a study of the ratio of plasma volume to coil volume. Coil forces are sustained by cryogenic trusses between like coils, fiberglass compression columns, and room temperature hoops. Radiation collimators in front of the high voltage electrodes greatly reduce the radiation impinging on the cathodes, helping to avoid breakdown and to prolong insulator life. The operating cycle consists of a startup period of about 20 s, followed by a fusion burn period lasting about 200 s (limited by impurity buildup) and a 20-s flushing period

  15. X-ray absorption in characterization of laser fusion targets

    International Nuclear Information System (INIS)

    Clement, X.; Coudeville, A.; Eyharts, P.; Perrine, J.P.; Rouillard, R.

    1982-11-01

    Many plastic or metal coated targets are opaque, so their thickness and thickness uniformity cannot be obtained by optical means. Therefore, we have built and tested a new system using monochromatic X-ray absorption measurements. This system is also able to perform non-destructive measurements of argon fill pressure in glass microballoons. The X-ray source is a diffraction tube with a chromium target and fine focus (0.4 x 0.8 mm 2 ). Since monochromatic calculations are involved in this method, we use electronic discrimination to isolate the chromium Kα line (5.4 keV) from the bremsstrahlung spectrum. The detectors are xenon-filled proportional counters. The system is composed of two beams (10 μm in diameter), one used as a reference and the other as the measurement arm. A PET desk computer is coupled ot the experiment. We achieved a precision better than 10% for gold layers in the range of 0.1 to 1 μm, and better than 20% for argon pressures in the range of 5 - 13 bars

  16. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

    A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations

  17. Design of the LBNF Beamline Target Station

    Energy Technology Data Exchange (ETDEWEB)

    Tariq, S. [Fermilab; Ammigan, K. [Fermilab; Anderson, K.; ; Buccellato, S. A. [Fermilab; Crowley, C. F. [Fermilab; Hartsell, B. D. [Fermilab; Hurh, P. [Fermilab; Hylen, J. [Fermilab; Kasper, P. [Fermilab; Krafczyk, G. E. [Fermilab; Lee, A. [Fermilab; Lundberg, B. [Fermilab; Reitzner, S. D. [Fermilab; Sidorov, V. [Fermilab; Stefanik, A. M. [Fermilab; Tropin, I. S. [Fermilab; Vaziri, K. [Fermilab; Williams, K. [Fermilab; Zwaska, R. M. [Fermilab; Densham, C. [RAL, Didcot

    2016-10-01

    The Long Baseline Neutrino Facility (LBNF) project will build a beamline located at Fermilab to create and aim an intense neutrino beam of appropriate energy range toward the DUNE detectors at the SURF facility in Lead, South Dakota. Neutrino production starts in the Target Station, which consists of a solid target, magnetic focusing horns, and the associated sub-systems and shielding infrastructure. Protons hit the target producing mesons which are then focused by the horns into a helium-filled decay pipe where they decay into muons and neutrinos. The target and horns are encased in actively cooled steel and concrete shielding in a chamber called the target chase. The reference design chase is filled with air, but nitrogen and helium are being evaluated as alternatives. A replaceable beam window separates the decay pipe from the target chase. The facility is designed for initial operation at 1.2 MW, with the ability to upgrade to 2.4 MW, and is taking advantage of the experience gained by operating Fermilab’s NuMI facility. We discuss here the design status, associated challenges, and ongoing R&D and physics-driven component optimization of the Target Station.

  18. Safety considerations in the design of the fusion engineering device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    Safety considerations play a significant role in the design of a near-term Fusion Engineering Device (FED). For the safety of the general public and the plant workers, the radiation environment caused by the reacting plasma and the potential release of tritium fuel are the dominant considerations. The U.S. Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris. The highly activated device components make it necessary to design many of the maintenance activities in the reactor building for totally remote operation. The hot cell facility has evolved as a totally remote maintenance facility due to the high radiation levels of the device components. Safety considerations have had substantial impacts on the design of FED. Several examples of safety-related design impacts are discussed in the paper. Feasible solutions have been identified for all outstanding safety-related items, and additional optimization of these solutions is anticipated in future design studies

  19. Inertial confinement fusion target component fabrication and technology development support: Annual report, October 1, 1995--September 30, 1996

    International Nuclear Information System (INIS)

    Hoppe, M.

    1997-02-01

    On December 30, 1990, the U.S. Department of Energy entered into a contract with General Atomics (GA) to be the Inertial Confinement Fusion (ICF) Target Component Fabrication and Technology Development Support contractor. In September 1995 this contract ended and a second contract was issued for us to continue this ICF target support work. This report documents the technical activities of the period October 1, 1995 through September 30, 1996. During this period, GA and our partners WJ Schafer Associates (WJSA) and Soane Technologies, Inc. (STI) were assigned 14 formal tasks in support of the Inertial Confinement Fusion program and its five laboratories. A portion of the effort on these tasks included providing direct open-quotes Onsite Supportclose quotes at Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), and Sandia National Laboratory Albuquerque (SNLA). We fabricated and delivered over 800 gold-plated hohlraum mandrels to LLNL, LANL and SNLA. We produced nearly 1,200 glass and plastic target capsules for LLNL, LANL, SNLA and University of Rochester/Laboratory for Laser Energetics (UR/LLE). We also delivered over 100 flat foil targets for Naval Research Lab (NRL) and SNLA in FY96. This report describes these target fabrication activities and the target fabrication and characterization development activities that made the deliveries possible. The ICF program is anticipating experiments at the OMEGA laser and the National Ignition Facility (NIF) which will require capsules containing cryogenic layered D 2 or deuterium-tritium (DT) fuel. We are part of the National Cryogenic Target Program to create and demonstrate viable ways to generate and characterize cryogenic layers. Substantial progress has been made on ways to both create and characterize viable layers. During FY96, significant progress was made in the design of the OMEGA Cryogenic Target System that will field cryogenic targets on OMEGA

  20. Towards a strategy of reliable fusion first-wall design

    International Nuclear Information System (INIS)

    Schultz, J.H.

    1981-05-01

    Fusion first walls are subject to a large number of possible failure mechanisms, including erosion due to sputtering, arcing, blistering and vaporization and crack growth due to thermal and magnetic stresses. Each of these failure mechanisms is poorly characterized and has the potential of being severe. A strategy for designing reliably in the face of great uncertainty is discussed. Topological features beneficial to reactor availability are identified. The integration of limiter pumping with rf wave launching is discussed, as a means of simplifying reactor design. The concept of a sewer limiter is introduced, as a possible long-life limiter topology. The concept of flexible armor is discussed, as a means of extending maximum life

  1. A conceptual design of LIB fusion reactor: UTLIF(2)

    International Nuclear Information System (INIS)

    Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

    1984-01-01

    UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

  2. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    Seki, Yasushi; Iida, Hiromasa; Honda, Tsutomu.

    1987-08-01

    This report describes the study on safety for FER(Fusion Experimental Reactor) which has been designed as a next step machine to the JT-60. Though the final purpose of this study is to have an image of design base accident, maximum credible accident and to assess their risk or probability, etc., as FER plant system, the emphasis of this years study is placed on fuel-gas circulation system where the tritium inventory is maximum. This report consists of two chapters. The first chapter of this report summaries the FER system and describes FMEA(Failure Mode and Effect Analysis) and related accident progression sequence for FER plant system as a whole. The second chapter of this report is focused on fuel-gas circulation system including the purification, isotope separation system and storage system. Here, probability of risk is assessed by the probabilistic risk analysis (PRA) procedure based on FMEA, ETA and FTA. (author)

  3. High efficiency targets for high gain inertial confinement fusion

    International Nuclear Information System (INIS)

    Gardner, J.H.; Bodner, S.E.

    1986-01-01

    Rocket efficiencies as high as 15% are possible using short wavelength lasers and moderately high aspect ratio pellet designs. These designs are made possible by two recent breakthroughs in physics constraints. First is the development of the Induced Spatial Incoherence (ISI) technique which allows uniform illumination of the pellet and relaxes the constraint of thermal smoothing, permitting the use of short wavelength laser light. Second is the discovery that the Rayleigh-Taylor growth rate is considerably reduced at the short laser wavelengths. By taking advantage of the reduced constraints imposed by nonuniform laser illumination and Rayleigh-Taylor instability, pellets using 1/4 micron laser light and initial aspect ratios of about 10 (with in flight aspect ratios of about 150 to 200) may produce energy gains as high as 200 to 250

  4. Conceptual design of China fusion power plant FDS-II

    International Nuclear Information System (INIS)

    Wu, Y.; Liu, S.; Chen, H.

    2007-01-01

    As one of the series of fusion system design concepts developed by the FDS Team of China, FDS-II is designated to exploit and evaluate potential attractiveness of fusion energy application for the generation of electricity on the basis of conservatively advanced plasma parameters, which can be limitedly extrapolated from the successful operation of ITER. The principle of the blanket design is established in both the feasibility and potential attractiveness of technology to meet the requirement for tritium self-sufficiency, safety margin, operation economy and environment protection etc. The plasma physics and engineering parameters of FDS-II are selected on the basis of the progress in recent experiments and associated theoretical studies of magnetic confinement fusion plasma with a fusion power of 2∝3 GW. The neutron wall load of 2∝3 MW/m 2 and the surface heat flux of 0.5∝1 MW/m 2 are considered for high effective power conversion. The ''multi-modules'' scenario is adopted in the FDS-II blanket design to reduce thermal stress and electromagnetic forces under plasma disruption, with liquid metal lithium lead (LiPb) as tritium breeder, the Reduced Activation Ferritic/Martensitic (RAFM) steel as structural material. Two options of specific liquid LiPb blanket concepts have been proposed, named the Dual-cooled Lithium Lead (DLL) breeder blanket and the Quasi-Static Lithium Lead (SLL) breeder blanket. The DLL blanket is a dual-cooled LiPb breeder system with helium gas to cool the first wall and main structure and LiPb eutectic to be self-cooled. The flow channel inserts (FCIs), e.g. SiCf/SiC composites, are designed as the thermal and electrical insulators inside the LiPb flow channels to reduce the magnetohydrodynamic (MHD) pressure drop and to allow the coolant LiPb outlet temperature up to 700 C for high thermal efficiency. The SLL blanket is another option of the FDS-II blanket with the technology developed relatively easily. To avoid or mitigate the

  5. The role of Z-pinches and related configurations in magnetized target fusion

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1997-01-01

    The use of a magnetic field within a fusion target is now known as Magnetized Target Fusion in the US and as MAGO (Magnitnoye Obzhatiye, or magnetic compression) in Russia. In contrast to direct, hydrodynamic compression of initially ambient-temperature fuel (e.g., ICF), MTF involves two steps: (a) formation of a warm, magnetized, wall-confined plasma of intermediate density within a fusion target prior to implosion; (b) subsequent quasi-adiabatic compression and heating of the plasma by imploding the confining wall, or pusher. In many ways, MTF can be considered a marriage between the more mature MFE and ICF approaches, and this marriage potentially eliminates some of the hurdles encountered in the other approaches. When compared to ICF, MTF requires lower implosion velocity, lower initial density, significantly lower radial convergence, and larger targets, all of which lead to substantially reduced driver intensity, power, and symmetry requirements. When compared to MFE, MTF does not require a vacuum separating the plasma from the wall, and, in fact, complete magnetic confinement, even if possible, may not be desirable. The higher density of MTF and much shorter confinement times should make magnetized plasma formation a much less difficult step than in MFE. The substantially lower driver requirements and implosion velocity of MTF make z-pinch magnetically driven liners, magnetically imploded by existing modern pulsed power electrical current sources, a leading candidate for the target pusher of an MTF system

  6. Hefei experimental hybrid fusion-fission reactor conceptual design

    International Nuclear Information System (INIS)

    Qiu Lijian; Luan Guishi; Xu Qiang

    1992-03-01

    A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

  7. Joint sparsity based heterogeneous data-level fusion for target detection and estimation

    Science.gov (United States)

    Niu, Ruixin; Zulch, Peter; Distasio, Marcello; Blasch, Erik; Shen, Dan; Chen, Genshe

    2017-05-01

    Typical surveillance systems employ decision- or feature-level fusion approaches to integrate heterogeneous sensor data, which are sub-optimal and incur information loss. In this paper, we investigate data-level heterogeneous sensor fusion. Since the sensors monitor the common targets of interest, whose states can be determined by only a few parameters, it is reasonable to assume that the measurement domain has a low intrinsic dimensionality. For heterogeneous sensor data, we develop a joint-sparse data-level fusion (JSDLF) approach based on the emerging joint sparse signal recovery techniques by discretizing the target state space. This approach is applied to fuse signals from multiple distributed radio frequency (RF) signal sensors and a video camera for joint target detection and state estimation. The JSDLF approach is data-driven and requires minimum prior information, since there is no need to know the time-varying RF signal amplitudes, or the image intensity of the targets. It can handle non-linearity in the sensor data due to state space discretization and the use of frequency/pixel selection matrices. Furthermore, for a multi-target case with J targets, the JSDLF approach only requires discretization in a single-target state space, instead of discretization in a J-target state space, as in the case of the generalized likelihood ratio test (GLRT) or the maximum likelihood estimator (MLE). Numerical examples are provided to demonstrate that the proposed JSDLF approach achieves excellent performance with near real-time accurate target position and velocity estimates.

  8. Moving target detection based on temporal-spatial information fusion for infrared image sequences

    Science.gov (United States)

    Toing, Wu-qin; Xiong, Jin-yu; Zeng, An-jun; Wu, Xiao-ping; Xu, Hao-peng

    2009-07-01

    Moving target detection and localization is one of the most fundamental tasks in visual surveillance. In this paper, through analyzing the advantages and disadvantages of the traditional approaches about moving target detection, a novel approach based on temporal-spatial information fusion is proposed for moving target detection. The proposed method combines the spatial feature in single frame and the temporal properties within multiple frames of an image sequence of moving target. First, the method uses the spatial image segmentation for target separation from background and uses the local temporal variance for extracting targets and wiping off the trail artifact. Second, the logical "and" operator is used to fuse the temporal and spatial information. In the end, to the fusion image sequence, the morphological filtering and blob analysis are used to acquire exact moving target. The algorithm not only requires minimal computation and memory but also quickly adapts to the change of background and environment. Comparing with other methods, such as the KDE, the Mixture of K Gaussians, etc., the simulation results show the proposed method has better validity and higher adaptive for moving target detection, especially in infrared image sequences with complex illumination change, noise change, and so on.

  9. Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

    International Nuclear Information System (INIS)

    Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

    1983-01-01

    The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps

  10. Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

    Energy Technology Data Exchange (ETDEWEB)

    Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

    1983-01-01

    The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps.

  11. The distribution of BRAF gene fusions in solid tumors and response to targeted therapy.

    Science.gov (United States)

    Ross, Jeffrey S; Wang, Kai; Chmielecki, Juliann; Gay, Laurie; Johnson, Adrienne; Chudnovsky, Jacob; Yelensky, Roman; Lipson, Doron; Ali, Siraj M; Elvin, Julia A; Vergilio, Jo-Anne; Roels, Steven; Miller, Vincent A; Nakamura, Brooke N; Gray, Adam; Wong, Michael K; Stephens, Philip J

    2016-02-15

    Although the BRAF V600E base substitution is an approved target for the BRAF inhibitors in melanoma, BRAF gene fusions have not been investigated as anticancer drug targets. In our study, a wide variety of tumors underwent comprehensive genomic profiling for hundreds of known cancer genes using the FoundationOne™ or FoundationOne Heme™ comprehensive genomic profiling assays. BRAF fusions involving the intact in-frame BRAF kinase domain were observed in 55 (0.3%) of 20,573 tumors, across 12 distinct tumor types, including 20 novel BRAF fusions. These comprised 29 unique 5' fusion partners, of which 31% (9) were known and 69% (20) were novel. BRAF fusions included 3% (14/531) of melanomas; 2% (15/701) of gliomas; 1.0% (3/294) of thyroid cancers; 0.3% (3/1,062) pancreatic carcinomas; 0.2% (8/4,013) nonsmall-cell lung cancers and 0.2% (4/2,154) of colorectal cancers, and were enriched in pilocytic (30%) vs. nonpilocytic gliomas (1%; p < 0.0001), Spitzoid (75%) vs. nonSpitzoid melanomas (1%; p = 0.0001), acinar (67%) vs. nonacinar pancreatic cancers (<1%; p < 0.0001) and papillary (3%) vs. nonpapillary thyroid cancers (0%; p < 0.03). Clinical responses to trametinib and sorafenib are presented. In conclusion, BRAF fusions are rare driver alterations in a wide variety of malignant neoplasms, but enriched in Spitzoid melanoma, pilocytic astrocytomas, pancreatic acinar and papillary thyroid cancers. © 2015 The Authors. Published by Wiley Periodicals, Inc. on behalf of UICC.

  12. Lawrence Livermore Laboratory heavy ion fusion program

    International Nuclear Information System (INIS)

    Bangerter, R.O.; Lee, E.P.; Monsler, M.J.; Yu, S.S.

    1978-01-01

    Target design at LLL for heavy ion fusion power production is discussed, including target development and beam-target interaction. The energy conversion chamber design, which utilizes a liquid lithium blanket, is described. Ion beam transport theory is discussed

  13. Divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

    Norajitra, P.; Ihli, T.; Janeschitz, G.; Abdel-Khalik, S.; Mazul, I.; Malang, S.

    2007-01-01

    The development of a divertor concept for post-ITER fusion power plants is deemed to be an urgent task to meet the EU Fast Track scenario. Developing a divertor is particularly challenging due to the wide range of requirements to be met including the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident a particles, radiation effects on the properties of structural materials, and efficient recovery and conversion of the divertor thermal power (∝15% of the total fusion thermal power) by maximizing the coolant operating temperature while minimizing the pumping power. In the course of the EU PPCS, three near-term (A, B and AB) and two advanced power plant models (C, D) were investigated. Model A utilizes a water-cooled lead-lithium (WCLL) blanket and a water-cooled divertor with a peak heat flux of 15 MW/m 2 . Model B uses a He-cooled ceramics/beryllium pebble bed (HCPB) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model AB uses a He-cooled lithium-lead (HCLL) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model C is based on a dual-coolant (DC) blanket (lead/lithium self-cooled bulk and He-cooled structures) and a He-cooled divertor (10 MW/m 2 ). Model D employs a self-cooled lead/lithium (SCLL) blanket and lead-lithiumcooled divertor (5 MW/m 2 ). The values in parenthesis correspond to the maximum peak heat fluxes required. It can be noted that the helium-cooled divertor is used in most of the EU plant models; it has also been proposed for the US ARIES-CS reactor study. Since 2002, it has been investigated extensively in Europe under the PPCS with the goal of reaching a maximum heat flux of at least 10 MW/m2. Work has covered many areas including conceptual design, analysis, material and fabrication issues, and experiments. Generally, the helium-cooled divertor is considered to be a suitable solution for fusion power plants, as it

  14. Fusion Engineering Device (FED) first wall/shield design

    International Nuclear Information System (INIS)

    Sager, P.H.; Fuller, G.; Cramer, B.; Davisson, J.; Haines, J.; Kirchner, J.

    1981-01-01

    The torus of the Fusion Engineering Device (FED) is comprised of the bulk shield and its associated spool lstructure and support system, the first wall water-cooled panel and armor systems, and the pumped limiter. The bulk shielding is provided by ten shield sectors that are installed in the spool structure in such a way as to permit extraction of the sectors through the openings between adjacent toroidal field coils with a direct radial movement. The first wall armor is installed on the inboard and top interior walls of these sectors, and the water-cooled panels are installed on the outboard interior walls and the pumped limiter in the bottom of the sectors. The overall design of the first wall and shield system is described in this paper

  15. Heavy ion fusion (HIF) impulse injector design, construction, and checkout

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M. J., LLNL

    1998-05-04

    The following report describes the design, construction, and checkout of a high-voltage (HV) impulser built for the heavy ion fusion (HIF) project. The purpose of this impulser is to provide an adjustable diode voltage source of sufficient quality and level to allow the optimization of beam transport and accelerator sections of HIF. An elegant, low-impedance, high-energy storage capacitor circuit has been selected for this application. A retrofit to the diode region has been included to provide additional beam stability and a controlled rise time. The critical part of this circuit that is common to all candidates is the impedance matching component. The following report provides a description of the implemented circuit, the basic circuit variables for wave shaping, component screening techniques, resulting operating parameters, diode modifications, operating considerations, and fault protection.

  16. OMEGA polar-drive target designs

    International Nuclear Information System (INIS)

    Radha, P. B.; Marozas, J. A.; Marshall, F. J.; Shvydky, A.; Collins, T. J. B.; Goncharov, V. N.; McKenty, P. W.; Sangster, T. C.; Skupsky, S.; McCrory, R. L.; Meyerhofer, D. D.

    2012-01-01

    Low-adiabat polar-drive (PD) [Skupsky et al., Phys. Plasmas 11, 2763 (2004)] implosion designs for the OMEGA [Boehly et al., Opt. Commun. 133, 495 (1997)] laser are described. These designs for cryogenic deuterium–tritium and warm plastic shells use a temporal laser pulse shape with three pickets followed by a main pulse [Goncharov et al., Phys. Rev. Lett. 104, 165001 (2010)]. The designs are at two different on-target laser intensities, with different in-flight aspect ratios (IFARs). These designs permit studies of implosion energetics and target performance closer to ignition-relevant intensities (∼7 × 10 14 W/cm 2 at the quarter-critical surface, where nonlocal heat conduction and laser–plasma interactions can play an important role) but at lower values of IFAR ∼ 22 or at lower intensity (∼3 × 10 14 W/cm 2 ) but at a higher IFAR (IFAR ∼ 32, where shell instability can play an important role). PD geometry requires repointing of laser beams to improve shell symmetry. The higher-intensity designs optimize target performance by repointing beams to a lesser extent, compensating for the reduced equatorial drive by increasing the energies of the repointed beams. They also use custom beam profiles that improve equatorial illumination at the expense of irradiation at higher latitudes. These latter designs will be studied when new phase plates for the OMEGA Laser System, corresponding to the custom beam profiles, are obtained.

  17. Inertial confinement fusion (ICF)

    International Nuclear Information System (INIS)

    Nuckolls, J.

    1977-01-01

    The principal goal of the inertial confinement fusion program is the development of a practical fusion power plant in this century. Rapid progress has been made in the four major areas of ICF--targets, drivers, fusion experiments, and reactors. High gain targets have been designed. Laser, electron beam, and heavy ion accelerator drivers appear to be feasible. Record-breaking thermonuclear conditions have been experimentally achieved. Detailed diagnostics of laser implosions have confirmed predictions of the LASNEX computer program. Experimental facilities are being planned and constructed capable of igniting high gain fusion microexplosions in the mid 1980's. A low cost long lifetime reactor design has been developed

  18. 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

  19. Conceptual design of laser fusion reactor, SENRI-I - 1. concept and system design

    International Nuclear Information System (INIS)

    Ido, S.; Naki, S.; Norimatsu, T.

    1981-01-01

    Design features of a laser fusion reactor concept SENRI-I and new concepts are reviewed and discussed. The unique feature is the utilization of a magnetic field to guide and control the inner liquid Li flow. Basic requirements and typical parameters used in the design are presented. Items to be discussed are constitution of the system, performance of liquid Li flow, neutronics, thermo-electric cycle, fuel cycle and new concepts

  20. Design and characterization of a neutralized-transport experiment for heavy-ion fusion

    Directory of Open Access Journals (Sweden)

    Enrique Henestroza

    2004-08-01

    Full Text Available In heavy-ion inertial-confinement fusion systems, intense beams of ions must be transported from the exit of the final-focus magnet system through the fusion chamber to hit spots on the target with radii of about 2 mm. For the heavy-ion-fusion power-plant scenarios presently favored in the U.S., a substantial fraction of the ion-beam space charge must be neutralized during this final transport. The most effective neutralization technique found in numerical simulations is to pass each beam through a low-density plasma after the final focusing. To provide quantitative comparisons of these theoretical predictions with experiment, the Virtual National Laboratory for Heavy Ion Fusion has completed the construction and has begun experimentation with the neutralized-transport experiment. The experiment consists of three main sections, each with its own physics issues. The injector is designed to generate a very high-brightness, space-charge-dominated potassium beam, while still allowing variable perveance by a beam aperturing technique. The magnetic-focusing section, consisting of four pulsed quadrupoles, permits the study of magnet tuning, as well as the effects of phase-space dilution due to higher-order nonlinear fields. In the final section, the converging ion beam exiting the magnetic section is transported through a drift region with plasma sources for beam neutralization, and the final spot size is measured under various conditions of neutralization. In this paper, we discuss the design and characterization of the three sections in detail and present initial results from the experiment.

  1. Design and evaluation of a laser fusion energy station for industrial applications

    International Nuclear Information System (INIS)

    Kok, K.D.; Bates, F.J.; Denning, R.S.; Triplett, M.B.; Waddell, J.D.

    1978-01-01

    The identification and development of long-term energy options is important in the continued growth of industry in the United States. Fusion and particularly laser fusion is one of the possible options. This paper applies the criteria used by industry in the selection of an energy source to the first of a series of conceptual designs for a laser fusion energy station. Several conclusions are presented including the constraints placed on the design by the criteria

  2. Inertial fusion energy power plant design using the Compact Torus Accelerator: HYLIFE-CT

    International Nuclear Information System (INIS)

    Moir, R.W.; Hammer, J.H.; Hartman, C.W.; Leber, R.L.; Logan, B.G.; Petzoldt, R.W.; Tabak, M.; Tobin, M.T.; Bieri, R.L.; Hoffman, M.A.

    1992-01-01

    The Compact Torus Accelerator (CTA), under development at Lawrence Livermore National Laboratory, offers the promise of a low-cost, high-efficiency, high energy, high-power-density driver for ICF and MICF (Magnetically Insulated ICF) type fusion systems. A CTA with 100 MJ driver capacitor bank energy is predicted to deliver ∼30 MJ CT kinetic energy to a 1 cm 2 target in several nanoseconds for a power density of ∼10 16 watts/cm 2 . The estimated cost of delivered energy is ∼3$/Joule, or $100M for 30 MJ. This driver appears to be cost-effective and, in this regard, is virtually alone among IFE drivers. We discuss indirect-drive ICF with a DT fusion energy gain Q = 70 for a total yield of 2 GJ. The CT can be guided to the target inside a several-meter-long disposable cone made of frozen Li 2 BeF 4 , the same material as the coolant. We have designed a power plant including CT injection, target emplacement, containment, energy recovery, and tritium breeding. The cost of electricity is predicted to be 4.8 cents/kWh, which is competitive with future coal and nuclear costs

  3. Identifying heavy-ion-beam fusion design and system features with high economic leverage

    International Nuclear Information System (INIS)

    Meier, W.R.; Hogan, W.J.

    1985-01-01

    In this article the authors consider a heavy-ion-beam (HIB) fusion power plant that consists of a driver, a target factory, and one or more power units. A power unit is defined as all the buildings and equipment needed to generate electric power, provided the target and beams are delivered to the reaction chamber. Because the maximum achievable pulse rate in a single chamber is limited, more than one reaction chamber may be required to achieve the desired output of a single power unit. They distinguish between multiple power units and multiple reaction chambers so that they can examine separately the effects of increasing the number of reaction chambers at a constant net power and of increasing the power level by driving more power units with a single driver. The authors conducted studies to investigate the effects on the cost of electricity (COE) of variations in several design parameters. In particular, they examined the effects of maximum achievable chamber pulse rate, driver cost, target gain, electric conversion efficiency, and net electric power. They found that with a combination of improvements over their base case, HIB fusion can be economically competitive with present and future power sources

  4. Copper-coated laser-fusion targets using molecular-beam levitation

    International Nuclear Information System (INIS)

    Rocke, M.J.

    1981-01-01

    A series of diagnostic experiments at the Shiva laser fusion facility required targets of glass microspheres coated with 1.5 to 3.0 μm of copper. Previous batch coating efforts using vibration techniques gave poor results due to microsphere sticking and vacuum welding. Molecular Beam Levitation (MBL) represented a noncontact method to produce a sputtered copper coating on a single glassmicrosphere. The coating specifications that were achieved resulted in a copper layer up to 3 μm thick with the allowance of a maximum variation of 10 nm in surface finish and thickness. These techniques developed with the MBL may be applied to sputter coat many soft metals for fusion target applications

  5. In Silico target fishing: addressing a "Big Data" problem by ligand-based similarity rankings with data fusion.

    Science.gov (United States)

    Liu, Xian; Xu, Yuan; Li, Shanshan; Wang, Yulan; Peng, Jianlong; Luo, Cheng; Luo, Xiaomin; Zheng, Mingyue; Chen, Kaixian; Jiang, Hualiang

    2014-01-01

    Ligand-based in silico target fishing can be used to identify the potential interacting target of bioactive ligands, which is useful for understanding the polypharmacology and safety profile of existing drugs. The underlying principle of the approach is that known bioactive ligands can be used as reference to predict the targets for a new compound. We tested a pipeline enabling large-scale target fishing and drug repositioning, based on simple fingerprint similarity rankings with data fusion. A large library containing 533 drug relevant targets with 179,807 active ligands was compiled, where each target was defined by its ligand set. For a given query molecule, its target profile is generated by similarity searching against the ligand sets assigned to each target, for which individual searches utilizing multiple reference structures are then fused into a single ranking list representing the potential target interaction profile of the query compound. The proposed approach was validated by 10-fold cross validation and two external tests using data from DrugBank and Therapeutic Target Database (TTD). The use of the approach was further demonstrated with some examples concerning the drug repositioning and drug side-effects prediction. The promising results suggest that the proposed method is useful for not only finding promiscuous drugs for their new usages, but also predicting some important toxic liabilities. With the rapid increasing volume and diversity of data concerning drug related targets and their ligands, the simple ligand-based target fishing approach would play an important role in assisting future drug design and discovery.

  6. Mechanical design for a large fusion laser system

    International Nuclear Information System (INIS)

    Hurley, C.A.

    1979-01-01

    The Nova Mechanical Systems Group at LLL is responsible for the design, fabrication, and installation of all laser chain components, for the stable support structure that holds them, and for the beam lines that transport the laser beam to the target system. This paper is an overview of the group's engineering effort, emphasizing new developments

  7. Inertial confinement fusion target component fabrication and technology development support: Annual report, October 1, 1997 - September 30, 1998

    International Nuclear Information System (INIS)

    Gibson, J.

    1998-12-01

    During this period, General Atomics (GA) and their partner Schafer Corporation were assigned 17 formal tasks in support of the Inertial Confinement Fusion (ICF) program and its five laboratories. A portion of the effort on these tasks included providing direct ''On-site Support'' at Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), and Sandia National Laboratory Albuquerque (SNLA). They fabricated and delivered over 1,200 hohlraum mandrels and numerous other micromachined components to LLNL, LANL, and SNLA. They produced more than 1,300 glass and plastic target capsules for LLNL, LANL, SNLA, and the University of Rochester/Laboratory for Laser Energetics (UR/LLE). They also delivered nearly 2,000 various target foils and films for Naval Research Lab (NRL) and UR/LLE in FY98. This report describes these target fabrication activities and the target fabrication and characterization development activities that made the deliveries possible. During FY98, great progress was made by the GA/Schafer-UR/LLE-LANL team in the design, procurement, installation, and testing of the OMEGA Cryogenic Target System (OCTS) that will field cryogenic targets on OMEGA. The design phase was concluded for all components of the OCTS and all major components were procured and nearly all were fabricated. Many of the components were assembled and tested, and some have been shipped to UR/LLE. The ICF program is anticipating experiments at the OMEGA laser and the National Ignition Facility (NIF) which will require targets containing cryogenic layered D 2 or deuterium-tritium (DT) fuel. They are part of the National Cryogenic Target Program and support experiments at LLNL and LANL to generate and characterize cryogenic layers for these targets. They also contributed cryogenic support and developed concepts for NIF cryogenic targets. This report summarizes and documents the technical progress made on these tasks

  8. Inertial confinement fusion target component fabrication and technology development support: Annual report, October 1, 1997--September 30, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J. [ed.

    1998-12-01

    During this period, General Atomics (GA) and their partner Schafer Corporation were assigned 17 formal tasks in support of the Inertial Confinement Fusion (ICF) program and its five laboratories. A portion of the effort on these tasks included providing direct ``On-site Support`` at Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), and Sandia National Laboratory Albuquerque (SNLA). They fabricated and delivered over 1,200 hohlraum mandrels and numerous other micromachined components to LLNL, LANL, and SNLA. They produced more than 1,300 glass and plastic target capsules for LLNL, LANL, SNLA, and the University of Rochester/Laboratory for Laser Energetics (UR/LLE). They also delivered nearly 2,000 various target foils and films for Naval Research Lab (NRL) and UR/LLE in FY98. This report describes these target fabrication activities and the target fabrication and characterization development activities that made the deliveries possible. During FY98, great progress was made by the GA/Schafer-UR/LLE-LANL team in the design, procurement, installation, and testing of the OMEGA Cryogenic Target System (OCTS) that will field cryogenic targets on OMEGA. The design phase was concluded for all components of the OCTS and all major components were procured and nearly all were fabricated. Many of the components were assembled and tested, and some have been shipped to UR/LLE. The ICF program is anticipating experiments at the OMEGA laser and the National Ignition Facility (NIF) which will require targets containing cryogenic layered D{sub 2} or deuterium-tritium (DT) fuel. They are part of the National Cryogenic Target Program and support experiments at LLNL and LANL to generate and characterize cryogenic layers for these targets. They also contributed cryogenic support and developed concepts for NIF cryogenic targets. This report summarizes and documents the technical progress made on these tasks.

  9. Minimizing scatter-losses during pre-heat for magneto-inertial fusion targets

    Science.gov (United States)

    Geissel, Matthias; Harvey-Thompson, Adam J.; Awe, Thomas J.; Bliss, David E.; Glinsky, Michael E.; Gomez, Matthew R.; Harding, Eric; Hansen, Stephanie B.; Jennings, Christopher; Kimmel, Mark W.; Knapp, Patrick; Lewis, Sean M.; Peterson, Kyle; Schollmeier, Marius; Schwarz, Jens; Shores, Jonathon E.; Slutz, Stephen A.; Sinars, Daniel B.; Smith, Ian C.; Speas, C. Shane; Vesey, Roger A.; Weis, Matthew R.; Porter, John L.

    2018-02-01

    The size, temporal and spatial shape, and energy content of a laser pulse for the pre-heat phase of magneto-inertial fusion affect the ability to penetrate the window of the laser-entrance-hole and to heat the fuel behind it. High laser intensities and dense targets are subject to laser-plasma-instabilities (LPI), which can lead to an effective loss of pre-heat energy or to pronounced heating of areas that should stay unexposed. While this problem has been the subject of many studies over the last decades, the investigated parameters were typically geared towards traditional laser driven Inertial Confinement Fusion (ICF) with densities either at 10% and above or at 1% and below the laser's critical density, electron temperatures of 3-5 keV, and laser powers near (or in excess of) 1 × 1015 W/cm2. In contrast, Magnetized Liner Inertial Fusion (MagLIF) [Slutz et al., Phys. Plasmas 17, 056303 (2010) and Slutz and Vesey, Phys. Rev. Lett. 108, 025003 (2012)] currently operates at 5% of the laser's critical density using much thicker windows (1.5-3.5 μm) than the sub-micron thick windows of traditional ICF hohlraum targets. This article describes the Pecos target area at Sandia National Laboratories using the Z-Beamlet Laser Facility [Rambo et al., Appl. Opt. 44(12), 2421 (2005)] as a platform to study laser induced pre-heat for magneto-inertial fusion targets, and the related progress for Sandia's MagLIF program. Forward and backward scattered light were measured and minimized at larger spatial scales with lower densities, temperatures, and powers compared to LPI studies available in literature.

  10. CFD aspects of ADSS target design

    International Nuclear Information System (INIS)

    Shashi Kumar, G.N.; Mahendra, A.K.; Sanyal, A.; Gouthaman, G.

    2004-03-01

    The preliminary studies on CFD aspects of Accelerator Driven Sub-critical System (ADSS) target design has been presented in this report. The studies involve the thermal hydraulic analysis of the Liquid Metal Spallation Target (LMST) using Lead Bismuth Eutectic (LBE) as the target material. Apart from acting as Spallation medium LBE is used to remove the heat deposited by High Energy Proton Beam. Window of the target ( one side vacuum and other side LBE) has been reported in literature to be the most critical zone where high temperatures are reached. Numerical Simulations are carried out with Artificial Neural Network coupled Computational Fluid Dynamics (CFD) code, Various studies were carried out after the verification and validation of the initial results. Window being, the main parameter to be optimised, various designs of window were tried, along with change in the window material. The best possible combination has been proposed. The thermal hydraulic studies were carried out to arrive at the acceptable operating conditions for the target. (author)

  11. Nuclear challenges and progress in designing stellarator fusion power plants

    International Nuclear Information System (INIS)

    El-Guebaly, L.A.; Wilson, P.; Henderson, D.; Sawan, M.; Sviatoslavsky, G.; Tautges, T.; Slaybaugh, R.; Kiedrowski, B.; Ibrahim, A.

    2008-01-01

    Over the past 5-6 decades, stellarator power plants have been studied in the US, Europe, and Japan as an alternate to the mainline magnetic fusion tokamaks, offering steady-state operation and eliminating the risk of plasma disruptions. The earlier 1980s studies suggested large-scale stellarator power plants with an average major radius exceeding 20 m. The most recent development of the compact stellarator concept delivered ARIES-CS - a compact stellarator with 7.75 m average major radius, approaching that of tokamaks. For stellarators, the most important engineering parameter that determines the machine size and cost is the minimum distance between the plasma boundary and mid-coil. Accommodating the breeding blanket and necessary shield within this distance to protect the ARIES-CS superconducting magnet represents a challenging task. Selecting the ARIES-CS nuclear and engineering parameters to produce an economic optimum, modeling the complex geometry for 3D nuclear analysis to confirm the key parameters, and minimizing the radwaste stream received considerable attention during the design process. These engineering design elements combined with advanced physics helped enable the compact stellarator to be a viable concept. This paper provides a brief historical overview of the progress in designing stellarator power plants and a perspective on the successful integration of the nuclear activity into the final ARIES-CS configuration

  12. Remote maintenance design for Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Tachikawa, K.; Iida, H.; Nishio, S.; Tone, T.; Aota, T.; Iwamoto, T.; Niikura, S.; Nishizawa, H.

    1984-01-01

    Design of Fusion Experimental Reactor, FER, has been conducted by Japan Atomic Energy Research Institute (JAERI) since 1981. Two typical reactors can be classified in general from the viewpoints of remote maintenance among four design concepts of FER. In the case of the type 1 FER, the torus module consists of shield structure and blanket, and the connective joints between toruses provided at the outer region of the reactor. As for the type 2 FER, the shield structure is joined with the vacuum cryostat, and only the blanket module is allowed to move, but connection between toruses are located in the inner region of the reactor. Comparing type 1 with type 2 FER, this paper describes on the remote maintenance of FER including reactor configurations, work procedures, remote systems/equipments, repairing facility and future R and D problems. Reviewing design studies and investigation for the existing robotics technologies, R and D for FER remote maintenance technology should be performed under the reasonable long-term program. The main items of remote technology required to start urgently are multi-purpose manipulator system with performance of dextrousity, tele-viewing system which reduces operator fatigue and remote tests for commercially available components

  13. Applying design principles to fusion reactor configurations for propulsion in space

    International Nuclear Information System (INIS)

    Carpenter, S.A.; Deveny, M.E.; Schulze, N.R.

    1993-01-01

    The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. Three design principles (DP's) were applied to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. A preliminary rating of these configurations was performed, and it was concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS)

  14. Data fusion for target tracking and classification with wireless sensor network

    Science.gov (United States)

    Pannetier, Benjamin; Doumerc, Robin; Moras, Julien; Dezert, Jean; Canevet, Loic

    2016-10-01

    In this paper, we address the problem of multiple ground target tracking and classification with information obtained from a unattended wireless sensor network. A multiple target tracking (MTT) algorithm, taking into account road and vegetation information, is proposed based on a centralized architecture. One of the key issue is how to adapt classical MTT approach to satisfy embedded processing. Based on track statistics, the classification algorithm uses estimated location, velocity and acceleration to help to classify targets. The algorithms enables tracking human and vehicles driving both on and off road. We integrate road or trail width and vegetation cover, as constraints in target motion models to improve performance of tracking under constraint with classification fusion. Our algorithm also presents different dynamic models, to palliate the maneuvers of targets. The tracking and classification algorithms are integrated into an operational platform (the fusion node). In order to handle realistic ground target tracking scenarios, we use an autonomous smart computer deposited in the surveillance area. After the calibration step of the heterogeneous sensor network, our system is able to handle real data from a wireless ground sensor network. The performance of system is evaluated in a real exercise for intelligence operation ("hunter hunt" scenario).

  15. Prostate cancer diagnosis: Efficacy of a simple electromagnetic MRI-TRUS fusion method to target biopsies

    International Nuclear Information System (INIS)

    Jelidi, Amina; Ohana, Mickael; Labani, Aïssam; Alemann, Guillaume; Lang, Herve; Roy, Catherine

    2017-01-01

    Highlights: • A very simple electromagnetic device for fusion with MRI examination during TRUS guided biopsies increases the detection of clinically significant prostate cancer. • This device has advantages: a short time for the fusion registration, no additional cumbersome material and no intense training to be fluent with. • Low or intermediate suspicious area for prostate carcinoma on mpMRI can be due to benign histological abnormalities or high grade prostatic intraepithelial neoplasia. - Abstract: Objective: To assess that transrectal ultrasound guidance (TRUS) targeted biopsies (TB) aimed with an easy to use electronic real-time fusion registration device have a higher rate of prostate cancer (PC) detection than standard biopsies (SB). Material and methods: This prospective study included 130 patients referred for TRUS biopsies after suspicious MRI. They underwent 16-core SB and 2 to 3 cores in each MRI suspicious area, using a fusion software. We noted SB and TB positivity for PC and Gleason score (GS). We used the McNemar test to compare SB and TB, with a statistical significance p < 0.05. Results: Among 130 patients, 68.5% had PC. Additional time due to the fusion registration was 3.3 min. One hundred fifteen patients (88.4%) had pathological findings on the histological analysis (prostate cancer n = 89, others n = 26). TB diagnosed PC in 75 patients with negative SB. Positivity rate for PC was significantly higher for TB than SB (p = 0.03). Among highly suspicious MRI lesions, detection rate of histological abnormalities using SB and TB was 96% with 79.7% of PC. Most PC that TB diagnosed alone were clinically significant (86.3%). Conclusion: TRUS biopsies performed with a simple MRI and US electronic fusion is an unrestrainedly method to increase PC diagnosis.

  16. Recent US target-physics-related research in heavy-ion inertial fusion: simulations for tamped targets and for disk experiments in accelerator test facilities

    International Nuclear Information System (INIS)

    Mark, J.W.K.

    1982-01-01

    Within the last few years, there have also appeared in the Heavy-Ion Fusion literature several studies of targets which have outer tampers. One-dimensional simulations indicate higher target gains with a judicious amount of tamping. But for these targets, a full investigation has not been carried through in regards to conservative criteria for fluid instabilities as well as reasonable imperfections in target fabrication and illumination symmetry which all affect target ignition and burn. Comparisons of these results with the gain survey of Part I would have to be performed with care. These calculations suggest that experiments relating to high temperature disk heating, as well as beam deposition, focusing and transport can be performed within the context of current design proposals for accelerator test-facilities. Since the test-facilities have lower ion kinetic energy and beam pulse power as compared to reactor drivers, we achieve high-beam intensities at the focal spot by using short focal distance and properly designed beam optics

  17. Test of a High Power Target Design

    CERN Multimedia

    2002-01-01

    %IS343 :\\\\ \\\\ A high power tantalum disc-foil target (RIST) has been developed for the proposed radioactive beam facility, SIRIUS, at the Rutherford Appleton Laboratory. The yield and release characteristics of the RIST target design have been measured at ISOLDE. The results indicate that the yields are at least as good as the best ISOLDE roll-foil targets and that the release curves are significantly faster in most cases. Both targets use 20 -25 $\\mu$m thick foils, but in a different internal geometry.\\\\ \\\\Investigations have continued at ISOLDE with targets having different foil thickness and internal geometries in an attempt to understand the release mechanisms and in particular to maximise the yield of short lived isotopes. A theoretical model has been developed which fits the release curves and gives physical values of the diffusion constants.\\\\ \\\\The latest target is constructed from 2 $\\mu$m thick tantalum foils (mass only 10 mg) and shows very short release times. The yield of $^{11}$Li (half-life of ...

  18. Safety and environmental requirements and design targets for TIBER-II

    International Nuclear Information System (INIS)

    Piet, S.J.

    1987-09-01

    A consistent set of safety and environmental requirements and design targets was proposed and adopted for the TIBER-II (Tokamak Ignition/Burn Experimental Reactor) design effort. TIBER-II is the most recent US version of a fusion experimental test reactor (ETR). These safety and environmental design targets were one contribution of the Fusion Safety Program in the TIBER-II design effort. The other contribution, safety analyses, is documented in the TIBER-II design report. The TIBER-II approach, described here, concentrated on logical development of, first, a complete and consistent set of safety and environmental requirements that are likely appropriate for an ETR, and, second, an initial set of design targets to guide TIBER-II. Because of limited time in the TIBER-II design effort, the iterative process only included one iteration - one set of targets and one design. Future ETR design efforts should therefore build on these design targets and the associated safety analyses. 29 refs., 5 figs., 3 tabs

  19. Chemical Hazards and Safety Issues in Fusion Safety Design

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    2003-01-01

    Radiological inventory releases have dominated accident consequences for fusion; these consequences are important to analyze and are generally the most severe result of a fusion facility accident event. However, the advent of, or plan for, large-scale usage of some toxic materials poses the additional hazard of chemical exposure from an accident event. Examples of toxic chemicals are beryllium for magnetic fusion and fluorine for laser fusion. Therefore, chemical exposure consequences must also be addressed in fusion safety assessment. This paper provides guidance for fusion safety analysis. US Department of Energy (DOE) chemical safety assessment practices for workers and the public are reviewed. The US Environmental Protection Agency (EPA) has published some guidance on public exposure to releases of mixtures of chemicals, this guidance has been used to create an initial guideline for treating mixed radiological and toxicological releases in fusion; for example, tritiated hazardous dust from a tokamak vacuum vessel. There is no convenient means to judge the hazard severity of exposure to mixed materials. The chemical fate of mixed material constituents must be reviewed to determine if there is a separate or combined radiological and toxicological carcinogenesis, or if other health threats exist with radiological carcinogenesis. Recommendations are made for fusion facility chemical safety evaluation and safety guidance for protecting the public from chemical releases, since such levels are not specifically identified in the DOE fusion safety standard

  20. Peptide-Based Membrane Fusion Inhibitors Targeting HCoV-229E Spike Protein HR1 and HR2 Domains

    Directory of Open Access Journals (Sweden)

    Shuai Xia

    2018-02-01

    Full Text Available Human coronavirus 229E (HCoV-229E infection in infants, elderly people, and immunocompromised patients can cause severe disease, thus calling for the development of effective and safe therapeutics to treat it. Here we reported the design, synthesis and characterization of two peptide-based membrane fusion inhibitors targeting HCoV-229E spike protein heptad repeat 1 (HR1 and heptad repeat 2 (HR2 domains, 229E-HR1P and 229E-HR2P, respectively. We found that 229E-HR1P and 229E-HR2P could interact to form a stable six-helix bundle and inhibit HCoV-229E spike protein-mediated cell-cell fusion with IC50 of 5.7 and 0.3 µM, respectively. 229E-HR2P effectively inhibited pseudotyped and live HCoV-229E infection with IC50 of 0.5 and 1.7 µM, respectively. In a mouse model, 229E-HR2P administered intranasally could widely distribute in the upper and lower respiratory tracts and maintain its fusion-inhibitory activity. Therefore, 229E-HR2P is a promising candidate for further development as an antiviral agent for the treatment and prevention of HCoV-229E infection.

  1. Conceptual design studies of experimental and demonstration fusion reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

  2. Muon catalyzed fusion at very low temperature: A new target system

    International Nuclear Information System (INIS)

    Mulhauser, F.; Beveridge, J.L.; Marshall, G.M.

    1994-10-01

    Muon catalyzed fusion (μCF) processes are usually studied in gases or liquids. A new target system allows experiments on muonic hydrogen isotopes in solid hydrogen layers at 3K, where processes of the μCF cycle can be separated and the energy dependence of reactions can be measured. Muonic tritium atomic beams with energy of the order of 1 eV have been produced via transfer and emission from solid hydrogen target containing small tritium concentrations. The μt energy distribution overlaps the predicted muonic molecular (dμt) formation resonances. Preliminary time of flight results are shown. (author). 9 refs., 5 figs

  3. Information-management data base for fusion-target fabrication processes

    International Nuclear Information System (INIS)

    Reynolds, J.

    1982-01-01

    A computer-based data-management system has been developed to handle data associated with target-fabrication processes including glass microballoon characterization, gas filling, materials coating, and storage locations. The system provides automatic data storage and computation, flexible data-entry procedures, fast access, automated report generation, and secure data transfer. It resides on a CDC CYBER 175 computer and is compatible with the CDC data-base-language Query Update, but is based on custom FORTRAN software interacting directly with the CYBER's file-management system. The described data base maintains detailed, accurate, and readily available records of fusion targets information

  4. Calculation of fusion gain in fast ignition with magnetic target by relativistic electrons and protons

    Directory of Open Access Journals (Sweden)

    A Parvazian

    2010-12-01

    Full Text Available Fast ignition is a new method for inertial confinement fusion (ICF in which the compression and ignition steps are separated. In the first stage, fuel is compressed by laser or ion beams. In the second phase, relativistic electrons are generated by pettawat laser in the fuel. Also, in the second phase 5-35 MeV protons can be generated in the fuel. Electrons or protons can penetrate in to the ultra-dense fuel and deposit their energy in the fuel . More recently, cylindrical rather than spherical fuel chambers with magnetic control in the plasma domain have been also considered. This is called magnetized target fusion (MTF. Magnetic field has effects on relativistic electrons energy deposition rate in fuel. In this work, fast ignition method in cylindrical fuel chambers is investigated and transportation of the relativistic electrons and protons is calculated using MCNPX and FLUKA codes with 0. 25 and 0. 5 tesla magnetic field in single and dual hot spot. Furthermore, the transfer rate of relativistic electrons and high energy protons to the fuel and fusion gain are calculated. The results show that the presence of external magnetic field guarantees higher fusion gain, and relativistic electrons are much more appropriate objects for ignition. MTF in dual hot spot can be considered as an appropriate substitution for the current ICF techniques.

  5. Automated computer analysis of x-ray radiographs greatly facilitates measurement of coating-thickness variations in laser-fusion targets

    International Nuclear Information System (INIS)

    Stupin, D.M.; Moore, K.R.; Thomas, G.D.; Whitman, R.L.

    1981-01-01

    An automated system was built to analyze x-ray radiographs of laser fusion targets which greatly facilitates the detection of coating thickness variations. Many laser fusion targets reqire opaque coatings 1 to 20 μm thick which have been deposited on small glass balloons 100 to 500 μm in diameter. These coatings must be uniformly thick to 1% for the targets to perform optimally. Our system is designed to detect variations as small as 100 A in 1-μm-thick coatings by converting the optical density variations of contact x-ray radiographs into coating thickness variations. Radiographic images are recorded in HRP emulsions and magnified by an optical microscope, imaged onto television camera, digitized and processed on a Data General S/230 computer with a code by Whitman. After an initial set-up by the operator, as many as 200 targets will be automatically characterized

  6. Safety design of the international fusion materials irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Yamaki, Daiju; Katsuta, Hiroji; Moeslang, Anton; Jameson, R.A.; Martone, Marcello; Shannon, T.E.

    1997-11-01

    In the Conceptual Design Activity of the IFMIF, major subsystems, as well as the entire facility is carefully designed to satisfy the safety requirements for any possible construction sites. Each subsystem is qualitatively analyzed to identify possible hazards to the workers, public and environments using Failure Mode and Effect Analysis (FMEA). The results are reflected in the design and operation procedure. Shielding of radiation, particularly neutron around the test cell is one of the most important issue in normal operation. Radiation due to beam halo and activation is a hazard for operation personnel in the accelerator system. For the maintenance, remote handling technology is designed to be applied in various facilities of the IFMIF. Lithium loop and target system hold the majority of the radioactive material in the facility. Tritium and beryllium-7 are generated by the nuclear reaction during operation and thus needed to be removed continuously. They are also the potential hazards of airborne source in off-normal events. Minimization of inventory, separation and immobilization, and multiple confinement are considered in the design. Generation of radioactive waste is anticipated to be minor, but waste treatment systems for gas, liquid and solid wastes are designed to minimize the environmental impact. Lithium leak followed by a fire is a major concern, and extensive prevention plan is made in the target design. One of the design option considered is composed of; primary enclosure of the lithium loop, secondary containment filled with positive pressure argon, and an air tight lithium cell made of concrete with a steel lining. This study will report some technical issues considered in the design of IFMIF. It was concluded that the IFMIF can be designed and constructed to meet or exceed current safely standards for workers, public and the environment with existing technology and reasonable construction cost. (J.P.N.)

  7. Cooling System Design Options for a Fusion Reactor

    Science.gov (United States)

    Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

    1997-06-01

    The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

  8. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    Nakashima, Kunihiko; Okano, Kunihiko; Miyamoto, Kazuhiro.

    1987-09-01

    This report describes the results of a conceptual study on the RF system in the typical candidates for the Fusion Experimental Reactor (FER), which were picked out through the '86FER scoping studies. According to the FER operation scenario, three RF systems, that is, ICRF (heating), LHRF (current drive and heating), ECRF (auxiliary heating) were studied. Main concern in these RF systems is the launcher, which may be so designed that required power match the geometrical constraints of the reactor. Then studies were concentrated on the launcher configuration. A prug-in concept of the launcher was adopted in each system and vacancies except transmission space were filled with water. The ICRF launcher had the 2 x 2 loop arrays antenna and the faraday shield area of 1.5 m x 1 m to provide a power of 20 MW. The LHRF launcher had the grillantenna with 28 x 8 open waveguides, and included multi junction-type power splitters which were connected to 56 transmission wave guides. The grild was designed to have two functions of current drive and heating, and provide a power of 20 MW each. The ECRF launcher had a boundle of open wave guides which a reflection mirror each, and three plain mirrors. Assuming a oscillator unit size of 200 kW, it had 40 oversized wave guides to provide a power of 3 MW. (author)

  9. Development of design system of manipulator for fusion reactor maintenance

    International Nuclear Information System (INIS)

    Ida, Toshio; Niikura, Setsuo; Ishiguro, Akiko; Yamada, Masao; Matsuoka, Fushiki

    1989-01-01

    A program which supports designers of a manipulator for in-vessel maintenance of a fusion reactor has been developed. The main purpose of this program is to provide the designer with a promising manipulator specification by furnishing useful information. It combines a technique of knowledge engineering with numerical solutions. This program consists of three parts: The first part is to generate candidates for the manipulator using a knowledge base; the second is to evaluate both static and dynamic properties of each candidate through numerical simulation of the maintenance task; and the third is to select the candidates having better performance and feasibility on the basis of the simulation data and knowledge base. The feasibility of the method used in this program is confirmed by the preliminary application. This application also emphasizes the importance of the knowledge base for the candidate generation and selection. Although the degree of freedom of the manipulator is restricted to less than 7 degrees in this study, further development of the capability of manipulator simulation (seven or more degrees of freedom) will enhance the effectiveness of this program. (orig.)

  10. Studies on the robustness of shock-ignited laser fusion targets

    International Nuclear Information System (INIS)

    Atzeni, S; Schiavi, A; Marocchino, A

    2011-01-01

    Several aspects of the sensitivity of a shock-ignited inertial fusion target to variation of parameters and errors or imperfections are studied by means of one-dimensional and two-dimensional numerical simulations. The study refers to a simple all-DT target, initially proposed for fast ignition (Atzeni et al 2007 Phys. Plasmas 7 052702) and subsequently shown to be also suitable for shock ignition (Ribeyre et al 2009 Plasma Phys. Control. Fusion 51 015013). It is shown that the growth of both Richtmyer-Meshkov and Rayleigh-Taylor instability (RTI) at the ablation front is reduced by laser pulses with an adiabat-shaping picket. An operating window for the parameters of the ignition laser spike is described; the threshold power depends on beam focusing and synchronization with the compression pulse. The time window for spike launch widens with beam power, while the minimum spike energy is independent of spike power. A large parametric scan indicates good tolerance (at the level of a few percent) to target mass and laser power errors. 2D simulations indicate that the strong igniting shock wave plays an important role in reducing deceleration-phase RTI growth. Instead, the high hot-spot convergence ratio (ratio of initial target radius to hot-spot radius at ignition) makes ignition highly sensitive to target mispositioning.

  11. Evaluation of Cytochalasin B-Induced Membrane Vesicles Fusion Specificity with Target Cells

    Directory of Open Access Journals (Sweden)

    Marina Gomzikova

    2018-01-01

    Full Text Available Extracellular vesicles (EV represent a promising vector system for biomolecules and drug delivery due to their natural origin and participation in intercellular communication. As the quantity of EVs is limited, it was proposed to induce the release of membrane vesicles from the surface of human cells by treatment with cytochalasin B. Cytochalasin B-induced membrane vesicles (CIMVs were successfully tested as a vector for delivery of dye, nanoparticles, and a chemotherapeutic. However, it remained unclear whether CIMVs possess fusion specificity with target cells and thus might be used for more targeted delivery of therapeutics. To answer this question, CIMVs were obtained from human prostate cancer PC3 cells. The diameter of obtained CIMVs was 962,13 ± 140,6 nm. We found that there is no statistically significant preference in PC3 CIMVs fusion with target cells of the same type. According to our observations, the greatest impact on CIMVs entry into target cells is by the heterophilic interaction of CIMV membrane receptors with the surface proteins of target cells.

  12. Conceptual design of the fusion-driven subcritical system FDS-I

    International Nuclear Information System (INIS)

    Wu, Y.; Zheng, S.; Zhu, X.; Wang, W.; Wang, H.; Liu, S.; Bai, Y.; Chen, H.; Hu, L.; Chen, M.; Huang, Q.; Huang, D.; Zhang, S.; Li, J.; Chu, D.; Jiang, J.; Song, Y.

    2006-01-01

    The fusion-driven subcritical system (named FDS-I) was previously proposed as an intermediate step toward the final application of fusion energy. A conceptual design of the FDS-I is presented, which consists of the fusion neutron driver with relatively easy-achieved plasma parameters, and the He-gas/liquid lithium-lead Dual-cooled subcritical Waste Transmutation (DWT) blanket used to transmute long-lived radioactive wastes and to generate energy on the basis of self-sustainable fission and fusion fuel cycle. An overview of the FDS-I is given and the specifications of the design analysis are summarized

  13. Updated reference design of a liquid metal cooled tandem mirror fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.; Gromada, R.J.; McCarville, T.J.; Moir, R.W.; Lee, J.D.; Bandini, B.R.; Fulton, F.J.; Wong, C.P.C.; Maya, I.; Hoot, C.G.; Schultz, K.R.; Miller, L.G.; Beeston, J.M.; Harris, B.L.; Westman, R.A.; Ghoniem, N.M.; Orient, G.; Wolfer, M.; DeVan, J.H.; Torterelli, P.

    1985-09-01

    Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future.

  14. Updated reference design of a liquid metal cooled tandem mirror fusion breeder

    International Nuclear Information System (INIS)

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.

    1985-09-01

    Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future

  15. Targeting and design of chilled water network

    International Nuclear Information System (INIS)

    Foo, Dominic C.Y.; Ng, Denny K.S.; Leong, Malwynn K.Y.; Chew, Irene M.L.; Subramaniam, Mahendran; Aziz, Ramlan; Lee, Jui-Yuan

    2014-01-01

    Highlights: • Minimum flowrate targeting for chilled water network. • Mixed series/parallel configuration of chilled water-using units. • Integrated cooling and chilled water networks. - Abstract: Chilled water is a common cooling agent used in various industrial, commercial and institutional facilities. In conventional practice, chilled water is distributed via chilled water networks (CHWNs) in parallel configuration to provide required air conditioning and/or equipment cooling in the heating, ventilating and air conditioning (HVAC) system. In this paper, process integration approach based on pinch analysis technique is used to address energy efficiency issues in the CHWN system for grassroots design problem. Graphical and algebraic targeting techniques are developed to identify the minimum chilled water flowrate needed to remove a given amount of heat load from the CHWN. Doing this leads to higher chilled water return temperature and enhanced energy efficiency of the HVAC system. A recent proposed network design technique is extended to synthesize the CHWN in a mixed series/parallel configuration. A novel concept of integrated cooling and chilled water networks (IWN) is also proposed in this work, with its targeting and design techniques presented. Hypothetical examples and an industrial case study are solved to elucidate the proposed approaches

  16. Verification of the hydraulic design of the FMIT liquid lithium target

    International Nuclear Information System (INIS)

    Miles, R.R.; Annese, C.E.; Ingham, J.G.

    1983-01-01

    A liquid lithium target is being developed to generate a neutron flux for material testing in a fusion-like environment. The target consists of a thin, high speed, curved wall jet of lithium which is formed by an asymmetric nozzle. A prototype target was designed using potential flow analysis and was tested in water. Measurements of jet thickness and velocity in water and thickness in lithium were compared with isothermal design predictions and were shown to match within 1% for thickness and 5% for jet velocity

  17. Cryogenic-laser-fusion target implosion studies performed with the OMEGA uv-laser system

    International Nuclear Information System (INIS)

    Marshall, F.J.; Letzring, S.A.; Verdon, C.P.; Skupsky, S.; Keck, R.L.; Knauer, J.P.; Kremens, R.L.; Bradley, D.K.; Kessler, T.; Delettrez, J.; and others.

    1989-01-01

    A series of direct-drive laser-fusion implosion experiments was performed on cryogenically cooled, DT-filled glass microballoons with the OMEGA 24-beam uv (351-nm) laser system. The targets consisted of glass microballoons having radii of 100 to 150 μm, wall thicknesses of 3 to 7 μm, filled with DT gas at pressures of 75 to 100 atm. The targets were cooled to below the freezing point of DT, in situ, by a cryogenic target system. The targets were irradiated by approximately 1 to 1.2 kJ of uv light in 650-ps Gaussian pulses. The on-target irradiation uniformity was enhanced for these experiments by the use of distributed phase plates, which brought the estimated irradiation nonuniformities to ∼12% (σ rms ). Target performance was diagnosed by an array of x-ray, plasma, and nuclear instruments. The measured target performance showed ∼70% absorption, thermonuclear yields of 10 6 to 10 8 neutrons, and final fuel areal densities of 20 to 40 mg/cm 2 for the optimum targets examined in these experiments. Fuel densities at the time of thermonuclear neutron production, inferred from direct measurements of the fuel areal density, were in the range of 20 to 50 g/cm 3 (100 to 200 times the density of liquid DT) for the optimum targets

  18. IFMIF target and test cell - design and integration

    International Nuclear Information System (INIS)

    Heinzel, V.

    2007-01-01

    The International Fusion Material Irradiation Facility (IFMIF) aims at the qualification of appropriate materials for a Demonstration Fusion Power Plant (DEMO) to a fluence of up to 150 dpa (displacement per atom) at a DEMO typical neutron spectrum. It comprises two accelerators each providing a deuteron beam with 125 mA and 40 MeV. The deuterons strike a lithium target and create via stripping reactions neutrons. The neutrons are mainly forward directed into the High-Flux-Test-Module (HFTM). The Medium Flux-Test-Modules (MFTM) and the Low-Flux-Test-Modules (LFTM) are arranged in beam direction behind. In the HFTM a damage rate in steel of more than 20 dpa/fpy (displacement per atome per full power year) will be provide in a volume of 0.5 litre. The neutron spectrum is prone to produce helium and tritium in steel like in the first wall of a DEMO reactor. The Medium- Flux-Test-Modules are designed for creep fatigues in situ and tritium release test. The test modules are cooled with helium. The target is a lithium jet with a free surface towards the deuteron beams. The jet follows a concave curved so called back wall. Centrifugal forces increase the static pressure, which prevents lithium boiling at the beam tube pressure and the power release of 10 MW due to the deuteron beams. The target and Test Cell (TTC) houses the target and the test modules as well as the lithium supply tubes and a quench tank into which the lithium splashes after the target. The lithium containing components have a temperature of 250 to 350 C. Nuclear reactions mainly in beam direction contribute to heat releases in TTC components. The TTC is filled with a noble gas with almost atmospheric pressure. Natural convection transfers heat to the walls but also mitigates temperature peaks. The Forschungszentrum Karlsruhe (FZK) has developed or validated tools for: - The extended Monte Carlo Code McDeLicious for calculations of the neutron source term, dpa rates in the material specimens, activation

  19. Design descriptions of the Prometheus-L and -H inertial fusion energy drivers

    Energy Technology Data Exchange (ETDEWEB)

    Linford, G.J. (TRW Inc., Bld. O1/1220, Redondo Beach, CA 90278 (United States)); Driemeyer, D.E. (McDonnell Douglas Aerospace Co. (MDAC), St. Louis, MO 63166 (United States)); Fornaca, S.W. (TRW Inc., Bld. O1/1220, Redondo Beach, CA 90278 (United States)); Maschke, A.W. (TRW Inc., Bld. O1/1220, Redondo Beach, CA 90278 (United States))

    1994-08-01

    Two innovative drivers have been designed for a prototype 1000MW thermonuclear power plant planned for operation early in the next century. The Prometheus-L driver is a 4MJ KrF master oscillator power amplifier laser system designed to operate at a 5.6Hz repetition rate. Output pulses from the KrF master oscillator are synchronized with the pulsed-power excitation of the KrF power amplifiers and the launching of the inertial fusion energy deuterium/tritium targets. The Prometheus-L laser architecture features 960 5kJ electric discharge KrF power amplifiers pumping 60 crossed stimulated rotational Raman scattering H[sub 2] amplifiers serving as beam accumulators. Pulse compression of the 60 accumulator beams is accomplished in 60 chirped, self-seeded SF[sub 6] stimulated Brillouin scattering pulse compressors. Grazing incidence metal focusing mirrors minimize back-streaming radiation damage from the target chamber. This architecture permits the laser driver to deliver spectrally broad-band, temporally complex optical pulses in 60 beam lines to implode the direct-drive IFE targets within a 5m radius target chamber.The Prometheus-H driver is a 7.8MJ 4GeV Pb[sup ++] heavy ion (HI) inertial fusion energy system designed to operate at a 3.5He repetition rate. The HI driver design is based on a short, ramped gradient, 5MeV accelerator, followed by a longer, 2km constant gradient, single beam linear accelerator operated in a 50kHz burst mode to generate sequentially 18 4GeV beamlets. A two-sided irradiation geometry was developed for indirect-drive HI targets. Six beamlets are used for the 45ns precursor HI pulses stored in two superconducting storage rings, 12 superconducting storage rings accumulate the 12 main beamlets, with a final buncher generating the 8ns HI pulses which arrive at the target chamber simultaneously. Final focusing is accomplished with large aperture triplet focusing magnets through Pb-vapor neutralization cells to reduce the effect of space charge.

  20. Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

    International Nuclear Information System (INIS)

    Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

    1975-01-01

    A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

  1. Review of fission-fusion pellet designs and inertial confinement system studies at EIR

    Energy Technology Data Exchange (ETDEWEB)

    Seifriz, W [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1978-01-01

    The article summarizes the work done so far at the Swiss Federal Institute for Reactor Research (EIR) in the field of the inertial confinement fusion technique. The following subjects are reviewed: a) fission fusion pellet designs using fissionable triggers, b) uranium tampered pellets, c) tampered pellets recycling unwanted actinide wastes from fission reactors in beam-driven micro-explosion reactors, and d) symbiotic fusion/fission reactor studies.

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

    Science.gov (United States)

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

    2016-10-01

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

  3. Improved prediction of drug-target interactions using regularized least squares integrating with kernel fusion technique

    Energy Technology Data Exchange (ETDEWEB)

    Hao, Ming; Wang, Yanli, E-mail: ywang@ncbi.nlm.nih.gov; Bryant, Stephen H., E-mail: bryant@ncbi.nlm.nih.gov

    2016-02-25

    Identification of drug-target interactions (DTI) is a central task in drug discovery processes. In this work, a simple but effective regularized least squares integrating with nonlinear kernel fusion (RLS-KF) algorithm is proposed to perform DTI predictions. Using benchmark DTI datasets, our proposed algorithm achieves the state-of-the-art results with area under precision–recall curve (AUPR) of 0.915, 0.925, 0.853 and 0.909 for enzymes, ion channels (IC), G protein-coupled receptors (GPCR) and nuclear receptors (NR) based on 10 fold cross-validation. The performance can further be improved by using a recalculated kernel matrix, especially for the small set of nuclear receptors with AUPR of 0.945. Importantly, most of the top ranked interaction predictions can be validated by experimental data reported in the literature, bioassay results in the PubChem BioAssay database, as well as other previous studies. Our analysis suggests that the proposed RLS-KF is helpful for studying DTI, drug repositioning as well as polypharmacology, and may help to accelerate drug discovery by identifying novel drug targets. - Graphical abstract: Flowchart of the proposed RLS-KF algorithm for drug-target interaction predictions. - Highlights: • A nonlinear kernel fusion algorithm is proposed to perform drug-target interaction predictions. • Performance can further be improved by using the recalculated kernel. • Top predictions can be validated by experimental data.

  4. Improved prediction of drug-target interactions using regularized least squares integrating with kernel fusion technique

    International Nuclear Information System (INIS)

    Hao, Ming; Wang, Yanli; Bryant, Stephen H.

    2016-01-01

    Identification of drug-target interactions (DTI) is a central task in drug discovery processes. In this work, a simple but effective regularized least squares integrating with nonlinear kernel fusion (RLS-KF) algorithm is proposed to perform DTI predictions. Using benchmark DTI datasets, our proposed algorithm achieves the state-of-the-art results with area under precision–recall curve (AUPR) of 0.915, 0.925, 0.853 and 0.909 for enzymes, ion channels (IC), G protein-coupled receptors (GPCR) and nuclear receptors (NR) based on 10 fold cross-validation. The performance can further be improved by using a recalculated kernel matrix, especially for the small set of nuclear receptors with AUPR of 0.945. Importantly, most of the top ranked interaction predictions can be validated by experimental data reported in the literature, bioassay results in the PubChem BioAssay database, as well as other previous studies. Our analysis suggests that the proposed RLS-KF is helpful for studying DTI, drug repositioning as well as polypharmacology, and may help to accelerate drug discovery by identifying novel drug targets. - Graphical abstract: Flowchart of the proposed RLS-KF algorithm for drug-target interaction predictions. - Highlights: • A nonlinear kernel fusion algorithm is proposed to perform drug-target interaction predictions. • Performance can further be improved by using the recalculated kernel. • Top predictions can be validated by experimental data.

  5. Enhanced neutralization potency of botulinum neurotoxin antibodies using a red blood cell-targeting fusion protein.

    Directory of Open Access Journals (Sweden)

    Sharad P Adekar

    2011-03-01

    Full Text Available Botulinum neurotoxin (BoNT potently inhibits cholinergic signaling at the neuromuscular junction. The ideal countermeasures for BoNT exposure are monoclonal antibodies or BoNT antisera, which form BoNT-containing immune complexes that are rapidly cleared from the general circulation. Clearance of opsonized toxins may involve complement receptor-mediated immunoadherence to red blood cells (RBC in primates or to platelets in rodents. Methods of enhancing immunoadherence of BoNT-specific antibodies may increase their potency in vivo. We designed a novel fusion protein (FP to link biotinylated molecules to glycophorin A (GPA on the RBC surface. The FP consists of an scFv specific for murine GPA fused to streptavidin. FP:mAb:BoNT complexes bound specifically to the RBC surface in vitro. In a mouse model of BoNT neutralization, the FP increased the potency of single and double antibody combinations in BoNT neutralization. A combination of two antibodies with the FP gave complete neutralization of 5,000 LD50 BoNT in mice. Neutralization in vivo was dependent on biotinylation of both antibodies and correlated with a reduction of plasma BoNT levels. In a post-exposure model of intoxication, FP:mAb complexes gave complete protection from a lethal BoNT/A1 dose when administered within 2 hours of toxin exposure. In a pre-exposure prophylaxis model, mice were fully protected for 72 hours following administration of the FP:mAb complex. These results demonstrate that RBC-targeted immunoadherence through the FP is a potent enhancer of BoNT neutralization by antibodies in vivo.

  6. Conceptual design study of fusion experimental reactor (FY86FER)

    International Nuclear Information System (INIS)

    Nakashima, Kunihiko; Yamamoto, Shin; Ohara, Yoshihiro; Watanabe, Kazuhiro; Mizuno, Makoto; Araki, Masanori; Uede, Taisei; Okano, Kunihiko.

    1987-09-01

    This report describes the results of applicability studies for the negative ion-based neutral beam injector to the Fusion Experimental Reactor (FER). The operation scenario of FER has been proposed to adopt the neutral injection method as one of candidates, which has three functions of heating, current drive and profile control. One of the fundamental requirements is the tangential injection of the neutral beam. For neutral beam injectors, three port sections are available. Supposing to adopt the beam line with the straight long neutralizer which has been designed in JAERI, the geometrical arrangement was determined so as to avoid any trouble to the reactor structure. The conceptual study for major components which compose the beam line system was carried out including the estimation of the neutron streaming. The power supply system was studied also and the work was concentrated on the acceleration power supply which requires the output voltage of 500 kV and fast cut-off time. A basic concept, in which a inverter with a AC switch is used and the frequency of the supplied AC line is increased was proposed. In these works, the configuration of the neutral beam injection system was detailed and it was shown that the beam line seems to be well implemented with the geometrical constraints related to the reactor configuration. (author)

  7. Critical safety issues in the design of fusion machines

    International Nuclear Information System (INIS)

    Kramer, W.

    1991-01-01

    In the course of developing fusion machines both general safety considerations and safety assessments for the various components and systems of actual machines increase in number and become more and more coherent. This is particularly true for the NET/ITER projects where safety analysis plays an increasing role for the design of the machine. Since in a D/T tokamak the radiological hazards will be dominant basic radiological safety objectives are discussed. Critical safety issues as identified in particular by the NET/ITER community are reviewed. Subsequently, issues of major concern are considered both for normal operation and for conceivable accidents. The following accidents are considered to be crucial: Loss of cooling in plasma facing components, loss of vacuum, tritium system failure, and magnet system failure. To mitigate accident consequences a confinement concept based on passive features and multiple barriers including detritiation and filtering has to be applied. The reactor building as final barrier needs special attention to cope with both internal and external hazards. (orig.)

  8. 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

  9. Collimator design for neutron imaging of laser-fusion targets

    International Nuclear Information System (INIS)

    Sommargren, G.E.; Lerche, R.A.

    1981-01-01

    Several pinhole collimator geometries for use in neutron imaging experiments have been modeled and compared. Point spread functions are shown for a cylinder, hyperbola, intersecting cones, and a five-zone approximation to the intersecting cones. Of the geometries studied, the intersecting cones appear the most promising with respect to neutron efficiency, field of view, and isoplanatism

  10. Shock Ignition Target Design for Inertial Fusion Energy

    Science.gov (United States)

    2010-01-01

    on the Nike KrF laser (0.248µm wavelength) are consistent with an intensity limit of about a factor of two higher8,9, and further experiments are...McLean, and C. Manka, “Observations of the Two Plasmon Decay Instability Driven by the Nike KrF Laser” submitted to Physical Review Letters. 9J. L...Lehmberg, J. A. Giuliani, and A. J. Schmitt, Journal Appl . Phys. 106, 023103 (2009). 17In general, we use V 2 ≡ I0.78R/ρ∆R, T ≡ R/V , E ≡ IR2T , and ṁ

  11. 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

  12. Conceptual fusion reactor designs based on the laser heat solenoid

    International Nuclear Information System (INIS)

    Steinhauer, L.C.

    1976-01-01

    The feasibility of the laser heated solenoid (LHS) as an approach to fusion and fusion-fission commercial power generation has been examined. The LHS concept is based on magnetic confinement of a long slender plasma column which is partly heated by the axially directed beam from a powerful long wavelength laser. As a pure fusion concept, the LHS configurations studied so far are characterized by fairly difficult engineering constraints, particularly on the magnet, a large laser, and a marginally acceptable system energy balance. As a fusion-fission system, however, the LHS is capable of a very attractive energy balance, has much more relaxed engineering constraints, requires a relatively modest laser, and as such holds great potential as a power generator and fissile fuel breeding scheme

  13. Trends in radiation protection: possible effects on fusion power plant design

    International Nuclear Information System (INIS)

    Eurajoki, Tapani; Frias, Manuel Pascual; Orlandi, Sergio

    2003-01-01

    Since the design of fusion power plants involves long-term issues, ranging over several decades, it is useful to try to foresee under what kind of regulations the first fusion plants are to be operated. Application of present radiological regulations and practice to a fusion power plant concept is considered. The current design phase of fusion power plants motivates the top-down dose assessment, but it is crucial to aim at bottom-up assessments to ensure radiation doses as low as reasonably achievable. Since several issues, relating both to our knowledge on radiation as well as to the practice of radiation protection, may change in the future, it is necessary to continuously follow the development in the further design of fusion power plants

  14. Spatially and temporally resolved x-ray emission from imploding laser fusion targets

    International Nuclear Information System (INIS)

    Attwood, D.T.; Coleman, L.W.; Boyle, M.J.; Phillion, D.W.; Swain, J.E.; Manes, K.R.; Larsen, J.T.

    1976-09-01

    The Livermore 15 psec x-ray streak camera has been used in conjunction with 6 μm diameter pinholes to record well resolved implosion histories of DT filled laser fusion targets. The space-time compression data provide clearly identified implosion velocities, typically 3 x 10 7 cm/sec for two-sided clamshell irradiation of a 70 μm/sup D/, .5 μm wall DT filled glass microshell. Single-sided irradiation results show hydrodynamic convergence at the target center, followed by an asymmetric but two-sided target disassembly. These experiments were performed at the two arm Janus Laser facility, which typically delivered a total of 0.4 TW in a 70 psec pulse for these experiments

  15. Potential design modifications for the High Yield Lithium Injection Fusion Energy (HYLIFE) reaction chamber

    International Nuclear Information System (INIS)

    Pitts, J.H.; Hovingh, J.; Meier, W.R.; Monsler, M.J.; Powell, E.G.; Walker, P.E.

    1979-01-01

    Generation of electric power from inertial confinement fusion requires a reaction chamber. One promising type, the High Yield Lithium Injection Fusion Energy (HYLIFE) chamber, includes a falling array of liquid lithium jets. These jets act as: (1) a renewable first wall and blanket to shield metal components from x-ray and neutron exposure, (2) a tritium breeder to replace tritium burned during the fusion process, and (3) an absorber and transfer medium for fusion energy. Over 90% of the energy produced in the reaction chamber is absorbed in the lithium jet fall. Design aspects are included

  16. Recent contributions to fusion reactor design and technology development

    International Nuclear Information System (INIS)

    1979-11-01

    The report contains a collection of 16 recent fusion technology papers on the STARFIRE Project, the study of alternate fusion fuel cycles, a maintainability study, magnet safety, neutral beam power supplies and pulsed superconducting magnets and energy transfer. This collection of papers contains contributions for Argonne National Laboratory, McDonnell Douglas Astronautics Company, General Atomic Company, The Ralph M. Parsons Company, the University of Illinois, and the University of Wisconsin. Separate abstracts are presented for each paper

  17. TALE-PvuII Fusion Proteins – Novel Tools for Gene Targeting

    Science.gov (United States)

    Yanik, Mert; Alzubi, Jamal; Lahaye, Thomas; Cathomen, Toni; Pingoud, Alfred; Wende, Wolfgang

    2013-01-01

    Zinc finger nucleases (ZFNs) consist of zinc fingers as DNA-binding module and the non-specific DNA-cleavage domain of the restriction endonuclease FokI as DNA-cleavage module. This architecture is also used by TALE nucleases (TALENs), in which the DNA-binding modules of the ZFNs have been replaced by DNA-binding domains based on transcription activator like effector (TALE) proteins. Both TALENs and ZFNs are programmable nucleases which rely on the dimerization of FokI to induce double-strand DNA cleavage at the target site after recognition of the target DNA by the respective DNA-binding module. TALENs seem to have an advantage over ZFNs, as the assembly of TALE proteins is easier than that of ZFNs. Here, we present evidence that variant TALENs can be produced by replacing the catalytic domain of FokI with the restriction endonuclease PvuII. These fusion proteins recognize only the composite recognition site consisting of the target site of the TALE protein and the PvuII recognition sequence (addressed site), but not isolated TALE or PvuII recognition sites (unaddressed sites), even at high excess of protein over DNA and long incubation times. In vitro, their preference for an addressed over an unaddressed site is > 34,000-fold. Moreover, TALE-PvuII fusion proteins are active in cellula with minimal cytotoxicity. PMID:24349308

  18. TALE-PvuII fusion proteins--novel tools for gene targeting.

    Science.gov (United States)

    Yanik, Mert; Alzubi, Jamal; Lahaye, Thomas; Cathomen, Toni; Pingoud, Alfred; Wende, Wolfgang

    2013-01-01

    Zinc finger nucleases (ZFNs) consist of zinc fingers as DNA-binding module and the non-specific DNA-cleavage domain of the restriction endonuclease FokI as DNA-cleavage module. This architecture is also used by TALE nucleases (TALENs), in which the DNA-binding modules of the ZFNs have been replaced by DNA-binding domains based on transcription activator like effector (TALE) proteins. Both TALENs and ZFNs are programmable nucleases which rely on the dimerization of FokI to induce double-strand DNA cleavage at the target site after recognition of the target DNA by the respective DNA-binding module. TALENs seem to have an advantage over ZFNs, as the assembly of TALE proteins is easier than that of ZFNs. Here, we present evidence that variant TALENs can be produced by replacing the catalytic domain of FokI with the restriction endonuclease PvuII. These fusion proteins recognize only the composite recognition site consisting of the target site of the TALE protein and the PvuII recognition sequence (addressed site), but not isolated TALE or PvuII recognition sites (unaddressed sites), even at high excess of protein over DNA and long incubation times. In vitro, their preference for an addressed over an unaddressed site is > 34,000-fold. Moreover, TALE-PvuII fusion proteins are active in cellula with minimal cytotoxicity.

  19. Multi-UAV Doppler Information Fusion for Target Tracking Based on Distributed High Degrees Information Filters

    Directory of Open Access Journals (Sweden)

    Hamza Benzerrouk

    2018-03-01

    Full Text Available Multi-Unmanned Aerial Vehicle (UAV Doppler-based target tracking has not been widely investigated, specifically when using modern nonlinear information filters. A high-degree Gauss–Hermite information filter, as well as a seventh-degree cubature information filter (CIF, is developed to improve the fifth-degree and third-degree CIFs proposed in the most recent related literature. These algorithms are applied to maneuvering target tracking based on Radar Doppler range/range rate signals. To achieve this purpose, different measurement models such as range-only, range rate, and bearing-only tracking are used in the simulations. In this paper, the mobile sensor target tracking problem is addressed and solved by a higher-degree class of quadrature information filters (HQIFs. A centralized fusion architecture based on distributed information filtering is proposed, and yielded excellent results. Three high dynamic UAVs are simulated with synchronized Doppler measurement broadcasted in parallel channels to the control center for global information fusion. Interesting results are obtained, with the superiority of certain classes of higher-degree quadrature information filters.

  20. Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

    Energy Technology Data Exchange (ETDEWEB)

    Benoît, Fabrice, E-mail: fabrice-2.benoit@cea.fr [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Allegretti, Ludovic [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Aumeunier, Marie-Hélène [OPTIS, ZE de La Farlède, F-83078 Toulon Cedex 9 (France); Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Robert, Julien [SOFYNE, F-69800 Saint Priest (France)

    2014-10-15

    Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule.

  1. Implementation of a design and configuration management platform for fusion components on the Tore Supra WEST Project

    International Nuclear Information System (INIS)

    Benoît, Fabrice; Allegretti, Ludovic; Aumeunier, Marie-Hélène; Bucalossi, Jérôme; Doceul, Louis; Faïsse, Frederic; Firdaouss, Medhi; Geynet, Michel; Houtte, Didier van; Larroque, Sébastien; Magaud, Philippe; Maini, Patrick; Missirlian, Marc; Parrat, Hélène; Robert, Julien

    2014-01-01

    Highlights: •A design and configuration management platform is under development for managing fusion components lifecycle at CEA. •Design platform ensures an efficient sharing of the data and provides connections between the different software and databases involved in fusion components design. •Design platform rollout on WEST project is ongoing as part of change control and configuration management implementation. -- Abstract: This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule

  2. Targeted proteins for diabetes drug design

    Science.gov (United States)

    Doan Trang Nguyen, Ngoc; Thi Le, Ly

    2012-03-01

    Type 2 diabetes mellitus is a common metabolism disorder characterized by high glucose in the bloodstream, especially in the case of insulin resistance and relative insulin deficiency. Nowadays, it is very common in middle-aged people and involves such dangerous symptoms as increasing risk of stroke, obesity and heart failure. In Vietnam, besides the common treatment of insulin injection, some herbal medication is used but no unified optimum remedy for the disease yet exists and there is no production of antidiabetic drugs in the domestic market yet. In the development of nanomedicine at the present time, drug design is considered as an innovative tool for researchers to study the mechanisms of diseases at the molecular level. The aim of this article is to review some common protein targets involved in type 2 diabetes, offering a new idea for designing new drug candidates to produce antidiabetic drugs against type 2 diabetes for Vietnamese people.

  3. Targeted proteins for diabetes drug design

    International Nuclear Information System (INIS)

    Trang Nguyen, Ngoc Doan; Le, Ly Thi

    2012-01-01

    Type 2 diabetes mellitus is a common metabolism disorder characterized by high glucose in the bloodstream, especially in the case of insulin resistance and relative insulin deficiency. Nowadays, it is very common in middle-aged people and involves such dangerous symptoms as increasing risk of stroke, obesity and heart failure. In Vietnam, besides the common treatment of insulin injection, some herbal medication is used but no unified optimum remedy for the disease yet exists and there is no production of antidiabetic drugs in the domestic market yet. In the development of nanomedicine at the present time, drug design is considered as an innovative tool for researchers to study the mechanisms of diseases at the molecular level. The aim of this article is to review some common protein targets involved in type 2 diabetes, offering a new idea for designing new drug candidates to produce antidiabetic drugs against type 2 diabetes for Vietnamese people. (review)

  4. Targeted recombinant fusion proteins of IFNγ and mimetic IFNγ with PDGFβR bicyclic peptide inhibits liver fibrogenesis in vivo.

    Directory of Open Access Journals (Sweden)

    Ruchi Bansal

    Full Text Available Hepatic stellate cells (HSCs, following transdifferentiation to myofibroblasts plays a key role in liver fibrosis. Therefore, attempts to attenuate this myofibroblastic phenotype would be a promising therapeutic approach. Interferon gamma (IFNγ is a potent anti-fibrotic cytokine, but its pleiotropic receptor expression leading to severe adverse effects has limited its clinical application. Since, activated HSC express high-level of platelet derived growth factor beta receptor (PDGFβR, we investigated the potential of PDGFβR-specific targeting of IFNγ and its signaling peptide that lacks IFNγR binding site (mimetic IFNγ or mimIFNγ in liver fibrosis. We prepared DNA constructs expressing IFNγ, mimIFNγ or BiPPB (PDGFβR-specific bicyclic peptide-IFNγ, BiPPB-mimIFNγ fusion proteins. Both chimeric proteins alongwith IFNγ and mimIFNγ were produced in E.coli. The expressed proteins were purified and analyzed for PDGFβR-specific binding and in vitro effects. Subsequently, these recombinant proteins were investigated for the liver uptake (pSTAT1α signaling pathway, for anti-fibrotic effects and adverse effects (platelet counts in CCl4-induced liver fibrogenesis in mice. The purified HSC-targeted IFNγ and mimIFNγ fusion proteins showed PDGFβR-specific binding and significantly reduced TGFβ-induced collagen-I expression in human HSC (LX2 cells, while mouse IFNγ and mimIFNγ did not show any effect. Conversely, mouse IFNγ and BiPPB-IFNγ induced activation and dose-dependent nitric oxide release in mouse macrophages (express IFNγR while lack PDGFβR, which was not observed with mimIFNγ and BiPPB-mimIFNγ, due to the lack of IFNγR binding sites. In vivo, targeted BiPPB-IFNγ and BiPPB-mimIFNγ significantly activated intrahepatic IFNγ-signaling pathway compared to IFNγ and mimIFNγ suggesting increased liver accumulation. Furthermore, the targeted fusion proteins ameliorated liver fibrogenesis in mice by significantly reducing

  5. The design status of the liquid lithium target facility of IFMIF at the end of the engineering design activities

    Energy Technology Data Exchange (ETDEWEB)

    Nitti, F.S., E-mail: francesco.nitti@enea.it [IFMIF/EVEDA Project Team, Rokkasho Japan (Japan); Ibarra, A. [CIEMAT, Madrid (Spain); Ida, M. [IHI Corporation, Tokyo (Japan); Favuzza, P. [ENEA Research Center Firenze (Italy); Furukawa, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Groeschel, F. [KIT Research Center, Karlsruhe (Germany); Heidinger, R. [F4E Research Center, Garching (Germany); Kanemura, T. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Knaster, J. [IFMIF/EVEDA Project Team, Rokkasho Japan (Japan); Kondo, H. [JAEA Research Center, Tokai-mura, Ibaraki (Japan); Micchiche, G. [ENEA Research Center, Brasimone (Italy); Sugimoto, M. [JAEA Research Center, Rokkasho Japan (Japan); Wakai, E. [JAEA Research Center, Tokai-mura, Ibaraki (Japan)

    2015-11-15

    Highlights: • Results of validation and design activity for the Li loop facility of IFMIF. • Demonstration of Li target stability, with surface disturbance <1 mm. • Demonstration of start-up and shut down procedures of Li loop. • Complete design of the heat removal system and C and O purification system. • Conceptual design of N and H isotopes purification systems. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed.

  6. The design status of the liquid lithium target facility of IFMIF at the end of the engineering design activities

    International Nuclear Information System (INIS)

    Nitti, F.S.; Ibarra, A.; Ida, M.; Favuzza, P.; Furukawa, T.; Groeschel, F.; Heidinger, R.; Kanemura, T.; Knaster, J.; Kondo, H.; Micchiche, G.; Sugimoto, M.; Wakai, E.

    2015-01-01

    Highlights: • Results of validation and design activity for the Li loop facility of IFMIF. • Demonstration of Li target stability, with surface disturbance <1 mm. • Demonstration of start-up and shut down procedures of Li loop. • Complete design of the heat removal system and C and O purification system. • Conceptual design of N and H isotopes purification systems. - Abstract: The International Fusion Material Irradiation Facility (IFMIF) is an experimental facility conceived for qualifying and characterizing structural materials for nuclear fusion applications. The Engineering Validation and Engineering Design Activity (EVEDA) is a fundamental step towards the final design. It presented two mandates: the Engineering Validation Activities (EVA), still on-going, and the Engineering Design Activities (EDA) accomplished on schedule in June 2013. Five main facilities are identified in IFMIF, among which the Lithium Target Facility constituted a technological challenge overcome thanks to the success of the main validation challenges impacting the design. The design of the liquid Lithium Target Facility at the end of the EDA phase is here detailed.

  7. Signal processing, sensor fusion, and target recognition; Proceedings of the Meeting, Orlando, FL, Apr. 20-22, 1992

    Science.gov (United States)

    Libby, Vibeke; Kadar, Ivan

    Consideration is given to a multiordered mapping technique for target prioritization, a neural network approach to multiple-target-tracking problems, a multisensor fusion algorithm for multitarget multibackground classification, deconvolutiom of multiple images of the same object, neural networks and genetic algorithms for combinatorial optimization of sensor data fusion, classification of atmospheric acoustic signals from fixed-wing aircraft, and an optics approach to sensor fusion for target recognition. Also treated are a zoom lens for automatic target recognition, a hybrid model for the analysis of radar sensors, an innovative test bed for developing and assessing air-to-air noncooperative target identification algorithms, SAR imagery scene segmentation using fractal processing, sonar feature-based bandwidth compression, laboratory experiments for a new sonar system, computational algorithms for discrete transform using fixed-size filter matrices, and pattern recognition for power systems.

  8. Broad target cell selectivity of Kaposi's sarcoma-associated herpesvirus glycoprotein-mediated cell fusion and virion entry

    International Nuclear Information System (INIS)

    Kaleeba, Johnan A.R.; Berger, Edward A.

    2006-01-01

    The molecular mechanism of Kaposi's sarcoma-associated herpesvirus (KSHV, human herpesvirus 8) entry is poorly understood. We tested a broad variety of cell types of diverse species and tissue origin for their ability to function as targets in a quantitative reporter gene assay for KSHV-glycoprotein-mediated cell fusion. Several human, non-human primate, and rabbit cell lines were efficient targets, whereas rodent and all human lymphoblastoid cell lines were weak targets. Parallel findings were obtained with a virion entry assay using a recombinant KSHV encoding a reporter gene. No correlation was observed between target cell activity and surface expression of α3β1 integrin, a proposed KSHV receptor. We hypothesize that target cell permissiveness in both the cell fusion and virion entry assays reflects the presence of a putative KSHV fusion-entry receptor

  9. A tri-modality image fusion method for target delineation of brain tumors in radiotherapy.

    Directory of Open Access Journals (Sweden)

    Lu Guo

    Full Text Available To develop a tri-modality image fusion method for better target delineation in image-guided radiotherapy for patients with brain tumors.A new method of tri-modality image fusion was developed, which can fuse and display all image sets in one panel and one operation. And a feasibility study in gross tumor volume (GTV delineation using data from three patients with brain tumors was conducted, which included images of simulation CT, MRI, and 18F-fluorodeoxyglucose positron emission tomography (18F-FDG PET examinations before radiotherapy. Tri-modality image fusion was implemented after image registrations of CT+PET and CT+MRI, and the transparency weight of each modality could be adjusted and set by users. Three radiation oncologists delineated GTVs for all patients using dual-modality (MRI/CT and tri-modality (MRI/CT/PET image fusion respectively. Inter-observer variation was assessed by the coefficient of variation (COV, the average distance between surface and centroid (ADSC, and the local standard deviation (SDlocal. Analysis of COV was also performed to evaluate intra-observer volume variation.The inter-observer variation analysis showed that, the mean COV was 0.14(± 0.09 and 0.07(± 0.01 for dual-modality and tri-modality respectively; the standard deviation of ADSC was significantly reduced (p<0.05 with tri-modality; SDlocal averaged over median GTV surface was reduced in patient 2 (from 0.57 cm to 0.39 cm and patient 3 (from 0.42 cm to 0.36 cm with the new method. The intra-observer volume variation was also significantly reduced (p = 0.00 with the tri-modality method as compared with using the dual-modality method.With the new tri-modality image fusion method smaller inter- and intra-observer variation in GTV definition for the brain tumors can be achieved, which improves the consistency and accuracy for target delineation in individualized radiotherapy.

  10. Expression and activity analysis of a new fusion protein targeting ovarian cancer cells.

    Science.gov (United States)

    Su, Manman; Chang, Weiqin; Wang, Dingding; Cui, Manhua; Lin, Yang; Wu, Shuying; Xu, Tianmin

    2015-09-01

    The aim of the present study was to develop a new therapeutic drug to improve the prognosis of ovarian cancer patients. Human urokinase-type plasminogen activator (uPA)17-34-kunitz-type protease inhibitor (KPI) eukaryotic expression vector was constructed and recombinant human uPA17-34-KPI (rhuPA17-34-KPI) in P. pastoris was expressed. In the present study, the DNA sequences that encode uPA 17-34 amino acids were created according to the native amino acids sequence and inserted into the KPI-pPICZαC vector, which was constructed. Then, uPA17‑34-KPI-pPICZαC was transformed into P. pastoris X-33, and rhuPA17-34-KPI was expressed by induction of methanol. The bioactivities of a recombinant fusion protein were detected with trypsin inhibition analysis, and the inhibitory effects on the growth of ovarian cancer cells were identified using the TUNEL assay, in vitro wound‑healing assay and Matrigel model analysis. The results of the DNA sequence analysis of the recombinant vector uPA17-34-KPI‑pPICZα demonstrated that the DNA‑encoding human uPA 17-34 amino acids, 285-288 amino acids of amyloid precursor protein (APP) and 1-57 amino acids of KPI were correctly inserted into the pPICZαC vector. Following induction by methonal, the fusion protein with a molecular weight of 8.8 kDa was observed using SDS-PAGE and western blot analysis. RhuPA17-34-KPI was expressed in P. pastoris with a yield of 50 mg/l in a 50-ml tube. The recombinant fusion protein was able to inhibit the activity of trypsin, inhibit growth and induce apoptosis of SKOV3 cells, and inhibit the invasion and metastasis of ovarian cancer cells. By considering uPA17-34 amino acid specific binding uPAR as the targeted part of fusion protein and utilizing the serine protease inhibitor activity of KPI, it was found that the recombinant fusion protein uPA17-34-KPI inhibited the invasion and metastasis of ovarian tumors, and may therefore be regarded as effective in targeted treatment.

  11. HYLIFE-II inertial fusion energy power plant design

    International Nuclear Information System (INIS)

    Moir, R.W.

    1992-01-01

    The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets, to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li 2 BeF 4 ) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. In addition, although not adequately considered for HYLIFE-I, there is liquid splash that must be forcibly cleared because gravity is too slow, at higher repetition rates than 1 Hz. Splash removal is accomplished in the central region by oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW·h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost, that is, a zero cost driver would give a calculated cost of electricity of 0.045 $/kWh

  12. HYLIFE-II inertial fusion energy power plant design

    International Nuclear Information System (INIS)

    Moir, R.W.

    1992-01-01

    The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets, to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li 2 BeF 4 ) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 Gj from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. In addition, although not adequately considered for HYLIFE-I, there is liquid splash that must be forcibly cleared because gravity is too slow, at higher repetition rates than 1 Hz. Splash removal is accomplished in the central region by oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW·h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost, that is, a zero cost driver would give a calculated cost of electricity of 0.045 $/kWh

  13. HYLIFE-II inertial confinement fusion reactor design

    International Nuclear Information System (INIS)

    Moir, R.W.

    1990-01-01

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li 2 BeF 4 ) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW·h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 15 refs., 9 figs., 3 tabs

  14. HYLIFE-II inertial confinement: Fusion power plant design

    International Nuclear Information System (INIS)

    Moir, R.W.

    1990-01-01

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li 2 BeF 4 ) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW·h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 16 refs., 6 figs., 2 tabs

  15. Target design for materials processing very far from equilibrium

    Science.gov (United States)

    Barnard, John J.; Schenkel, Thomas

    2016-10-01

    Local heating and electronic excitations can trigger phase transitions or novel material states that can be stabilized by rapid quenching. An example on the few nanometer scale are phase transitions induced by the passage of swift heavy ions in solids where nitrogen-vacancy color centers form locally in diamonds when ions heat the diamond matrix to warm dense matter conditions at 0.5 eV. We optimize mask geometries for target materials such as silicon and diamond to induce phase transitions by intense ion pulses (e. g. from NDCX-II or from laser-plasma acceleration). The goal is to rapidly heat a solid target volumetrically and to trigger a phase transition or local lattice reconstruction followed by rapid cooling. The stabilized phase can then be studied ex situ. We performed HYDRA simulations that calculate peak temperatures for a series of excitation conditions and cooling rates of crystal targets with micro-structured masks. A simple analytical model, that includes ion heating and radial, diffusive cooling, was developed that agrees closely with the HYDRA simulations. The model gives scaling laws that can guide the design of targets over a wide range of parameters including those for NDCX-II and the proposed BELLA-i. This work was performed under the auspices of the U.S. DOE under contracts DE-AC52-07NA27344 (LLNL), DE-AC02-05CH11231 (LBNL) and was supported by the US DOE Office of Science, Fusion Energy Sciences. LLNL-ABS-697271.

  16. Ubiquitin fusion expression and tissue-dependent targeting of hG-CSF in transgenic tobacco

    Science.gov (United States)

    2011-01-01

    Background Human granulocyte colony-stimulating factor (hG-CSF) is an important human cytokine which has been widely used in oncology and infection protection. To satisfy clinical needs, expression of recombinant hG-CSF has been studied in several organisms, including rice cell suspension culture and transient expression in tobacco leaves, but there was no published report on its expression in stably transformed plants which can serve as a more economical expression platform with potential industrial application. Results In this study, hG-CSF expression was investigated in transgenic tobacco leaves and seeds in which the accumulation of hG-CSF could be enhanced through fusion with ubiquitin by up to 7 fold in leaves and 2 fold in seeds, leading to an accumulation level of 2.5 mg/g total soluble protein (TSP) in leaves and 1.3 mg/g TSP in seeds, relative to hG-CSF expressed without a fusion partner. Immunoblot analysis showed that ubiquitin was processed from the final protein product, and ubiquitination was up-regulated in all transgenic plants analyzed. Driven by CaMV 35S promoter and phaseolin signal peptide, hG-CSF was observed to be secreted into apoplast in leaves but deposited in protein storage vacuole (PSV) in seeds, indicating that targeting of the hG-CSF was tissue-dependent in transgenic tobacco. Bioactivity assay showed that hG-CSF expressed in both seeds and leaves was bioactive to support the proliferation of NFS-60 cells. Conclusions In this study, the expression of bioactive hG-CSF in transgenic plants was improved through ubiquitin fusion strategy, demonstrating that protein expression can be enhanced in both plant leaves and seeds through fusion with ubiquitin and providing a typical case of tissue-dependent expression of recombinant protein in transgenic plants. PMID:21985646

  17. Technique for thick polymer coating of inertial-confinement-fusion targets

    International Nuclear Information System (INIS)

    Lee, M.C.; Feng, I.; Wang, T.G.; Kim, H.

    1983-01-01

    A novel technique has been developed to coat a thick layer (15--50 μm) of polymer materials on inertial-confinement-fusion (ICF) targets. In this technique, the target and the coating material are independently positioned and manipulated. The coating material is first dissolved in an appropriate solvent to form a polymer solution. The solution is then atomized, transported, and allowed to coalesce into a droplet in a stable acoustic levitating field. The ICF target mounted on a stalk is moved into the acoustic field by manipulating a three-dimensional (3-D) positioner to penetrate the surface membrane of the droplet and thus the target is immersed in the levitated coating solution. The 3-D coordinates of the target inside the droplet are obtained using two orthogonally placed television cameras. The target is positioned at the geometric center of the droplet and maintained at that location by continuously manipulating the 3-D device until the coating layer is dried. Tests of this technique using a polymer solution have been highly successful

  18. On stimulated scattering of laser light in inertial fusion energy targets

    International Nuclear Information System (INIS)

    Nikolic, Lj; Skoric, M.M.; Ishiguro, S.; Sato, T.

    2002-11-01

    Propagation of a laser light through regions of an underdense plasma is an active research topic in laser fusion. In particular, a large effort has been invested in studies of stimulated Raman scattering (SRS) and stimulated Brillouin scattering (SBS) which can reflect laser energy and produce energetic particles to preheat a fusion energy target. Experiments, theory and simulations agree on a complex interplay between various laser-plasma instabilities. By particle-in-cell simulations of an underdense electron-plasma, we have found, apart from the standard SRS, a strong backscattering near the electron plasma frequency at densities beyond the quarter critical. This novel instability, recognized in recent experiments as stimulated laser scattering on a trapped electron-acoustic mode (SEAS), is absent from a classical theory of laser-parametric instabilities. A parametric excitation of SEAS instability, is explained by a three-wave resonant decay of the incident laser light into a standing backscattered wave and a slow trapped electron acoustic wave (ω p ). Large SEAS pulsations, eventually suppressed by relativistic heating of electrons, are observed in our simulations. This phenomenon seems relevant to future hohlraum target and fast ignition experiments. (author)

  19. Enhancing Ignition Probability and Fusion Yield in NIF Indirect Drive Targets with Applied Magnetic Fields

    Science.gov (United States)

    Perkins, L. John; Logan, B. Grant; Ho, Darwin; Zimmerman, George; Rhodes, Mark; Blackfield, Donald; Hawkins, Steven

    2017-10-01

    Imposed magnetic fields of tens of Tesla that increase to greater than 10 kT (100 MGauss) under capsule compression may relax conditions for ignition and propagating burn in indirect-drive ICF targets. This may allow attainment of ignition, or at least significant fusion energy yields, in presently-performing ICF targets on the National Ignition Facility that today are sub-marginal for thermonuclear burn through adverse hydrodynamic conditions at stagnation. Results of detailed 2D radiation-hydrodynamic-burn simulations applied to NIF capsule implosions with low-mode shape perturbations and residual kinetic energy loss indicate that such compressed fields may increase the probability for ignition through range reduction of fusion alpha particles, suppression of electron heat conduction and stabilization of higher-mode RT instabilities. Optimum initial applied fields are around 50 T. Off-line testing has been performed of a hohlraum coil and pulsed power supply that could be integrated on NIF; axial fields of 58T were obtained. Given the full plasma structure at capsule stagnation may be governed by 3-D resistive MHD, the formation of closed magnetic field lines might further augment ignition prospects. Experiments are now required to assess the potential of applied magnetic fields to NIF ICF ignition and burn. Work performed under auspices of U.S. DOE by LLNL under Contract DE-AC52-07NA27344.

  20. Modelling of surface evolution of rough surface on divertor target in fusion devices

    International Nuclear Information System (INIS)

    Dai, Shuyu; Liu, Shengguang; Sun, Jizhong; Kirschner, A.; Kawamura, G.; Tskhakaya, D.; Ding, Rui; Luo, Guangnan; Wang, Dezhen

    2015-01-01

    Highlights: • We study the surface evolution of rough surface on divertor target in fusion devices. • The effects of gyration motion and E × B drift affect 3D angular distribution. • A larger magnetic field angle leads to a reduced net eroded areal density. • The rough surface evolution affects the physical sputtering yield. - Abstract: The 3D Monte-Carlo code SURO has been used to study the surface evolution of rough surface on the divertor target in fusion devices. The edge plasma at divertor region is modelled by the SDPIC code and used as input data for SURO. Coupled with SDPIC, SURO can perform more sophisticated simulations to calculate the local angle and surface evolution of rough surface. The simulation results show that the incident direction of magnetic field, gyration and E × B force has a significant impact on 3D angular distribution of background plasma and accordingly on the erosion of rough surface. The net eroded areal density of rough surface is studied by varying the magnetic field angle with surface normal. The evolution of the microscopic morphology of rough surface can lead to a significant change in the physical sputtering yield

  1. Conceptual design of the Purdue compact torus/passive liner fusion reactor

    International Nuclear Information System (INIS)

    Terry, W.K.

    1981-01-01

    This proposal describes a program for the conceptual development of a novel fusion reactor design, the Purdue Compact Torus/Passive Liner Reactor. The key features of the concept are described and a comparison is made with a conventional tokamak

  2. Engineering design of a fusion test reactor (FTR) and fusion engineering research facility (FERF) based on a toroidal theta pinch

    International Nuclear Information System (INIS)

    Abdou, M.; Burke, R.J.; Dauzvardis, P.V.; Foss, M.; Gerstl, S.A.W.; Maroni, V.A.; Pierce, A.W.; Turner, A.F.; Krakowski, R.A.; Linford, R.K.; Oliphant, T.A.; Ribe, F.L.; Thomassen, K.I.

    1975-01-01

    This paper describes two advanced toroidal theta-pinch devices which are being proposed for future construction. The Fusion Test Reactor (FTR) is being designed to produce thermonuclear energy (at 20 MeV/neutron) equal to the maximum plasma energy (Q=1) and to demonstrate α-particle heating. The Fusion Engineering and Research Facility (FERF) is being designed to test materials in a fusion environment where the average 14-MeV neutron flux from the plasma is greater than or of the order of 5.10 13 n/cm 2 .s over large surface areas. These devices employ the staged theta-pinch principle where the heating is accomplished by rapid (about 0.1 μs) implosion and expansion followed by a slow compression of the plasma. The rapid implosion injects as much heat as possible at as large a plasma radious as possible so that the plasma remains stable even after further compression. The final compression to ignition requires the transfer of a large amount of magnetic energy which implies a long transfer time (about 1 ms) for realistic voltages in the driving circuit. Throughout the heating and burn cycle the plasma must remain in equilibrium and stable to the dominant MHD-modes. A sufficiently large plasma radius guarantees stability against the m = 1 modes. These equilibrium and stability conditions and the requirements on thermonuclear burn determine the design parameters for either machine. The design parameters must also be consistent with economic limitations and technological feasibility of components. In addition to these requirements, the FERF must provide a steady and reliable source of fusion neutrons. (author)

  3. Safety and environmental aspects of the HYLIFE-II and ARIES fusion reactor designs

    International Nuclear Information System (INIS)

    Dolan, T.J.; Longhurst, G.R.; Herring, J.S.

    1993-01-01

    The HYLIFE-II inertial confinement fusion reactor design uses jets of Flibe molten salt to protect the blast chamber walls and to breed tritium. It has a low tritium inventory and effective tritium removal. The issue with this design is not one of safety but of economics. The ARIES reactor designs have safety concerns associated with fires. These reactors designs are described

  4. Accuracy of position measurement method using Arago spot for inertial fusion energy target tracking system

    International Nuclear Information System (INIS)

    Saruta, Koichi; Tsuji, Ryusuke

    2007-01-01

    The accuracy of a position measurement method using the Arago spot is reported for an inertial fusion energy (IFE) target tracking system, where the position of the target is determined by the position of the Arago spot, which is a bright spot appearing in the central portion of the diffraction pattern of a spherical obstacle. We use a He-Ne laser as the light source and a charge-coupled device (CCD) camera with a microscope objective lens to magnify and record the diffraction pattern of a spherical target. We examine two different algorithms to determine the center of the Arago spot in order to compare the measurement performances. The experimental results show that the position of a 5-mm-diameter target can be obtained with a measurement resolution of 1 μm and an rms measurement error of less than 0.2μm for both algorithms when the distance between the target and the microscope objective lens is 5 cm. (author)

  5. 3D Surface Mapping of Capsule Fill-Tube Assemblies used in Laser-Driven Fusion Targets

    Energy Technology Data Exchange (ETDEWEB)

    Buice, E S; Alger, E T; Antipa, N A; Bhandarkar, S D; Biesiada, T A; Conder, A D; Dzenitis, E G; Flegel, M S; Hamza, A V; Heinbockel, C L; Horner, J; Johnson, M A; Kegelmeyer, L M; Meyer, J S; Montesanti, R C; Reynolds, J L; Taylor, J S; Wegner, P J

    2011-02-18

    This paper presents the development of a 3D surface mapping system used to measure the surface of a fusion target Capsule Fill-Tube Assembly (CFTA). The CFTA consists of a hollow Ge-doped plastic sphere, called a capsule, ranging in outer diameter between 2.2 mm and 2.6 mm and an attached 150 {micro}m diameter glass-core fill-tube that tapers down to a 10{micro} diameter at the capsule. The mapping system is an enabling technology to facilitate a quality assurance program and to archive 3D surface information of each capsule used in fusion ignition experiments that are currently being performed at the National Ignition Facility (NIF). The 3D Surface Mapping System is designed to locate and quantify surface features with a height of 50 nm and 300 nm in width or larger. Additionally, the system will be calibrated such that the 3D measured surface can be related to the capsule surface angular coordinate system to within 0.25 degree (1{sigma}), which corresponds to approximately 5 {micro}m linear error on the capsule surface.

  6. 3D Surface Mapping of Capsule Fill-Tube Assemblies used in Laser-Driven Fusion Targets

    International Nuclear Information System (INIS)

    Buice, E.S.; Alger, E.T.; Antipa, N.A.; Bhandarkar, S.D.; Biesiada, T.A.; Conder, A.D.; Dzenitis, E.G.; Flegel, M.S.; Hamza, A.V.; Heinbockel, C.L.; Horner, J.; Johnson, M.A.; Kegelmeyer, L.M.; Meyer, J.S.; Montesanti, R.C.; Reynolds, J.L.; Taylor, J.S.; Wegner, P.J.

    2011-01-01

    This paper presents the development of a 3D surface mapping system used to measure the surface of a fusion target Capsule Fill-Tube Assembly (CFTA). The CFTA consists of a hollow Ge-doped plastic sphere, called a capsule, ranging in outer diameter between 2.2 mm and 2.6 mm and an attached 150 (micro)m diameter glass-core fill-tube that tapers down to a 10(micro) diameter at the capsule. The mapping system is an enabling technology to facilitate a quality assurance program and to archive 3D surface information of each capsule used in fusion ignition experiments that are currently being performed at the National Ignition Facility (NIF). The 3D Surface Mapping System is designed to locate and quantify surface features with a height of 50 nm and 300 nm in width or larger. Additionally, the system will be calibrated such that the 3D measured surface can be related to the capsule surface angular coordinate system to within 0.25 degree (1σ), which corresponds to approximately 5 (micro)m linear error on the capsule surface.

  7. Investigation of different cage designs and mechano-regulation algorithms in the lumbar interbody fusion process - a finite element analysis.

    Science.gov (United States)

    Postigo, Sergio; Schmidt, Hendrik; Rohlmann, Antonius; Putzier, Michael; Simón, Antonio; Duda, Georg; Checa, Sara

    2014-04-11

    Lumbar interbody fusion cages are commonly used to treat painful spinal degeneration and instability by achieving bony fusion. Many different cage designs exist, however the effect of cage morphology and material properties on the fusion process remains largely unknown. This finite element model study aims to investigate the influence of different cage designs on bone fusion using two mechano-regulation algorithms of tissue formation. It could be observed that different cages play a distinct key role in the mechanical conditions within the fusion region and therefore regulate the time course of the fusion process. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. Thermal-hydraulic analysis of low activity fusion blanket designs

    International Nuclear Information System (INIS)

    Fillo, J.A.; Powell, J.; Yu, W.S.

    1977-01-01

    The heat transfer aspects of fusion blankets are considered where: (a) conduction and (b) boiling and condensation are the dominant heat transfer mechanisms. In some cases, unique heat transfer problems arise and additional heat transfer data and analyses may be required

  9. 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

  10. Neutron pre-emission at the fusion of 11 Li halo nuclei with Si targets

    International Nuclear Information System (INIS)

    Petrascu, M.; Isbasescu, A.; Petrascu, H.; Bordeanu, C.; David, I.; Lazar, I.; Mihai, I.; Vaman, G.; Tanihata, I.; Kobayashi, T.; Korsheninnikov, A.; Fukuda, S.; Kumagai, H.; Momota, S.; Ozawa, A.; Yoshida, K.; Nikolski, E.; Giurgiu, M.

    1997-01-01

    In this contribution, the first experiment on fusion of 11 Li halo nuclei with Si targets is reported. A novel effect consisting of a large neutron pre-emission probability in the fusion process was observed. The neutron halo nuclei are characterized by very large matter radii, small separation energy and small internal momentum of the valence neutrons. Until now, the halo nuclei were investigated mostly by elastic, inelastic scattering and breakup processes. It was recently predicted that due to the very large dimension of 11 Li, one may expect, that in a fusion experiment on a light target, the valence neutrons will not be absorbed together with the 9 Li core, but will be emitted in the early stage of the reaction process. The experiment aiming to check this expectation, performed at RIKEN-RIPS facility, is described. In the experimental arrangement, three main parts are present: the first part contains the detectors used for the control, identification and determination of the beam characteristics; the second part consists of a Multiple Sampling Ionisation Chamber (MUSIC), used for identification of the inclusive evaporation residue spectra produced in the detector-target; the third part consists of two wall neutron detectors, each made up of 15 plastic scintillators. This detector was used for the energy and position determination of the neutrons originating from the target. The projectile energy range was 11.2 - 15.2 AMeV, centered at 13 AMeV. The neutrons resulting from the reaction were measured by time-of-light technique. The position on the 'wall' of the detected neutrons could be also determined. The measured neutron spectra from 11 Li and 9 Li are shown. A marked different between the two spectra was found and it is explained by the contribution of a large amount of pre-emission (breakup) processes, in case of 11 Li projectiles. The position spectra point out the evaporation origin of the neutrons in case of 9 Li projectiles while for 11 Li only the

  11. Identification of target genes of synovial sarcoma-associated fusion oncoprotein using human pluripotent stem cells

    Energy Technology Data Exchange (ETDEWEB)

    Hayakawa, Kazuo [Department of Tissue Regeneration, Institute for Frontier Medical Sciences, Kyoto University, Kyoto (Japan); Department of Cell Growth and Differentiation, Center for iPS Cell Research and Application, Kyoto University, Kyoto (Japan); Department of Orthopaedic Surgery, Graduate School of Medical Sciences, Nagoya City University, Nagoya (Japan); Ikeya, Makoto [Department of Cell Growth and Differentiation, Center for iPS Cell Research and Application, Kyoto University, Kyoto (Japan); Fukuta, Makoto [Department of Tissue Regeneration, Institute for Frontier Medical Sciences, Kyoto University, Kyoto (Japan); Department of Cell Growth and Differentiation, Center for iPS Cell Research and Application, Kyoto University, Kyoto (Japan); Department of Orthopaedic Surgery, Graduate School of Medical Sciences, Nagoya City University, Nagoya (Japan); Woltjen, Knut [Department of Reprogramming Sciences, Center for iPS Cell Research and Application, Kyoto University, Kyoto (Japan); Tamaki, Sakura; Takahara, Naoko; Kato, Tomohisa; Sato, Shingo [Department of Tissue Regeneration, Institute for Frontier Medical Sciences, Kyoto University, Kyoto (Japan); Otsuka, Takanobu [Department of Orthopaedic Surgery, Graduate School of Medical Sciences, Nagoya City University, Nagoya (Japan); Toguchida, Junya, E-mail: togjun@frontier.kyoto-u.ac.jp [Department of Tissue Regeneration, Institute for Frontier Medical Sciences, Kyoto University, Kyoto (Japan); Department of Cell Growth and Differentiation, Center for iPS Cell Research and Application, Kyoto University, Kyoto (Japan); Department of Orthopaedic Surgery, Graduate School of Medicine, Kyoto University, Kyoto (Japan)

    2013-03-22

    Highlights: ► We tried to identify targets of synovial sarcoma (SS)-associated SYT–SSX fusion gene. ► We established pluripotent stem cell (PSC) lines with inducible SYT–SSX gene. ► SYT–SSX responsive genes were identified by the induction of SYT–SSX in PSC. ► SS-related genes were selected from database by in silico analyses. ► 51 genes were finally identified among SS-related genes as targets of SYT–SSX in PSC. -- Abstract: Synovial sarcoma (SS) is a malignant soft tissue tumor harboring chromosomal translocation t(X; 18)(p11.2; q11.2), which produces SS-specific fusion gene, SYT–SSX. Although precise function of SYT–SSX remains to be investigated, accumulating evidences suggest its role in gene regulation via epigenetic mechanisms, and the product of SYT–SSX target genes may serve as biomarkers of SS. Lack of knowledge about the cell-of-origin of SS, however, has placed obstacle in the way of target identification. Here we report a novel approach to identify SYT–SSX2 target genes using human pluripotent stem cells (hPSCs) containing a doxycycline-inducible SYT–SSX2 gene. SYT–SSX2 was efficiently induced both at mRNA and protein levels within three hours after doxycycline administration, while no morphological change of hPSCs was observed until 24 h. Serial microarray analyses identified genes of which the expression level changed more than twofold within 24 h. Surprisingly, the majority (297/312, 95.2%) were up-regulated genes and a result inconsistent with the current concept of SYT–SSX as a transcriptional repressor. Comparing these genes with SS-related genes which were selected by a series of in silico analyses, 49 and 2 genes were finally identified as candidates of up- and down-regulated target of SYT–SSX, respectively. Association of these genes with SYT–SSX in SS cells was confirmed by knockdown experiments. Expression profiles of SS-related genes in hPSCs and human mesenchymal stem cells (hMSCs) were strikingly

  12. Conceptual design of a moving-ring fusion reactor. AP-3229 Research Project 922, annual report, October 1983

    International Nuclear Information System (INIS)

    1983-10-01

    This report concludes a two-year project to design a prototype (roughly 100 MW(e) net) fusion power plant which places emphasis on commercial needs and to outline a development plan for it. This project is a follow-on to an earlier design of a small (11 MW(e) net) pilot power plant (EPRI Report AP-1544), which is included in the development plan as a step to be built before this prototype. The prototype design put major attention on minimizing nuclear issues. Other commercial matters, including suitability for scaling to a wide range of sizes, were also very important. The design, the earlier pilot design, and the extrapolation of these to commercial size designs identify gaps between required technical features and current technical knowledge. These gaps set targets for experiments and technology programs. These programs and designs define the development plan

  13. Design of a fusion reaction-history measurement system with high temporal resolution

    International Nuclear Information System (INIS)

    Peng Xiaoshi; Wang Feng; Liu Shenye; Jiang Xiaohua; Tang Qi

    2010-01-01

    In order to accurately measure the history of fusion reaction for experimental study of inertial confinement fusion, we advance the design of a fusion reaction-history measurement system with high temporal resolution. The diagnostic system is composed of plastic scintillator and nose cone, an optical imaging system and the system of optic streak camera. Analyzing the capability of the system indicated that the instrument measured fusion reaction history at temporal resolution as low as 55ps and 40ps correspond to 2.45MeV DD neutrons and 14.03MeV DT neutrons. The instrument is able to measure the fusion reaction history at yields 1.5 x 10 9 DD neutrons, about 4 x 10 8 DT neutrons are required for a similar quality signal. (authors)

  14. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.; Ott, K.O.; Terry, W.K.

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

  15. Range shortening, radiation transport, and Rayleigh-Taylor instability phenomena in ion-beam-driven inertial-fusion-reactor-size targets: Implosion, ignition, and burn phases

    International Nuclear Information System (INIS)

    Long, K.A.; Tahir, N.A.

    1987-01-01

    In this paper we present an analysis of the theory of the energy deposition of ions in cold materials and hot dense plasmas together with numerical calculations for heavy and light ions of interest to ion-beam fusion. We have used the gorgon computer code of Long, Moritz, and Tahir (which is an extension of the code originally written for protons by Nardi, Peleg, and Zinamon) to carry out these calculations. The energy-deposition data calculated in this manner has been used in the design of heavy-ion-beam-driven fusion targets suitable for a reactor, by its inclusion in the medusa code of Christiansen, Ashby, and Roberts as extended by Tahir and Long. A number of other improvements have been made in this code and these are also discussed. Various aspects of the theoretical analysis of such targets are discussed including the calculation of the hydrodynamic stability, the hydrodynamic efficiency, and the gain. Various different target designs have been used, some of them new. In general these targets are driven by Bi + ions of energy 8--12 GeV, with an input energy of 4--6.5 MJ, with output energies in the range 600--900 MJ, and with gains in the range 120--180. The peak powers are in the range of 500--750 TW. We present detailed calculations of the ablation, compression, ignition, and burn phases. By the application of a new stability analysis which includes ablation and density-gradient effects we show that these targets appear to implode in a stable manner. Thus the targets designed offer working examples suited for use in a future inertial-confinement fusion reactor

  16. Single-particle fusion of influenza viruses reveals complex interactions with target membranes

    Science.gov (United States)

    van der Borg, Guus; Braddock, Scarlett; Blijleven, Jelle S.; van Oijen, Antoine M.; Roos, Wouter H.

    2018-05-01

    The first step in infection of influenza A virus is contact with the host cell membrane, with which it later fuses. The composition of the target bilayer exerts a complex influence on both fusion efficiency and time. Here, an in vitro, single-particle approach is used to study this effect. Using total internal reflection fluorescence (TIRF) microscopy and a microfluidic flow cell, the hemifusion of single virions is visualized. Hemifusion efficiency and kinetics are studied while altering target bilayer cholesterol content and sialic-acid donor. Cholesterol ratios tested were 0%, 10%, 20%, and 40%. Sialic-acid donors GD1a and GYPA were used. Both cholesterol ratio and sialic-acid donors proved to have a significant effect on hemifusion efficiency. Furthermore, comparison between GD1a and GYPA conditions shows that the cholesterol dependence of the hemifusion time is severely affected by the sialic-acid donor. Only GD1a shows a clear increasing trend in hemifusion efficiency and time with increasing cholesterol concentration of the target bilayer with maximum rates for GD1A and 40% cholesterol. Overall our results show that sialic acid donor and target bilayer composition should be carefully chosen, depending on the desired hemifusion time and efficiency in the experiment.

  17. First generation of fusion power plants: Design and technology. Proceedings of the 2. IAEA technical meeting

    International Nuclear Information System (INIS)

    2008-01-01

    This series of meetings has been initiated under recommendation of the International Fusion Research Council for the IAEA and is expected to initiate, develop and mature ideas on fusion strategy that would be of benefit for all players. The present objectives of this meeting are to provide a forum to discuss concepts, technology and environmental aspects of future fusion power plants, the next step following ITER, their role in future energy mix and to assess a selection of urgent topics aiming at identifying the physics and the technological requirements that ITER and a fusion grade materials developing programme will have to address to support the construction of a DEMO(s) fusion power plant(s) prototype demonstrating viable economics. The meeting was organized in five sessions addressing five topics: - (PPCA) Power Plant Concepts and systems Analysis. - (MCP) Materials analysis/Components design/Plasma requirements - (NE) Non-Electric applications of fusion - (SESE) Social, Economic, Safety and Environmental aspects of fusion - (EP) Energy Policy, strategy and scenario for fusion development. A summary session took place at the end of the meeting. Thirty-three participants representing 12 Countries and 3 International Organizations were present at the meeting

  18. Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

    Science.gov (United States)

    Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

    2018-06-01

    The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit

  19. Novel treatment option for MUC16-positive malignancies with the targeted TRAIL-based fusion protein Meso-TR3

    International Nuclear Information System (INIS)

    Garg, Gunjal; Spitzer, Dirk; Gibbs, Jesse; Belt, Brian; Powell, Matthew A; Mutch, David G; Goedegebuure, Peter; Collins, Lynne; Piwnica-Worms, David; Hawkins, William G

    2014-01-01

    The targeted delivery of cancer therapeutics represents an ongoing challenge in the field of drug development. TRAIL is a promising cancer drug but its activity profile could benefit from a cancer-selective delivery mechanism, which would reduce potential side effects and increase treatment efficiencies. We recently developed the novel TRAIL-based drug platform TR3, a genetically fused trimer with the capacity for further molecular modifications such as the addition of tumor-directed targeting moieties. MUC16 (CA125) is a well characterized biomarker in several human malignancies including ovarian, pancreatic and breast cancer. Mesothelin is known to interact with MUC16 with high affinity. In order to deliver TR3 selectively to MUC16-expressing cancers, we investigated the possibility of targeted TR3 delivery employing the high affinity mesothelin/MUC16 ligand/receptor interaction. Using genetic engineering, we designed the novel cancer drug Meso-TR3, a fusion protein between native mesothelin and TR3. The recombinant proteins were produced with mammalian HEK293T cells. Meso-TR3 was characterized for binding selectivity and killing efficacy against MUC16-positive cancer cells and controls that lack MUC16 expression. Drug efficacy experiments were performed in vitro and in vivo employing an intraperitoneal xenograft mouse model of ovarian cancer. Similar to soluble mesothelin itself, the strong MUC16 binding property was retained in the Meso-TR3 fusion protein. The high affinity ligand/receptor interaction was associated with a selective accumulation of the cancer drug on MUC16-expressing cancer targets and directly correlated with increased killing activity in vitro and in a xenograft mouse model of ovarian cancer. The relevance of the mesothelin/MUC16 interaction for attaching Meso-TR3 to the cancer cells was verified by competitive blocking experiments using soluble mesothelin. Mechanistic studies using soluble DR5-Fc and caspase blocking assays confirmed

  20. Interfacial Stability of Spherically Converging Plasma Jets for Magnetized Target Fusion

    Science.gov (United States)

    Thio, Y. C. Francis; Cassibry, Jason; Wu, S. T.; Eskridge, Richard; Smith, James; Lee, Michael; Rodgers, Stephen L. (Technical Monitor)

    2000-01-01

    A fusion propulsion scheme has been proposed that makes use of the merging of a spherical distribution of plasma jets to dynamically form a gaseous liner to implode a magnetized target to produce the fusion reaction. In this paper, a study is made of the interfacial stability of the interaction of these jets. Specifically, the Orr-Sommerfeld equation is integrated to obtain the growth rate of a perturbation to the primary flow at the interface between the colliding jets. The results lead to an estimate on the tolerances on the relative flow velocities of the merging plasma jets to form a stable, imploding liner. The results show that the maximum temporal growth rate of the perturbed flow at the jet interface is very small in comparison with the time to full compression of the liner. These data suggest that, as far as the stability of the interface between the merging jets is concerned, the formation of the gaseous liner can withstand velocity variation of the order of 10% between the neighboring jets over the density and temperature ranges investigated.

  1. US-DOE Fusion-Breeder Program: blanket design and system performance

    International Nuclear Information System (INIS)

    Lee, J.D.

    1983-01-01

    Conceptual design studies are being used to assess the technical and economic feasibility of fusion's potential to produce fissile fuel. A reference design of a fission-suppressed blanket using conventional materials is under development. Theoretically, a fusion breeder that incorporates this fusion-suppressed blanket surrounding a 3000-MW tandem mirror fusion core produces its own tritium plus 5600 kg of 233 U per year. The 233 U could then provide fissile makeup for 21 GWe of light-water reactor (LWR) power using a denatured thorium fuel cycle with full recycle. This is 16 times the net electric power produced by the fusion breeder (1.3 GWe). The cost of electricity from this fusion-fission system is estimated to be only 23% higher than the cost from LWRs that have makeup from U 3 O 8 at present costs (55 $/kg). Nuclear performance, magnetohydrodynamics (MHD), radiation effects, and other issues concerning the fission-suppressed blanket are summarized, as are some of the present and future objectives of the fusion breeder program

  2. HIGH SPATIAL RESOLUTION IMAGING OF INERTIAL FUSION TARGET PLASMAS USING BUBBLE NEUTRON DETECTORS

    International Nuclear Information System (INIS)

    FISHER, R.K.

    2003-01-01

    OAK B202 HIGH SPATIAL RESOLUTION IMAGING OF INERTIAL FUSION TARGET PLASMAS USING BUBBLE NEUTRON DETECTORS. Bubble detectors, which can detect neutrons with a spatial 5 to 30 (micro), are the most promising approach to imaging NIF target plasmas with the desired 5 (micro) spatial resolution in the target plane. Gel bubble detectors are being tested to record neutron images of ICF implosions in OMEGA experiments. By improving the noise reduction techniques used in analyzing the data taken in June 2000, we have been able to image the neutron emission from 6 · 10 13 yield DT target plasmas with a target plane spatial resolution of ∼ 140 (micro). As expected, the spatial resolution was limited by counting statistics as a result of the low neutron detection efficiency of the easy-to-use gel bubble detectors. The results have been submitted for publication and will be the subject of an invited talk at the October 2001 Meeting of the Division of Plasma Physics of the American Physical Society. To improve the counting statistics, data was taken in May 2001 using a stack of four gel detectors and integrated over a series of up to seven high-yield DT shots. Analysis of the 2001 data is still in its early stages. Gel detectors were chosen for these initial tests since the bubbles can be photographed several hours after the neutron exposure. They consist of ∼ 5000 drops (∼ 100 (micro) in diameter) of bubble detector liquid/cm 3 suspended in an inactive support gel that occupies ∼ 99% of the detector volume. Using a liquid bubble chamber detector and a light scattering system to record the bubble locations a few microseconds after the neutron exposure when the bubbles are ∼ 10 (micro) in diameter, should result in ∼ 1000 times higher neutron detection efficiency and a target plane resolution on OMEGA of ∼ 10 to 50 (micro)

  3. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    International Nuclear Information System (INIS)

    1997-01-01

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  4. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  5. Design aspects of a multipurpose fusion power plant for desalination and agrochemical processes

    International Nuclear Information System (INIS)

    Sabri, Z.A.

    1975-02-01

    A description is given of the skeletal structure of a multipurpose fusion power plant, designed for desalination and agrochemicals production. The plant is a complex that comprises dual purpose power and desalination units, separation and processing units for recovery of soluble salts in the effluent of the desalination unit, mariculture units for production of algae for food and as food for shrimp and other fish species. The electrical power unit is a two-component fusion device that burns deuterium and helium-3 utilizing a fast fusion cycle

  6. The reversed-field-pinch (RFP) fusion neutron source: A conceptual design

    International Nuclear Information System (INIS)

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

    1989-01-01

    The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at ∼5-MW/m 2 steady-state DT fusion neutron wall loading and ∼124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low input power (∼206 MW) source of DT neutrons for large-volume (∼10 m 3 ), high-fluence (3.4 MW yr/m 2 ) fusion nuclear materials and technology testing. 19 refs., 15 figs., 9 tabs

  7. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    International Nuclear Information System (INIS)

    Martone, M.

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  8. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Martone, M [ENEA, Centro Ricerche Frascati, Rome (Italy)

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  9. Progress on the conceptual design of a mirror hybrid fusion--fission reactor

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Burleigh, R.J.

    1975-01-01

    A conceptual design study was made of a fusion-fission reactor for the purpose of producing fissile material and electricity. The fusion component is a D-T plasma confined by a pair of magnetic mirror coils in a Yin-Yang configuration and is sustained by neutral beam injection. The neutrons from the fusion plasma drive the fission assembly which is composed of natural uranium carbide fuel rods clad with stainless steel and helium cooled. It was shown conceptually how the reactor might be built using essentially present-day technology and how the uranium-bearing blanket modules can be routinely changed to allow separation of the bred fissile fuel

  10. Design system for in-vessel mainipulator of fusion reactor 'DESIM'

    International Nuclear Information System (INIS)

    Adachi, Junihci; Kobayashi, Takeshi; Ise, Hideo; Sato, Keisuke; Matsuda, Hirotsugu

    1989-01-01

    A computer aided design system 'DESIM' for the in-vessel manipulators of nuclear fusion reactors has been developed to design the manipulators efficiently. The DESIM consists of the following subsystems: (1) the design system for arm mechanisms to realize optimum manipulation performance in the specified workspace; (2) the robot simulator to study manipulator movement, postures and interference problems; (3) the CAD system which is used to define the structure object data for robots, and the interface system for the data conversion from the CAD system to the robot simulator. The DESIM has been used to design the in-vessel manipulator for the Fusion Experimental Reactor (FER) to confirm the effectiveness. (author)

  11. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  12. Research on Key Technologies of Network Centric System Distributed Target Track Fusion

    Directory of Open Access Journals (Sweden)

    Yi Mao

    2017-01-01

    Full Text Available To realize common tactical picture in network-centered system, this paper proposes a layered architecture for distributed information processing and a method for distributed track fusion on the basis of analyzing the characteristics of network-centered systems. Basing on the noncorrelation of three-dimensional measurement of surveillance and reconnaissance sensors under polar coordinates, it also puts forward an algorithm for evaluating track quality (TQ using statistical decision theory. According to simulation results, the TQ value is associated with the measurement accuracy of sensors and the motion state of targets, which is well matched with the convergence process of tracking filters. Besides, the proposed algorithm has good reliability and timeliness in track quality evaluation.

  13. Manufacturing W fibre-reinforced Cu composite pipes for application as heat sink in divertor targets of future nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Alexander v.; You, Jeong-Ha [Max-Planck-Institut fuer Plasmaphysik, 85748 Garching (Germany); Ewert, Dagmar [Institut fuer Textil- und Verfahrenstechnik Denkendorf, 73770 Denkendorf (Germany); Siefken, Udo [Louis Renner GmbH, 85221 Dachau (Germany)

    2016-07-01

    An important plasma-facing component (PFC) in future nuclear fusion reactors is the so-called divertor which allows power exhaust and removal of impurities from the main plasma. The most highly loaded parts of a divertor are the target plates which have to withstand intense particle bombardment. This intense particle bombardment leads to high heat fluxes onto the target plates which in turn lead to severe thermomechanical loads. With regard to future nuclear fusion reactors, an improvement of the performance of divertor targets is desirable in order to ensure reliable long term operation of such PFCs. The performance of a divertor target is most closely linked to the properties of the materials that are used for its design. W fibre-reinforced Cu (Wf/Cu) composites are regarded as promising heat sink materials in this respect. These materials do not only feature adequate thermophysical and mechanical properties, they do also offer metallurgical flexibility as their microstructure and hence their macroscopic properties can be tailored. The contribution will point out how Wf/Cu composites can be used to realise an advanced design of a divertor target and how these materials can be fabricated by means of liquid Cu infiltration.

  14. Primary heat transfer loop design for the Cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Murray, K.A.; McDowell, M.W.

    1984-05-01

    This study investigates a heat exchanger and balance of plant design to accompany the Cascade inertial confinement fusion reaction chamber concept. The concept uses solid Li 2 O or other lithium-ceramic granules, held to the wall of a rotating reaction chamber by centrifugal action, as a tritium breeding blanket and first wall protection. The Li 2 O granules enter the chamber at 800 K and exit at 1200 K after absorbing the thermal energy produced by the fusion process

  15. Double emulsion generation in the mass production of inertial confinement fusion targets using T-junctions

    Science.gov (United States)

    Moynihan, Matthew J.

    This work demonstrates a new device for the continuous controlled production of double emulsions for the manufacturing of inertial confinement fusion targets. This device can be integrated into a microfluidic approach to produce targets which should increase the yield and quality of the targets and at a lower cost. The device is a double T-Junction, which has been scaled, optimized and built to produce oil-in-water-in-oil double emulsions from diameters of roughly 4 mm or less. A T-Junction is an intersection of two channels at a right angle where fluid emerges and is broken off to form droplets. A systematic study presented here has shown that a single T-Junction has four modes of operation: squeezing, dripping, transition and streaming. The droplet size may be controlled by controlling the fluid flow rate through the channels; the droplet increases with increasing dispersed flow and decreasing continuous flow. The device was utilized to produce hundreds of ˜ 2.5 mm diameter resorcinol formaldehyde double emulsions with better than 2 percent reproducibility in diameter. The device was used to produce 2.0 mm shells with an average wall thickness of 510 microns.

  16. Chamber wall response to target implosion in inertial fusion reactors: new and critical assessments

    International Nuclear Information System (INIS)

    Hassanein, A.; Morozov, V.

    2002-01-01

    The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to next target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The focus of this study is to critically assess the reliability and the dynamic response of chamber walls in IFE systems. Of particular concern is the effect on wall erosion lifetime due to various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls

  17. Chamber wall response to target implosion in inertial fusion reactors : new and critical assessments

    International Nuclear Information System (INIS)

    Hassanein, A.; Morozov, V.

    2002-01-01

    The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The hydrodynamic response of gas filled cavities and photon radiation transport of the deposited energy has been calculated by means of new and advanced numerical techniques. Fragmentation models of liquid jets as a result of the deposited energy have also been developed, and the impact on chamber clearing dynamics has been evaluated. Th focus of this study is to critically assess the reliability and the dynamic response of chamber walls in various proposed protection methods for IFE systems. Of particular concern is the effect on wall erosion lifetime of various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls

  18. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  19. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  20. Combinatorial synthesis and screening of cancer cell-specific nanomedicines targeted via phage fusion proteins

    Directory of Open Access Journals (Sweden)

    James W. Gillespie

    2015-06-01

    Full Text Available Active tumor targeting of nanomedicines has recently shown significant improvements in the therapeutic activity of currently existing drug delivery systems, such as liposomal doxorubicin (Doxil/Caelyx/Lipodox. Previously, we have shown that isolated pVIII major coat proteins of the fd tet filamentous phage vector, containing cancer cell-specific peptide fusions at their N terminus, can be used as active targeting ligands in a liposomal doxorubicin delivery system in vitro and in vivo. Here, we show a novel major coat protein isolation procedure in 2-propanol that allows spontaneous incorporation of the hydrophobic protein core into preformed liposomal doxorubicin with minimal damage or drug loss while still retaining the targeting ligand exposed for cell-specific targeting. Using a panel of 12 structurally unique ligands with specificity towards breast, lung, and/or pancreatic cancer, we showed the feasibility of pVIII major coat proteins to significantly increase the throughput of targeting ligand screening in a common nanomedicine core. Phage protein-modified Lipodox samples showed an average doxorubicin recovery of 82.8% across all samples with 100% of protein incorporation in the correct orientation (N-terminus exposed. Following cytotoxicity screening in a doxorubicin-sensitive breast cancer line (MCF-7, three major groups of ligands were identified. Ligands showing the most improved cytotoxicity included: DMPGTVLP, ANGRPSMT, VNGRAEAP, and ANDVYLD showing a 25-fold improvement (p < 0.05 in toxicity. Similarly DGQYLGSQ, ETYNQPYL, and GSSEQLYL ligands with specificity towards a doxorubicin-insensitive pancreatic cancer line (PANC-1 showed significant increases in toxicity (2-fold; p < 0.05. Thus, we demonstrated proof-of-concept that pVIII major coat proteins can be screened in significantly higher throughput to identify novel ligands displaying improved therapeutic activity in a desired cancer phenotype.

  1. Breakup and fusion cross sections of the 6Li nucleus with targets of mass A = 58, 144 and 208

    Science.gov (United States)

    Mukeru, B.; Rampho, G. J.; Lekala, M. L.

    2018-04-01

    We use the continuum discretized coupled channels method to investigate the effects of continuum-continuum coupling on the breakup and fusion cross sections of the weakly bound 6Li nucleus with the 58Ni, 144Sm and 208Pb nuclear targets. The cross sections were analyzed at incident energies E cm below, close to and above the Coulomb barrier V B. We found that for the medium and heavy targets, the breakup cross sections are enhanced at energies below the Coulomb barrier (E cm/V B ≤ 0.8) owing to these couplings. For the lighter target, relatively small enhancement of the breakup cross sections appear at energies well below the barrier (E cm/V B ≤ 0.6). At energies E cm/V B > 0.8 for medium and heavy targets, and E cm/V B > 0.6 for the light target, the continuum-continuum couplings substantially suppress the breakup cross sections. On the other hand, the fusion cross sections are enhanced at energies E cm/V B fusion cross sections. We also compared the breakup and fusion cross sections, and found that below the barrier, the breakup cross sections are more dominant regardless of whether continuum-continuum couplings are included.

  2. The RCC-MR design code for LMFBR components. A useful basic for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1985-11-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials (Stainless steels), temperature service level (550-600 0 C), loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain

  3. The RCC-MR design code for LMFBR components. A useful basis for fusion reactor design tools development

    International Nuclear Information System (INIS)

    Acker, D.; Chevereau, G.

    1986-01-01

    LMFBR and fusion reactors exhibit common features with regard to structural materials, temperature service level, loading types. So, design and construction rules used in France for LMFBR, that is to say RCC-MR Code, can constitute a good basis for fusion reactors design. Some original aspects of RCC-MR design rules are described, relating to unsignificant creep, ratchetting effect, fatigue and creep damage limits, creep damage evaluation, fatigue damage evaluation, buckling. The main originality of RCC-MR consists to propose comprehensive simplified rules based on elastic calculations and extended from classical cold temperatures to the elevated temperature domain. (author)

  4. Flunarizine Prevents Hepatitis C Virus Membrane Fusion in a Genotype-dependent Manner by Targeting the Potential Fusion Peptide within E1

    Science.gov (United States)

    Perin, Paula M.; Haid, Sibylle; Brown, Richard J. P.; Doerrbecker, Juliane; Schulze, Kai; Zeilinger, Carsten; von Schaewen, Markus; Heller, Brigitte; Vercauteren, Koen; Luxenburger, Eva; Baktash, Yasmine M.; Vondran, Florian W. R.; Speerstra, Sietkse; Awadh, Abdullah; Mukhtarov, Furkat; Schang, Luis M; Kirschning, Andreas; Müller, Rolf; Guzman, Carlos A.; Kaderali, Lars; Randall, Glenn; Meuleman, Philip; Ploss, Alexander; Pietschmann, Thomas

    2015-01-01

    To explore mechanisms of hepatitis C virus (HCV) replication we screened a compound library including licensed drugs. Flunarizine, a diphenylmethylpiperazine used to treat migraine, inhibited HCV cell entry in vitro and in vivo in a genotype-dependent fashion. Analysis of mosaic viruses between susceptible and resistant strains revealed that E1 and E2 glycoproteins confer susceptibility to flunarizine. Time of addition experiments and single particle tracking of HCV demonstrated that flunarizine specifically prevents membrane fusion. Related phenothiazines and pimozide also inhibited HCV infection and preferentially targeted HCV genotype 2 viruses. However, phenothiazines and pimozide exhibited improved genotype coverage including the difficult to treat genotype 3. Flunarizine-resistant HCV carried mutations within the alleged fusion peptide and displayed cross-resistance to these compounds, indicating that these drugs have a common mode of action. Conclusion: These observations reveal novel details about HCV membrane fusion. Moreover, flunarizine and related compounds represent first-in-class HCV fusion inhibitors that merit consideration for repurposing as cost-effective component of HCV combination therapies. PMID:26248546

  5. Conceptual design study of quasi-steady state fusion experimental reactor (FEQ-Q), part 1

    International Nuclear Information System (INIS)

    1985-12-01

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 JER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included; core plasma, reactor structure, reactor core components, magnets. (author)

  6. Conceptual design study of quasi-steady state fusion experimental reactor (FER-Q), part 2

    International Nuclear Information System (INIS)

    1985-12-01

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 FER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included: heating/current drive system, plasma position control, power supply, diagnostics, neutronics, blanket test module, repair and maintenance and safety. (author)

  7. KrF amplifier design issues and application to inertial confinement fusion system design

    International Nuclear Information System (INIS)

    Sullivan, J.A.; Allen, G.R.; Berggren, R.R.

    1993-01-01

    Los Alamos National Laboratory has assembled an array of experimental and theoretical tools to optimize amplifier design for future single-pulse KrF lasers. The next opportunity to exercise these tools is with the design of the second-generation NIKE system under construction at the Naval Research Laboratory with the collaboration of Los Alamos National Laboratory. Los Alamos has applied these amplifier design tools to the conceptual design of a 100-kJ Laser Target Test Facility and a 3-MJ Laboratory Microfusion Facility. (author)

  8. Developing a commercial production process for 500,000 targets per day: A key challenge for inertial fusion energy

    International Nuclear Information System (INIS)

    Goodin, D.T.; Alexander, N.B.; Besenbruch, G.E.; Bozek, A.S.; Brown, L.C.; Flint, G.W.; Kilkenny, J.D.; McQuillan, B.W.; Nikroo, A.; Paguio, R.R.; Petzoldt, R.W.; Schroen, D.G.; Sheliak, J.D.; Vermillion, B.A.; Carlson, L.C.; Goodman, P.; Maksaereekul, W.; Raffray, R.; Spalding, J.; Tillack, M.S.

    2006-01-01

    As is true for current-day commercial power plants, a reliable and economic fuel supply is essential for the viability of future Inertial Fusion Energy (IFE) [Energy From Inertial Fusion, edited by W. J. Hogan (International Atomic Energy Agency, Vienna, 1995)] power plants. While IFE power plants will utilize deuterium-tritium (DT) bred in-house as the fusion fuel, the 'target' is the vehicle by which the fuel is delivered to the reaction chamber. Thus the cost of the target becomes a critical issue in regard to fuel cost. Typically six targets per second, or about 500 000/day are required for a nominal 1000 MW(e) power plant. The electricity value within a typical target is about $3, allocating 10% for fuel cost gives only 30 cents per target as-delivered to the chamber center. Complicating this economic goal, the target supply has many significant technical challenge - fabricating the precision fuel-containing capsule, filling it with DT, cooling it to cryogenic temperatures, layering the DT into a uniform layer, characterizing the finished product, accelerating it to high velocity for injection into the chamber, and tracking the target to steer the driver beams to meet it with micron-precision at the chamber center

  9. S-layer fusion protein as a tool functionalizing emulsomes and CurcuEmulsomes for antibody binding and targeting.

    Science.gov (United States)

    Ucisik, Mehmet H; Küpcü, Seta; Breitwieser, Andreas; Gelbmann, Nicola; Schuster, Bernhard; Sleytr, Uwe B

    2015-04-01

    Selective targeting of tumor cells by nanoparticle-based drug delivery systems is highly desirable because it maximizes the drug concentration at the desired target while simultaneously protecting the surrounding healthy tissues. Here, we show a design for smart nanocarriers based on a biomimetic approach that utilizes the building principle of virus envelope structures. Emulsomes and CurcuEmulsomes comprising a tripalmitin solid core surrounded by phospholipid layers are modified by S-layer proteins that self-assemble into a two-dimensional array to form a surface layer. One significant advantage of this nanoformulation is that it increases the solubility of the lipophilic anti-cancer agent curcumin in the CurcuEmulsomes by a factor of 2700. In order to make the emulsomes specific for IgG, the S-layer protein is fused with two protein G domains. This S-layer fusion protein preserves its recrystallization characteristics, forming an ordered surface layer (square lattice with 13 nm unit-by-unit distance). The GG domains are presented in a predicted orientation and exhibit a selective binding affinity for IgG. Copyright © 2015 The Authors. Published by Elsevier B.V. All rights reserved.

  10. Employment of colorimetric enzyme assay for monitoring expression and solubility of GST fusion proteins targeted to inclusion bodies.

    Science.gov (United States)

    Mačinković, Igor S; Abughren, Mohamed; Mrkic, Ivan; Grozdanović, Milica M; Prodanović, Radivoje; Gavrović-Jankulović, Marija

    2013-12-01

    High levels of recombinant protein expression can lead to the formation of insoluble inclusion bodies. These complex aggregates are commonly solubilized in strong denaturants, such as 6-8M urea, although, if possible, solubilization under milder conditions could facilitate subsequent refolding and purification of bioactive proteins. Commercially available GST-tag assays are designed for quantitative measurement of GST activity under native conditions. GST fusion proteins accumulated in inclusion bodies are considered to be undetectable by such assays. In this work, solubilization of recombinantly produced proteins was performed in 4M urea. The activity of rGST was assayed in 2M urea and it was shown that rGST preserves 85% of its activity under such denaturing conditions. A colorimetric GST activity assay with 1-chloro-2, 4-dinitrobenzene (CDNB) was examined for use in rapid detection of expression targeted to inclusion bodies and for the identification of inclusion body proteins which can be solubilized in low concentrations of chaotropic agents. Applicability of the assay was evaluated by tracking protein expression of two GST-fused allergens of biopharmaceutical value in E. coli, GST-Der p 2 and GST-Mus a 5, both targeted to inclusion bodies. Copyright © 2013 Elsevier B.V. All rights reserved.

  11. Engineering design of the Nova Laser Facility for inertial-confinement fusion

    International Nuclear Information System (INIS)

    Simmons, W.W.; Godwin, R.O.; Hurley, C.A.

    1982-01-01

    The design of the Nova Laser Facility for inertial confinement fusion experiments at Lawrence Livermore National Laboratory is presented from an engineering perspective. Emphasis is placed upon design-to-performance requirements as they impact the various subsystems that comprise this complex experimental facility

  12. Accuracy verification of PET-CT image fusion and its utilization in target delineation of radiotherapy

    International Nuclear Information System (INIS)

    Wang Xuetao; Yu Jinming; Yang Guoren; Gong Heyi

    2005-01-01

    Objective: Evaluate the accuracy of co-registration of PET and CT (PET-CT) images on line with phantom, and utilize it on patients to provide clinical evidence for target delineation in radiotherapy. Methods: A phantom with markers and different volume cylinders was infused with various concentrations of 18 FDG, and scanned at 4 mm by PET and CT respectively. After having been transmitted into GE eNTEGRA and treatment planning system (TPS) workstations, the images were fused and reconstructed. The distance between the markers and the errors were monitored in PET and CT images respectively. The volume of cylinder in PET and CT images were measured and compared by certain pixel value proportion deduction method. The same procedure was performed on the pulmonary tumor image in ten patients. Results: eNTEGRA and TPS workstations had a good length linearity, but the fusion error of the latter was markedly greater than the former. Tumors in different volume filled by varying concentrations of 18 FDG required different pixel deduction proportion. The cylinder volume of PET and CT images were almost the same, so were the images of pulmonary tumor of ten patients. Conclusions: The accuracy of image co-registration of PET-CT on line may fulfill the clinical demand. Pixel value proportion deduction method can be used for target delineation on PET image. (authors)

  13. Miniature proportional counter for compression measurements of laser-fusion targets

    International Nuclear Information System (INIS)

    Lane, S.M.; Dellis, J.H.; Bennett, C.K.; Campbell, E.M.

    1981-10-01

    Direct drive laser fusion targets consisting of DT gas encapsulated in glass microshells produce 14.1 MeV neutrons that can interact with silicon-28 nuclei in the glass to produce a 2.2 minute aluminum-28 activity. From the number of 28 Al nuclei created and the neutron yield, the compressed glass areal density can be found. To determine the number of activated atoms created, we collect approximately one-half of the target debris on a thin metal foil which is transferred to our beta-gamma coincidence detector. This detector consists of a 25 cm x 25 cm NaI(Tl) crystal having a 5 cm x 15 cm well. We have recently built a miniature proportional counter that fits into this well and is used to detect beta particles. It is constructed of .025 cm thick copper and has nine separate chambers through which methane flows. The coincidence background is 0.14 cpm and the measured beta efficiency is 45%. We are now building a .0125 cm thick counter made of aluminum having a predicted efficiency of > 90%

  14. Design of intense neutron source for fusion material study and the role of universities

    International Nuclear Information System (INIS)

    Ishino, Shiori

    1993-01-01

    Need and requirement for the intense neutron source for fusion materials study have been discussed for many years. Recently, international climate has been becoming gradually maturing to consider this problem more seriously because of the recognition of crucial importance of solving materials problems for fusion energy development. The present symposium was designed to discuss the problems associated with the intense neutron source for material irradiation studies which will have a potential for the National Institute for Fusion Science to become one of the important future research areas. The symposium comprises five sessions; first, the role of materials research in fusion development strategies was discussed followed by a brief summary of current IFMIF (International Fusion Materials Irradiation Facility) activity. Despite the pressing need for intense fusion neutron source, currently available neutron sources are reactor or accelerator based sources of which FFTF and LASREF were discussed. Then, various concepts of intense neutron source candidates were presented including ESNIT, which are currently under design by JAERI. In the fourth session, discussions were made on the study of materials with the intense neutron source from the viewpoint of materials scientists and engineers as the user of the facility. This is followed by discussions on the role of universities from the two stand points, namely, fusion irradiation studies and fusion materials development. Finally summary discussions were made by the participants, indicating important role fundamental studies in universities for the full utilization of irradiation data and the need of pure 14 MeV neutron source for fundamental studies together with the intense surrogate neutron sources. (author)

  15. 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.

  16. 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.).

  17. Inhibition of Nipah virus infection in vivo: targeting an early stage of paramyxovirus fusion activation during viral entry.

    Directory of Open Access Journals (Sweden)

    Matteo Porotto

    2010-10-01

    Full Text Available In the paramyxovirus cell entry process, receptor binding triggers conformational changes in the fusion protein (F leading to viral and cellular membrane fusion. Peptides derived from C-terminal heptad repeat (HRC regions in F have been shown to inhibit fusion by preventing formation of the fusogenic six-helix bundle. We recently showed that the addition of a cholesterol group to HRC peptides active against Nipah virus targets these peptides to the membrane where fusion occurs, dramatically increasing their antiviral effect. In this work, we report that unlike the untagged HRC peptides, which bind to the postulated extended intermediate state bridging the viral and cell membranes, the cholesterol tagged HRC-derived peptides interact with F before the fusion peptide inserts into the target cell membrane, thus capturing an earlier stage in the F-activation process. Furthermore, we show that cholesterol tagging renders these peptides active in vivo: the cholesterol-tagged peptides cross the blood brain barrier, and effectively prevent and treat in an established animal model what would otherwise be fatal Nipah virus encephalitis. The in vivo efficacy of cholesterol-tagged peptides, and in particular their ability to penetrate the CNS, suggests that they are promising candidates for the prevention or therapy of infection by Nipah and other lethal paramyxoviruses.

  18. A study on the fusion reactor - A study on the design feature of fusion reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Jin [Chosun University, Kwangju (Korea, Republic of); Paek, Won Pil; Jang, Soon Hong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Sim, Young Jae [Kyungsang University, Jinju (Korea, Republic of)

    1996-09-01

    The contents and scope of the project can be summarized as, - study on the trend of divertor design - study on characteristics of coolant materials - study on characteristics of divertor materials - study on the thermal analysis method of divertor design. 36 refs., 12 tabs., 16 figs. (author)

  19. Design of the target area for the National Ignition Facility

    International Nuclear Information System (INIS)

    Foley, R.J.; Karpenko, V.P.; Adams, C.H.

    1997-01-01

    The preliminary design of the target area for the National Ignition Facility has been completed. The target area is required to meet a challenging set of engineering system design requirements and user needs. The target area must provide the appropriate conditions before, during, and after each shot. The repeated introduction of large amounts of laser energy into the chamber and subsequent target emissions represent new design challenges for ICF facility design. Prior to each shot, the target area must provide the required target illumination, target chamber vacuum, diagnostics, and optically stable structures. During the shot, the impact of the target emissions on the target chamber, diagnostics, and optical elements is minimized and the workers and public are protected from excessive prompt radiation doses. After the shot, residual radioactivation is managed to allow the required accessibility. Diagnostic data is retrieved, operations and maintenance activities are conducted, and the facility is ready for the next shot. The target area subsystems include the target chamber, target positioner, structural systems, target diagnostics, environmental systems, and the final optics assembly. The engineering design of the major elements of the target area requires a unique combination of precision engineering, structural analysis, opto-mechanical design, random vibration suppression, thermal stability, materials engineering, robotics, and optical cleanliness. The facility has been designed to conduct both x- ray driven targets and to be converted at a later date for direct drive experiments. The NIF has been configured to provide a wide range of experimental environments for the anticipated user groups of the facility. The design status of the major elements of the target area is described

  20. Present status of design, research and development of nuclear fusion reactors and problems

    International Nuclear Information System (INIS)

    1983-04-01

    Seven years have elapsed since the publication of ''Progress of nuclear fusion research and perspective toward the development of power reactors'' by the Atomic Energy Society of Japan in August, 1976. During this period, the research and development of nuclear fusion have changed from plasma physics to reactor technology, being conscious of the realization of fusion reactors. There are the R project in the Institute of Plasma Physics, Nagoya University, and the design and construction of JT-60 in Japan Atomic Energy Research Institute, to put it concretely. Now the research and development taking the economical efficiency into account are adopted. However, the type of fusion reactors is not reduced to tokamak type, accordingly the research and development to meet the diverse possibilities are forwarded. The progress of tokamak reactor research, core plasma design, nuclear design and shielding design, thermal structure design, the design of superconducting magnets, disassembling and repair, safety, economical efficiency, the conceptual design of other types than tokamak and others are reported. (Kako, I.)