WorldWideScience

Sample records for confinement fusion devices

  1. Health physics appraisal guidelines for fusion/confinement devices

    International Nuclear Information System (INIS)

    Neeson, P.M.

    1987-01-01

    Several types of fusion/confinement devices have been developed for a variety of research applications. The health physics considerations for these devices can vary, depending on a number of parameters. This paper presents guidelines for health physics appraisal of such devices, which can be tailored to apply to specific systems. The guidelines can also be useful for establishing ongoing health physics programs for safe operation of the devices

  2. Fusion energy in an inertial electrostatic confinement device using a magnetically shielded grid

    Energy Technology Data Exchange (ETDEWEB)

    Hedditch, John, E-mail: john.hedditch@sydney.edu.au; Bowden-Reid, Richard, E-mail: rbow3948@physics.usyd.edu.au; Khachan, Joe, E-mail: joe.khachan@sydney.edu.au [School of Physics, The University of Sydney, Sydney, New South Whales 2006 (Australia)

    2015-10-15

    Theory for a gridded inertial electrostatic confinement (IEC) fusion system is presented, which shows a net energy gain is possible if the grid is magnetically shielded from ion impact. A simplified grid geometry is studied, consisting of two negatively biased coaxial current-carrying rings, oriented such that their opposing magnetic fields produce a spindle cusp. Our analysis indicates that better than break-even performance is possible even in a deuterium-deuterium system at bench-top scales. The proposed device has the unusual property that it can avoid both the cusp losses of traditional magnetic fusion systems and the grid losses of traditional IEC configurations.

  3. Benefits and drawbacks of low magnetic shears on the confinement in magnetic fusion toroidal devices

    Science.gov (United States)

    Firpo, Marie-Christine; Constantinescu, Dana

    2012-10-01

    The issue of confinement in magnetic fusion devices is addressed within a purely magnetic approach. As it is well known, the magnetic field being divergence-free, the equations of its field lines can be cast in Hamiltonian form. Using then some Hamiltonian models for the magnetic field lines, the dual impact of low magnetic shear is demonstrated. Away from resonances, it induces a drastic enhancement of magnetic confinement that favors robust internal transport barriers (ITBs) and turbulence reduction. However, when low-shear occurs for values of the winding of the magnetic field lines close to low-order rationals, the amplitude thresholds of the resonant modes that break internal transport barriers by allowing a radial stochastic transport of the magnetic field lines may be much lower than the ones obtained for strong shear profiles. The approach can be applied to assess the robustness versus magnetic perturbations of general almost-integrable magnetic steady states, including non-axisymmetric ones such as the important single helicity steady states. This analysis puts a constraint on the tolerable mode amplitudes compatible with ITBs and may be proposed as a possible explanation of diverse experimental and numerical signatures of their collapses.

  4. Fusion, magnetic confinement

    International Nuclear Information System (INIS)

    Berk, H.L.

    1992-01-01

    An overview is presented of the principles of magnetic confinement of plasmas for the purpose of achieving controlled fusion conditions. Sec. 1 discusses the different nuclear fusion reactions which can be exploited in prospective fusion reactors and explains why special technologies need to be developed for the supply of tritium or 3 He, the probable fuels. In Sec. 2 the Lawson condition, a criterion that is a measure of the quality of confinement relative to achieving fusion conditions, is explained. In Sec. 3 fluid equations are used to describe plasma confinement. Specific confinement configurations are considered. In Sec. 4 the orbits of particle sin magneti and electric fields are discussed. In Sec. 5 stability considerations are discussed. It is noted that confinement systems usually need to satisfy stability constraints imposed by ideal magnetohydrodynamic (MHD) theory. The paper culminates with a summary of experimental progress in magnetic confinement. Present experiments in tokamaks have reached the point that the conditions necessary to achieve fusion are being satisfied

  5. 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

  6. Utilization of a Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research. Report of a Coordinated Research Project 2011–2016

    International Nuclear Information System (INIS)

    2016-12-01

    The IAEA actively promotes the development of controlled fusion as a source of energy. Through its coordinated research activities, the IAEA helps Member States to exchange and establish scientific and technical knowledge required for the design, construction and operation of a fusion reactor. Due to their compactness, flexibility and low operation costs, small fusion devices are a great resource for supporting and accelerating the development of mainstream fusion research on large fusion devices such as the International Thermonuclear Experimental Reactor. They play an important role in investigating the physics of controlled fusion, developing innovative technologies and diagnostics, testing new materials, training highly qualified personnel for larger fusion facilities, and supporting educational programmes for young scientists. This publication reports on the research work accomplished within the framework of the Coordinated Research Project (CRP) on Utilization of the Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research, organized and conducted by the IAEA in 2011–2016. The CRP has contributed to the coordination of a network of research institutions, thereby enhancing international collaboration through scientific visits, joint experiments and the exchange of information and equipment. A total of 16 institutions and 14 devices from 13 Member States participated in this CRP (Belgium, Bulgaria, Canada, China, Costa Rica, the Czech Republic, the Islamic Republic of Iran, Kazakhstan, Pakistan, Portugal, the Russian Federation, Ukraine and the United Kingdom).

  7. Fusion devices

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1977-01-01

    Three types of thermonuclear fusion devices currently under development are reviewed for an electric utilities management audience. Overall design features of laser fusion, tokamak, and magnetic mirror type reactors are described and illustrated. Thrusts and trends in current research on these devices that promise to improve performance are briefly reviewed. Twenty photographs and drawings are included

  8. Apparatus and method for removing particle species from fusion-plasma-confinement devices

    Science.gov (United States)

    Hamilton, G.W.

    1981-10-26

    In a mirror fusion plasma confinement apparatus, method and apparatus are provided for selectively removing (pumping) trapped low energy (thermal) particle species from the end cell region, without removing the still useful high energy particle species, and without requiring large power input to accomplish the pumping. Perturbation magnets are placed in the thermal barrier region of the end cell region at the turning point characteristic of trapped thermal particles, thus deflecting the thermal particles from their closed trajectory, causing them to drift sufficiently to exit the thermal barrier.

  9. Inertial confinement fusion

    International Nuclear Information System (INIS)

    Nuckolls, J.H.; Wood, L.L.

    1988-01-01

    Edward Teller has been a strong proponent of harnessing nuclear explosions for peaceful purposes. There are two approaches: Plowshare, which utilizes macro- explosions, and inertial confinement fusion, which utilizes microexplosions. The development of practical fusion power plants is a principal goal of the inertial program. It is remarkable that Teller's original thermonuclear problem, how to make super high yield nuclear explosions, and the opposite problem, how to make ultra low yield nuclear explosions, may both be solved by Teller's radiation implosion scheme. This paper reports on the essential physics of these two thermonuclear domains, which are separated by nine orders of magnitude in yield, provided by Teller's similarity theorem and its exceptions. Higher density makes possible thermonuclear burn of smaller masses of fuel. The leverage is high: the scale of the explosion diminishes with the square of the increase in density. The extraordinary compressibility of matter, first noticed by Teller during the Los Alamos atomic bomb program, provides an almost incredible opportunity to harness fusion. The energy density of thermonuclear fuels isentropically compressed to super high-- -densities---even to ten thousand times solid density---is small compared to the energy density at thermonuclear ignition temperatures. In small masses of fuel imploded to these super high matter densities, the energy required to achieve ignition may be greatly reduced by exploiting thermonuclear propagation from a relatively small hot spot

  10. 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)

  11. High-speed repetitive pellet injector for plasma fueling of magnetic confinement fusion devices

    International Nuclear Information System (INIS)

    Combs, S.K.; Baylor, L.R.; Foust, C.R.

    1993-01-01

    The projected fueling requirements of future magnetic confinement devices for controlled thermonuclear research [e.g., the International Thermonuclear Experimental Reactor (ITER)] indicate that a flexible plasma fueling capability is required. This includes a mix of traditional gas puffing and low- and high-velocity deuterium-tritium pellets. Conventional pellet injectors (based on light gas guns or centrifugal accelerators) can reliably provide frozen hydrogen pellets (1- to 6-mm-diam sizes tested) up to ∼1.3-km/s velocity at the appropriate pellet fueling rates (1 to 10 Hz or greater). For long-pulse operation in a higher velocity regime (>2 km/s), an experiment in collaboration between Oak Ridge National Laboratory (ORNL) and ENEA Frascati is under way. This activity will be carried out in the framework of a collaborative agreement between the US Department of Energy and European Atomic Energy Community -- ENEA Association. In this experiment, an existing ORNL hydrogen extruder (equipped with a pellet chambering mechanism/gun barrel assembly) and a Frascati two-stage light gas gun driver have been combined on a test facility at ORNL. Initial testing has been carried out with single deuterium pellets accelerated up to 2.05 km/s with the two-stage driver; in addition, some preliminary repetitive testing (to commission the diagnostics) was performed at reduced speeds, including sequences at 0.5 to 1 Hz and 10 to 30 pellets. The primary objective of this study is to demonstrate repetitive operation (up to ∼1 Hz) with speeds in the 2- to 3-km/s range. In addition, the strength of extruded hydrogen ice as opposed to that produced in situ by direct condensation in pipe guns can be investigated. The equipment and initial experimental results are described

  12. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems.

  13. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    International Nuclear Information System (INIS)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems

  14. High-speed repetitive pellet injector prototype for magnetic confinement fusion devices

    International Nuclear Information System (INIS)

    Frattolillo, A.; Gasparotto, M.; Migliori, S.; Angelone, G.; Baldarelli, M.; Scaramuzzi, F.; Ronci, G.; Reggiori, A.; Riva, G.; Carlevaro, R.; Daminelli, G.B.

    1992-01-01

    The design of a test facility aimed at demonstrating the feasibility of high-speed repetitive acceleration of solid D 2 pellets for fusion applications, developed in a collaboration between Oak Ridge National Laboratory and ENEA Frascati, is presented. The results of tests performed at the CNPM/CNR on the piston wear in a repetitively operating two-stage gun are also reported

  15. Alternate fusion -- continuous inertial confinement

    International Nuclear Information System (INIS)

    Barnes, D.C.; Turner, L.; Nebel, R.A.

    1993-01-01

    The authors argue that alternate approaches to large tokamak confinement are appropriate for fusion applications if: (1) They do not require magnetic confinement of a much higher quality than demonstrated in tokamaks; (2) Their physics basis may be succinctly stated and experimentally tested; (3) They offer near-term applications to important technical problems; and (4) Their cost to proof-of-principle is low enough to be consistent with current budget realities. An approach satisfying all of these criteria is presented. Fusion systems based on continuous inertial confinement are described. In these approaches, the inertia of a nonequilibrium plasma is used to produce local concentrations of plasma density in space and/or time. One implementation (inertial electrostatic confinement) which has been investigated both experimentally and theoretically uses a system of electrostatic grids to accelerate plasma ions toward a spherical focus. This system produced a steady 2 x 10 10 D-T neutrons/second with an overall fusion gain of 10 -5 in a sphere of about 9 cm radius. Recent theoretical developments show how to raise the fusion gain to order unity or greater by replacing the internal grids by a combination of applied magnetic and electrostatic fields. In these approaches, useful thermonuclear conditions may be produced in a system as small as a few mm radius. Confinement is that of a nonneutralized plasma. A pure electron plasma with a radial beam velocity distribution is absolutely confined by an applied Penning trap field. Spherical convergence of the confined electrons forms a deep virtual cathode near r = 0, in which thermonuclear ions are absolutely confined at useful densities. The authors have examined the equilibrium, stability, and classical relaxation of such systems, and obtained many positive physics results. Equilibria exist for both pure electron and partially charge-neutralized systems with arbitrarily high core-plasma densities

  16. 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

  17. Inertial confinement fusion at NRL

    International Nuclear Information System (INIS)

    Bodner, S.E.; Boris, J.P.; Cooperstein, G.

    1979-01-01

    The NRL Inertial Confinement Fusion Program's emphasis has moved toward pellet concepts which use longer (approximately 10ns) lower intensity driver pulses than previously assumed. For laser drivers, this change was motivated by recent experiments at NRL with enhanced stimulated Brillouin backscatter. For ion drivers, the motivation is the possibility that substantial energy at 10-ns pulse lengths may soon be available. To accept these 10-ns pulses, it may be necessary to consider pellets of larger radius and thinner shell. The computational studies of Rayleigh-Taylor instability at NRL indicate the possibility of a dynamic stabilization of these thinner shells. (author)

  18. Summary on inertial confinement fusion

    International Nuclear Information System (INIS)

    Meyer-Ter-Vehn, J.

    1995-01-01

    Highlights on inertial confinement during the fifteenth international conference on plasma physics and controlled nuclear fusion are briefly summarized. Specifically the following topics are discussed: the US National Ignition Facility presently planned by the US Department of Energy; demonstration of diagnostics for hot spot formation; declassification of Hohlraum target design; fusion targets, in particular, the Hohlraum target design for the National Ignition Facility (NIF), Hohlraum experiments, direct drive implosions, ablative Rayleigh-Taylor instabilities, laser imprinting (of perturbations by the laser on the laser target surface), hot spot formation and mixing, hot spot implosion experiments at Lawrence Livermore National Laboratory, Livermore, USA, time resolving hot spot dynamics at the Institute of Laser Engineering (ILE), Osaka, Japan, laser-plasma interaction

  19. Magnetic confinement fusion energy research

    International Nuclear Information System (INIS)

    Grad, H.

    1977-03-01

    Controlled Thermonuclear Fusion offers probably the only relatively clean energy solution with completely inexhaustible fuel and unlimited power capacity. The scientific and technological problem consists in magnetically confining a hot, dense plasma (pressure several to hundreds of atmospheres, temperature 10 8 degrees or more) for an appreciable fraction of a second. The scientific and mathematical problem is to describe the behavior, such as confinement, stability, flow, compression, heating, energy transfer and diffusion of this medium in the presence of electromagnetic fields just as we now can for air or steam. Some of the extant theory consists of applications, routine or ingenious, of known mathematical structures in the theory of differential equations and in traditional analysis. Other applications of known mathematical structures offer surprises and new insights: the coordination between sub-supersonic and elliptic-hyperbolic is fractured; supersonic propagation goes upstream; etc. Other completely nonstandard mathematical structures with significant theory are being rapidly uncovered (and somewhat less rapidly understood) such as non-elliptic variational equations and new types of weak solutions. It is these new mathematical structures which one should expect to supply the foundation for the next generation's pure mathematics, if history is a guide. Despite the substantial effort over a period of some twenty years, there are still basic and important scintific and mathematical discoveries to be made, lying just beneath the surface

  20. Generalized Lawson Criteria for Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Tipton, Robert E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-08-27

    The Lawson Criterion was proposed by John D. Lawson in 1955 as a general measure of the conditions necessary for a magnetic fusion device to reach thermonuclear ignition. Over the years, similar ignition criteria have been proposed which would be suitable for Inertial Confinement Fusion (ICF) designs. This paper will compare and contrast several ICF ignition criteria based on Lawson’s original ideas. Both analytical and numerical results will be presented which will demonstrate that although the various criteria differ in some details, they are closely related and perform similarly as ignition criteria. A simple approximation will also be presented which allows the inference of each ignition parameter directly from the measured data taken on most shots fired at the National Ignition Facility (NIF) with a minimum reliance on computer simulations. Evidence will be presented which indicates that the experimentally inferred ignition parameters on the best NIF shots are very close to the ignition threshold.

  1. Electrostatic-Dipole (ED) Fusion Confinement Studies

    Science.gov (United States)

    Miley, George H.; Shrestha, Prajakti J.; Yang, Yang; Thomas, Robert

    2004-11-01

    The Electrostatic-Dipole (ED) concept significantly differs from a "pure" dipole confinement device [1] in that the charged particles are preferentially confined to the high-pressure region interior of the dipole coil by the assistance of a surrounding spherical electrostatic grid. In present ED experiments, a current carrying coil is embedded inside the grid of an IEC such as to produce a magnetic dipole field. Charged particles are injected axisymmetrically from an ion gun (or duo-plasmatron) into the center of the ED confinement grid/dipole ring where they oscillate along the magnetic field lines and pass the peak field region at the center of the dipole region. As particles begin accelerating away from the center region towards the outer electrostatic grid region, they encounter a strong electrostatic potential (order of 10's of kilovolts) retarding force. The particles then decelerate, reverse direction and re-enter the dipole field region where again magnetic confinement dominates. This process continues, emulating a complex harmonic oscillator motion. The resulting pressure profile averaged over the field curvature offers good plasma stability in the ED configuration. The basic concept and results from preliminary experiments will be described. [1] M.E. Mauel, et al. "Dipole Equilibrium and Stability," 18th IAEA Conference of Plasma Phys. and Control. Nuclear Fusion, Varenna, Italy 2000, IAEA-F1-CN-70/TH

  2. West European magnetic confinement fusion research

    International Nuclear Information System (INIS)

    McKenney, B.L.; McGrain, M.; Hogan, J.T.; Porkolab, M.; Thomassen, K.I.

    1990-01-01

    This report presents a technical assessment and review of the West European program in magnetic confinement fusion by a panel of US scientists and engineers active in fusion research. Findings are based on the scientific and technical literature, on laboratory reports and preprints, and on the personal experiences and collaborations of the panel members. Concerned primarily with developments during the past 10 years, from 1979 to 1989, the report assesses West European fusion research in seven technical areas: tokamak experiments; magnetic confinement technology and engineering; fusion nuclear technology; alternate concepts; theory; fusion computations; and program organization. The main conclusion emerging from the analysis is that West European fusion research has attained a position of leadership in the international fusion program. This distinction reflects in large measure the remarkable achievements of the Joint European Torus (JET). However, West European fusion prominence extends beyond tokamak experimental physics: the program has demonstrated a breadth of skill in fusion science and technology that is not excelled in the international effort. It is expected that the West European primacy in central areas of confinement physics will be maintained or even increased during the early 1990s. The program's maturity and commitment kindle expectations of dramatic West European advances toward the fusion energy goal. For example, achievement of fusion breakeven is expected first in JET, before 1995

  3. Review of Inertial Confinement Fusion

    Science.gov (United States)

    Haines, M. G.

    The physics of inertial confinement fusion is reviewed. The trend to short-wavelength lasers is argued, and the distinction between direct and indirect (soft X-ray) drive is made. Key present issues include the non-linear growth of Rayleigh-Taylor (R-T) instabilities, the seeding of this instability by the initial laser imprint, the relevance of self-generated magnetic fields, and the importance of parametric instabilities (stimulated Brillouin and Raman scattering) in gas-filled hohlraums. Experiments are reviewed which explore the R-T instability in both planar and converging geometry. The employment of various optical smoothing techniques is contrasted with the overcoating of the capsule by gold coated plastic foams to reduce considerably the imprint problem. The role of spontaneously generated magnetic fields in non-symmetric plasmas is discussed. Recent hohlraum compression results are presented together with gas bag targets which replicate the long-scale-length low density plasmas expected in NIF gas filled hohlraums. The onset of first Brillouin and then Raman scattering is observed. The fast ignitor scheme is a proposal to use an intense short pulse laser to drill a hole through the coronal plasma and then, with laser excited fast electrons, create a propagating thermonuclear spark in a dense, relatively cold laser-compressed target. Some preliminary results of laser hole drilling and 2-D and 3-D PIC simulations of this and the > 10^8 Gauss self-generated magnetic fields are presented. The proposed National Ignition Facility (NIF) is described.

  4. Measurement of inertial confinement fusion reaction rate

    International Nuclear Information System (INIS)

    Peng Xiaoshi; Wang Feng; Tang Daorun; Liu Shenye; Huang Tianxuan; Liu Yonggang; Xu Tao; Chen Ming; Mei Yu

    2011-01-01

    Fusion reaction rate is an important parameter for measuring compression during the implosion in inertial confinement fusion experiment. We have developed a system for fusion reaction history measurement with high temporal resolution. The system is composed of plastic scintillator and nose cone, optical system and streak camera. We have applied this system on the SG-III prototype for fusion reaction rate measuring. For the first time, fusion reaction rate history have been measured for deuterium-tritium filled targets with neutrons yields about 10 10 . We have analyzed possible influence factor during fusion reaction rate measuring. It indicates that the instrument measures fusion reaction bang time at temporal resolutions as low as 30 ps.(authors)

  5. Commercial applications of inertial confinement fusion

    International Nuclear Information System (INIS)

    Booth, L.A.; Frank, T.G.

    1977-05-01

    This report describes the fundamentals of inertial-confinement fusion, some laser-fusion reactor (LFR) concepts, and attendant means of utilizing the thermonuclear energy for commercial electric power generation. In addition, other commercial energy-related applications, such as the production of fissionable fuels, of synthetic hydrocarbon-based fuels, and of process heat for a variety of uses, as well as the environmental and safety aspects of fusion energy, are discussed. Finally, the requirements for commercialization of laser fusion technologies are described

  6. Maximum neutron yeidls in experimental fusion devices

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1979-02-01

    The optimal performances of 12 types of fusion devices are compared with regard to neutron production rate, neutrons per pulse, and fusion energy multiplication, Q/sub p/ (converted to the equivalent value in D-T operation). The record values in all categories are held by the beam-injected tokamak plasma, followed by other beam-target systems. The achieved values of Q/sub p/ for nearly all laboratory plasma fusion devices (magnetically or inertially confined) are found to roughly satisfy a common empirical scaling, Q/sub p/ approx. 10 -6 E/sub in//sup 3/2/, where E/sub in/ is the energy (in kilojoules) injected into the plasma during one or two energy confinement times, or the total energy delivered to the target for inertially confined systems. Fusion energy break-even (Q/sub p/ = 1) in any system apparently requires E/sub in/ approx. 10,000 kJ

  7. Cold nuclear fusion device

    International Nuclear Information System (INIS)

    Ogino, Shinji.

    1991-01-01

    Selection of cathode material is a key to the attainment of cold nuclear fusion. However, there are only few reports on the cathode material at present and an effective development has been demanded. The device comprises an anode and a cathode and an electrolytic bath having metal salts dissolved therein and containing heavy water in a glass container. The anode is made of gold or platinum and the cathode is made of metals of V, Sr, Y, Nb, Hf or Ta, and a voltage of 3-25V is applied by way of a DC power source between them. The metal comprising V, Sr, Y, Nb, Hf or Ta absorbs deuterium formed by electrolysis of heavy water effectively to cause nuclear fusion reaction at substantially the same frequency and energy efficiency as palladium and titanium. Accordingly, a cold nuclear fusion device having high nuclear fusion generation frequency can be obtained. (N.H.)

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

    Science.gov (United States)

    Fukuda, Takeshi

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

  9. Twenty years of ''Nuclear Fusion''. Inertial confinement

    International Nuclear Information System (INIS)

    Yamanaka, C.

    1980-01-01

    Inertial confinement (ICF) fusion research is directed towards demonstrating the feasibility of very rapidly heating and compressing small pellets of suitable fuel until conditions exist where thermonuclear fusion can occur and useful amounts of power can be produced. Major problems which have to be solved are the following: 1) pellet design based on driver-plasma coupling; 2) the technology of energy drivers; 3) feasibility of ICF reactor systems

  10. Inertial Confinement Fusion at Los Alamos

    International Nuclear Information System (INIS)

    Cartwright, D.C.

    1989-09-01

    This report discusses the following topics on Inertial Confinement Fusion: ICF contributions to science and technology; target fabrication; laser-target interaction; KrF laser development; advanced KrF lasers; KrF laser technology; and plasma physics for light-ion program

  11. Inertial Confinement Fusion at Los Alamos

    International Nuclear Information System (INIS)

    Cartwright, D.C.

    1989-09-01

    This report discusses the following topics on inertial confinement fusion: distribution of electron-beam energy in KrF laser media; electron collision processes in KrF laser media; Krf laser kinetics; and properties of the KrF laser medium

  12. Nuclear diagnostics for inertial confinement fusion implosions

    International Nuclear Information System (INIS)

    Murphy, T.J.

    1997-01-01

    This abstract contains viewgraphs on nuclear diagnostic techniques for inertial confinement fusion implosions. The viewgraphs contain information on: reactions of interest in ICF; advantages and disadvantages of these methods; the properties nuclear techniques can measure; and some specifics on the detectors used

  13. Hohlraum manufacture for inertial confinement fusion

    International Nuclear Information System (INIS)

    Foreman, L.R.; Gobby, P.; Bartos, J.

    1994-01-01

    Hohlraums are an integral part of indirect drive targets for Inertial Confinement Fusion (ICF) research. Hohlraums are made by an electroforming process that combines elements of micromachining and coating technology. The authors describe how these target element are made and extension of the method that allow fabrication of other, more complex target components

  14. Progress in high gain inertial confinement fusion

    International Nuclear Information System (INIS)

    Sun Jingwen

    2001-01-01

    The author reviews the progress in laboratory high gain inertial confinement fusion (ICF), including ICF capsule physics, high-energy-density science, inertial fusion energy, the National Ignition Facility (NIF) and its design of ignition targets and the peta watt laser breakthrough. High power laser, particle beam, and pulsed power facilities around the world have established the new laboratory field of high-energy- density plasma physics and have furthered development of inertial fusion. New capabilities such as those provided by high-brightness peta watt lasers have enabled the study of matter feasible in conditions previously unachievable on earth. Science and technology developed in inertial fusion research have found near-term commercial use and have enabled steady progress toward the goal of fusion ignition and high gain in the laboratory, and have opened up new fields of study for the 21 st century

  15. Magnetic confinement in plasmas in nuclear devices

    International Nuclear Information System (INIS)

    Tull, C.G.

    1979-01-01

    The main emphasis of the magnetic fusion energy research program today lies in the development of two types of confinement schemes: magnetic mirrors and tokamaks. Experimental programs for both of these confinement schemes have shown steady progress toward achieving fusion power breakeven. The scaling of the current machines to a reactor operating regime and newly developed methods for plasma heating will very likely produce power breakeven within the next decade. Predictions are that the efficiency in a fusion power plant should exceed 32%

  16. Process and device for energy production from thermonuclear fusion reactions

    International Nuclear Information System (INIS)

    Bussard, R.W.; Coppi, Bruno.

    1977-01-01

    An energy generating system is described using a fusion reaction. It includes several contrivances for confining a plasma in an area, a protective device around a significant part of each of these confinement contrivances, an appliance for introducing a fusion reaction fuel in each of the confinements so that the plasma may be formed. Each confinement can be separated from the protective device so that it may be replaced by another. The system is connected to the confinements, to the protective devices or to both. It enables the thermal energy to be extracted and transformed into another form, electric, mechanical or both [fr

  17. 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

  18. Ceramics for fusion devices

    International Nuclear Information System (INIS)

    Clinard, F.W. Jr.

    1984-01-01

    Ceramics are required for a number of applications in fusion devices, among the most critical of which are magnetic coil insulators, windows for RF heating systems, and structural uses. Radiation effects dominate consideration of candidate materials, although good pre-irradiation properties are a requisite. Materials and components can be optimized by careful control of chemical and microstructural content, and application of brittle material design and testing techniques. Future directions for research and development should include further extension of the data base in the areas of electrical, structural, and thermal properties; establishment of a fission neutron/fusion neutron correlation including transmutation gas effects; and development of new materials tailored to meet the specific needs of fusion reactors

  19. Data-Acquisition Systems for Fusion Devices

    NARCIS (Netherlands)

    van Haren, P. C.; Oomens, N. A.

    1993-01-01

    During the last two decades, computerized data acquisition systems (DASs) have been applied at magnetic confinement fusion devices. Present-day data acquisition is done by means of distributed computer systems and transient recorders in CAMAC systems. The development of DASs has been technology

  20. 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

  1. 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)

  2. Laser drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Holzrichter, J.F.

    1983-01-01

    Inertial Confinement Fusion (ICF) is the technology that we are developing to access the vast stored energy potential of deuterium fuel located in the world's water supply. This form of fusion is accomplished by compressing and heating small volumes of D-T fuel to very high temperatures (greater than 100M 0 C) and to very high densities (greater than 1000 times the normal liquid density). Under these fuel conditions, a thermonuclear reaction can occur, leading to a net energy release compared to the energy used to heat the fuel initially. To accomplish the condition where fusion reactions begin, effective drivers are required. These are lasers or particle beam accelerators which can provide greater than 10 14 W/cm 2 over millimeter scale targets with an appropriately programmed intensity vs time. At present, we are using research lasers to obtain an understanding of the physics and engineering of fuel compression

  3. Inertial confinement fusion and related topics

    International Nuclear Information System (INIS)

    Starodub, A. N.

    2007-01-01

    The current state of different approaches (laser fusion, light and heavy ions, electron beam) to the realization of inertial confinement fusion is considered. From comparative analysis a conclusion is made that from the viewpoint of physics, technology, safety, and economics the most realistic way to future energetics is an electric power plant based on a hybrid fission-fusion reactor which consists of an external source of neutrons (based on laser fusion) and a subcritical two-cascade nuclear blanket, which yields the energy under the action of 14 MeV neutrons. The main topics on inertial confinement fusion such as the energy driver, the interaction between plasmas and driver beam, the target design are discussed. New concept of creation of a laser driver for IFE based on generation and amplification of radiation with controllable coherence is reported. The performed studies demonstrate that the laser based on generation and amplification of radiation with controllable coherence (CCR laser) has a number of advantages as compared to conventional schemes of lasers. The carried out experiments have shown a possibility of suppression of small-scale self-focusing, formation of laser radiation pulses with required characteristics, simplification of an optical scheme of the laser, good matching of laser-target system and achievement of homogeneous irradiation and high output laser energy density without using traditional correcting systems (phase plates, adaptive optics, space filters etc.). The results of the latest experiments to reach ultimate energy characteristics of the developed laser system are also reported. Recent results from the experiments aimed at studying of the physical processes in targets under illumination by the laser with controllable coherence of radiation are presented and discussed, especially such important laser-matter interaction phenomena as absorption and scattering of the laser radiation, the laser radiation harmonic generation, X

  4. Magnetic confinement fusion plasma theory, Task 1

    International Nuclear Information System (INIS)

    Callen, J.D.

    1991-07-01

    The research performed under this grant during the current year has concentrated on a few key tokamak plasma confinement and heating theory issues: extensive development of a new Chapman-Enskog-like fluid/kinetic hybrid approach to deriving rigorously valid fluid moment equations; applications (neoclassical viscous force, instabilities in the banana-plateau collisionality regime, nonlinear gyroviscous force, unified plasma microinstability equations and their implications, semi-collisional presheath modeling, etc.) of this new formalism; interactions of fluctuating bootstrap-current-driven magnetic islands; determination of net transport processes and equations for a tokamak; and some other topics (extracting more information from heat-pulse-propagation data, modeling of BES fluctuation data, exploring sawtooth effects on energy confinement in DIII-D, divertor X-point modeling). Recent progress and publications in these areas, and in the management of the local NERSC node and fusion theory DECstation 5000 at UW-Madison are summarized briefly in this report

  5. Hydrodynamic instabilities in inertial confinement fusion

    International Nuclear Information System (INIS)

    Hoffman, N.M.

    1995-01-01

    The focus of these (two) lectures is on buoyancy-driven instabilities of the Rayleigh-Taylor type, which are commonly regarded as the most important kind of hydrodynamic instability in inertial-confinement-fusion implosions. The paper is intended to be pedagogical rather than research-oriented, and so is by no means a comprehensive review of work in this field. Rather, it is hoped that the student will find here a foundation on which to build an understanding of current research, and the experienced researcher will find a compilation of useful results. (author)

  6. 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

  7. Pulsed power systems for inertial confinement fusion

    International Nuclear Information System (INIS)

    VanDevender, J.P.

    1979-01-01

    Sandis's Particle Beam Fusion Program is investigating pulsed electron and light ion beam accelerators with the goal of demonstrating the practical application of such drivers as igniters in inertial confinement fusion (ICF) reactors. The power and energy requirements for net energy gain are 10 14 to 10 15 W and 1 to 10 MJ. Recent advances in pulsed power and power flow technologies permit suitable accelerators to be built. The first accelerator of this new generation is PBFA I. It operates at 2 MV, 15 MA, 30 TW for 35 ns and is scheduled for completion in June 1980. The principles of this new accelerator technology and their application to ICF will be presented

  8. Hydrodynamic instabilities in inertial confinement fusion

    International Nuclear Information System (INIS)

    Freeman, J.R.

    1977-01-01

    Inertial confinement fusion targets generally consist of hollow high-density spheres filled with low density thermonuclear fuel. Targets driven ablatively by electrons, ions, or lasers are potentially unstable during the initial acceleration phase. Later in time, the relatively low density fuel decelerates the dense inner portion of the sphere (termed the pusher), permitting unstable growth at the fuel-pusher interface. The instabilities are of the Rayleigh-Taylor variety, modified by thermal and viscous diffusion and convection. These problems have been analyzed by many in recent years using both linearized perturbation methods and direct numerical simulation. Examples of two-dimensional simulations of the fuel-pusher instability in electron beam fusion targets will be presented, along with a review of possible stabilization mechanisms

  9. Definition of Ignition in Inertial Confinement Fusion

    Science.gov (United States)

    Christopherson, A. R.; Betti, R.

    2017-10-01

    Defining ignition in inertial confinement fusion (ICF) is an unresolved problem. In ICF, a distinction must be made between the ignition of the hot spot and the propagation of the burn wave in the surrounding dense fuel. Burn propagation requires that the hot spot is robustly ignited and the dense shell exhibits enough areal density. Since most of the energy gain comes from burning the dense shell, in a scale of increasing yields, hot-spot ignition comes before high gains. Identifying this transition from hot-spot ignition to burn-wave propagation is key to defining ignition in general terms applicable to all fusion approaches that use solid DT fuel. Ad hoc definitions such as gain = 1 or doubling the temperature are not generally valid. In this work, we show that it is possible to identify the onset of ignition through a unique value of the yield amplification defined as the ratio of the fusion yield including alpha-particle deposition to the fusion yield without alphas. Since the yield amplification is a function of the fractional alpha energy fα =EαEα 2Ehs 2Ehs (a measurable quantity), it appears possible not only to define ignition but also to measure the onset of ignition by the experimental inference of the fractional alpha energy and yield amplification. This material is based upon work supported by the Department of Energy Office of Fusion Energy Services under Award Number DE-FC02-04ER54789 and National Nuclear Security Administration under Award Number DE-NA0001944.

  10. 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.

  11. 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

  12. 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

  13. Inertial electrostatic confinement (IEC) fusion fundamentals and applications

    CERN Document Server

    Miley, George H

    2014-01-01

    This book provides readers with an introductory understanding of Inertial Electrostatic Confinement (IEC), a type of fusion meant to retain plasma using an electrostatic field. IEC provides a unique approach for plasma confinement, as it offers a number of spin-off applications, such as a small neutron source for Neutron Activity Analysis (NAA), that all work towards creating fusion power. The IEC has been identified in recent times as an ideal fusion power unit because of its ability to burn aneutronic fuels like p-B11 as a result of its non-Maxwellian plasma dominated by beam-like ions. This type of fusion also takes place in a simple mechanical structure small in size, which also contributes to its viability as a source of power. This book posits that the ability to study the physics of IEC in very small volume plasmas makes it possible to rapidly investigate a design to create a power-producing device on a much larger scale. Along with this hypothesis the book also includes a conceptual experiment propose...

  14. Intense ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    Mehlhorn, T.A.

    1997-01-01

    Intense beams of light of heavy ions are being studied as inertial confinement fusion (ICF) drivers for high yield and energy. Heavy and light ions have common interests in beam transport, targets, and alternative accelerators. Self-pinched transport is being jointly studied. This article reviews the development of intense ion beams for ICF. Light-ion drivers are highlighted because they are compact, modular, efficient and low cost. Issues facing light ions are: (1) decreasing beam divergence; (2) increasing beam brightness; and (3) demonstrating self-pinched transport. Applied-B ion diodes are favored because of efficiency, beam brightness, perceived scalability, achievable focal intensity, and multistage capability. A light-ion concept addressing these issues uses: (1) an injector divergence of ≤ 24 mrad at 9 MeV; (2) two-stage acceleration to reduce divergence to ≤ 12 mrad at 35 MeV; and (3) self-pinched transport accepting divergences up to 12 mrad. Substantial progress in ion-driven target physics and repetitive ion diode technology is also presented. Z-pinch drivers are being pursued as the shortest pulsed power path to target physics experiments and high-yield fusion. However, light ions remain the pulsed power ICF driver of choice for high-yield fusion energy applications that require driver standoff and repetitive operation. 100 refs

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

    International Nuclear Information System (INIS)

    Callen, J.D.

    1993-01-01

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

  16. An experiment to test centrifugal confinement for fusion

    International Nuclear Information System (INIS)

    Ellis, R.F.; Hassam, A.B.; Messer, S.; Osborn, B.R.

    2001-01-01

    The basic idea of centrifugal confinement is to use centrifugal forces from supersonic rotation to augment conventional magnetic confinement. Optimizing this 'knob' results in a fusion device that features four advantages: steady state, no disruptions, superior cross-field confinement, and a simpler coil configuration. The idea rests on two prongs: first, centrifugal forces can confine plasmas to desired regions of shaped magnetic fields; second, the accompanying large velocity shear can stabilize even magnetohydrodynamic (MHD) instabilities. A third feature is that the velocity shear also viscously heats the plasma; no auxiliary heating is necessary to reach fusion temperatures. Regarding transport, the velocity shear can also quell microturbulence, leading to fully classical confinement, as there are no neoclassical effects. Classical parallel electron transport then sets the confinement time. These losses are minimized by a large Pastukhov factor resulting from the deep centrifugal potential well: at Mach 4-5, the Lawson criterion is accessible. One key issue is whether velocity shear will be sufficient by itself to stabilize MHD interchanges. Numerical simulations indicate that laminar equilibria can be obtained at Mach numbers of 4-5 but that the progression toward laminarity with increasing Mach number is accompanied by residual convection from the interchanges. The central goal of the Maryland Centrifugal Torus (MCT) [R. F. Ellis et al., Bull. Am. Phys. Soc. 44, 48 (1998)] is to obtain MHD stability from velocity shear. As an assist to accessing laminarity, MCT will incorporate two unique features: plasma elongation and toroidal magnetic field. The former raises velocity shear efficiency, and modest magnetic shear should suppress residual convection

  17. Elmo bumpy square plasma confinement device

    Science.gov (United States)

    Owen, L.W.

    1985-01-01

    The invention is an Elmo bumpy type plasma confinement device having a polygonal configuration of closed magnet field lines for improved plasma confinement. In the preferred embodiment, the device is of a square configuration which is referred to as an Elmo bumpy square (EBS). The EBS is formed by four linear magnetic mirror sections each comprising a plurality of axisymmetric assemblies connected in series and linked by 90/sup 0/ sections of a high magnetic field toroidal solenoid type field generating coils. These coils provide corner confinement with a minimum of radial dispersion of the confined plasma to minimize the detrimental effects of the toroidal curvature of the magnetic field. Each corner is formed by a plurality of circular or elliptical coils aligned about the corner radius to provide maximum continuity in the closing of the magnetic field lines about the square configuration confining the plasma within a vacuum vessel located within the various coils forming the square configuration confinement geometry.

  18. Fast ignition schemes for inertial confinement fusion

    International Nuclear Information System (INIS)

    Deutsch, C.

    2003-01-01

    The controlled production of a local hot spot in super-compressed deuterium + tritium fuel is examined in details. Relativistic electron beams (REB) in the MeV and proton beams in the few tens MeV energy range produced by PW-lasers are respectively considered. A strong emphasis is given to the propagation issues due to large density gradients in the outer core of compressed fuel. A specific attention is also paid to the final and complete particle stopping resulting in hot spot generation as well as to the interplay of collective vs. particle stopping at the entrance channel on the low density side in plasma target. Moreover, REB production and fast acceleration mechanisms are also given their due attention. Proton fast ignition looks promising as well as the wedged (cone angle) approach circumventing most of transport uncertainties between critical layer and hot spot. Global engineering perspectives for fast ignition scenario (FIS) driven inertial confinement fusion are also detailed. (author)

  19. Open-ended fusion devices and reactors

    International Nuclear Information System (INIS)

    Kawabe, T.; Nariai, H.

    1983-01-01

    Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown

  20. Inertial confinement fusion with light ion beams

    International Nuclear Information System (INIS)

    VanDevender, J.P.; Cook, D.L.

    1986-01-01

    The Particle Beam Fusion Accelerator II (PBFA II) is presently under construction and is the only existing facility with the potential of igniting thermonuclear fuel in the laboratory. The accelerator will generate up to 5 megamperes of lithium ions at 30 million electron volts and will focus them onto an inertial confinement fusion (ICF) target after beam production and focusing have been optimized. Since its inception, the light ion approach to ICF has been considered the one that combines low cost, high risk, and high payoff. The beams are of such high density that their self-generated electric and magnetic fields were thought to prohibit high focal intensities. Recent advances in beam production and focusing demonstrate that these self-forces can be controlled to the degree required for ignition, break-even, and high gain experiments. ICF has been pursued primarily for its potential military applications. However, the high efficiency and cost-effectiveness of the light ion approach enhance its potential for commercial energy application as well

  1. Thermonuclear plasma physic: inertial confinement fusion

    International Nuclear Information System (INIS)

    Bayer, Ch.; Juraszek, D.

    2001-01-01

    Inertial Confinement Fusion (ICF) is an approach to thermonuclear fusion in which the fuel contained in a spherical capsule is strongly compressed and heated to achieve ignition and burn. The released thermonuclear energy can be much higher than the driver energy, making energetic applications attractive. Many complex physical phenomena are involved by the compression process, but it is possible to use simple analytical models to analyze the main critical points. We first determine the conditions to obtain fuel ignition. High thermonuclear gains are achieved if only a small fraction of the fuel called hot spot is used to trigger burn in the main fuel compressed on a low isentrope. A simple hot spot model will be described. The high pressure needed to drive the capsule compression are obtained by the ablation process. A simple Rocket model describe the main features of the implosion phase. Several parameters have to be controlled during the compression: irradiation symmetry, hydrodynamical stability and when the driver is a laser, the problems arising from interaction of the EM wave with the plasma. Two different schemes are examined: Indirect Drive which uses X-ray generated in a cavity to drive the implosion and the Fast Ignitor concept using a ultra intense laser beam to create the hot spot. At the end we present the Laser Megajoule (LMJ) project. LMJ is scaled to a thermonuclear gain of the order of ten. (authors)

  2. The role of alpha particles in magnetically confined fusion plasmas

    International Nuclear Information System (INIS)

    Lisak, M.; Wilhelmsson, H.

    1986-01-01

    Recent progress in the confinement of hot plasmas in magnetic fusion experiments throughout the world has intensified interest and research in the physics of D-T burning plasmas especially in the wide range of unresolved theoretical as well as experimental questions associated with the role of alpha particles in such devices. In order to review the state-of-the- art in this field, and to identify new issues and problems for further research, the Symposium on the Role of Alpha Particles in Magnetically Confined Fusion Plasmas was held from 24 to 26 June 1986 at Aspenaesgaarden near Goeteborg, Sweden. About 25 leading experts from nine countries attended the Symposium and gave invited talks. The major part of the programme was devoted to alpha-particle effects in tokamaks but some aspects of open systems were also discussed. The possibilities of obtaining ignition in JET and TFTR as well as physics issues for the compact ignition experiments were considered in particular. A special session was devoted to the diagnostics of alpha particles and other fusion products. In this report are summarised some of the highlights of the symposium. (authors)

  3. Inertial Electrostatic Confinement (IEC) devices

    International Nuclear Information System (INIS)

    Nebel, R.A.; Turner, L.; Tiouririne, T.N.; Barnes, D.C.; Nystrom, W.D.; Bussard, R.W.; Miley, G.H.; Javedani, J.; Yamamoto, Y.

    1994-01-01

    Inertial Electrostatic Confinement (IEC) is one of the earliest plasma confinement concepts, having first been suggested by P. T. Farnsworth in the 1950s. The concept involves a simple apparatus of concentric spherical electrostatic grids or a combination of grids and magnetic fields. An electrostatic structure is formed from the confluence of electron or ion beams. Gridded IEC systems have demonstrated neutron yields as high as 2 * 10 10 neutrons/sec. These systems have considerable potential as small, inexpensive, portable neutron sources for assaying applications. Neutron tomography is also a potential application. Atomic physics effects strongly influence the performance of all of these systems. Important atomic effects include elastic scattering, ionization, excitation, and charge exchange. This paper discusses how an IEC system is influenced by these effects and how to design around them. Theoretical modeling and experimental results are presented

  4. Arcing phenomena in fusion devices workshop

    International Nuclear Information System (INIS)

    Clausing, R.E.

    1979-01-01

    The workshop on arcing phenomena in fusion devices was organized (1) to review the pesent status of our understanding of arcing as it relates to confinement devices, (2) to determine what informaion is needed to suppress arcing and (3) to define both laboratory and in-situ experiments which can ultimately lead to reduction of impurities in the plasma caused by arcing. The workshop was attended by experts in the area of vacuum arc electrode phenomena and ion source technology, materials scientists, and both theoreticians and experimentalists engaged in assessing the importance of unipolar arcing in today's tokamaks. Abstracts for papers presented at the workshop are included

  5. Beam dancer fusion device

    International Nuclear Information System (INIS)

    Maier, H.B.

    1984-01-01

    To accomplish fusion of two or more fusion fuel elements numerous minute spots of energy or laser light are directed to a micro target area, there to be moved or danced about by a precision mechanical controlling apparatus at the source of the laser light or electromagnetic energy beams, so that merging and coinciding patterns of light or energy beams can occur around the area of the fuel atoms or ions. The projecting of these merging patterns may be considered as target searching techniques to locate responsive clusters of fuel elements and to compress such elements into a condition in which fusion may occur. Computerized programming may be used

  6. Control of Internal Transport Barriers in Magnetically Confined Fusion Plasmas

    Science.gov (United States)

    Panta, Soma; Newman, David; Sanchez, Raul; Terry, Paul

    2016-10-01

    In magnetic confinement fusion devices the best performance often involves some sort of transport barriers to reduce the energy and particle flow from core to edge. Those barriers create gradients in the temperature and density profiles. If gradients in the profiles are too steep that can lead to instabilities and the system collapses. Control of these barriers is therefore an important challenge for fusion devices (burning plasmas). In this work we focus on the dynamics of internal transport barriers. Using a simple 7 field transport model, extensively used for barrier dynamics and control studies, we explore the use of RF heating to control the local gradients and therefore the growth rates and shearing rates for barrier initiation and control in self-heated fusion plasmas. Ion channel barriers can be formed in self-heated plasmas with some NBI heating but electron channel barriers are very sensitive. They can be formed in self-heated plasmas with additional auxiliary heating i.e. NBI and radio-frequency(RF). Using RF heating on both electrons and ions at proper locations, electron channel barriers along with ion channel barriers can be formed and removed demonstrating a control technique. Investigating the role of pellet injection in controlling the barriers is our next goal. Work supported by DOE Grant DE-FG02-04ER54741.

  7. Electromagnetic computations for fusion devices

    International Nuclear Information System (INIS)

    Turner, L.R.

    1989-09-01

    Among the difficulties in making nuclear fusion a useful energy source, two important ones are producing the magnetic fields needed to drive and confine the plasma, and controlling the eddy currents induced in electrically conducting components by changing fields. All over the world, researchers are developing electromagnetic codes and employing them to compute electromagnetic effects. Ferromagnetic components of a fusion reactor introduce field distortions. Eddy currents are induced in the vacuum vessel, blanket and other torus components of a tokamak when the plasma current disrupts. These eddy currents lead to large forces, and 3-D codes are being developed to study the currents and forces. 35 refs., 6 figs

  8. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohnishi, Masami, E-mail: onishi@kansai-u.ac.jp [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Yamamoto, Yasushi; Osawa, Hodaka [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Hatano, Yuji; Torikai, Yuji [Hydrogen Isotope Science Center, University of Toyama, Gofuku, Toyama 930-8555 (Japan); Murata, Isao [Faculty of Engineering Environment and Energy Department, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Masuda, Kai [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hotta, Eiki [Interdisciplinary Graduate School of Science and Engineering, Tokyo Institute of Technology, 4259 Nagatsuda-cho, Midori-ku, Yokohama 226-8503 (Japan)

    2016-11-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10{sup 8}(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  9. Tritium burning in inertial electrostatic confinement fusion facility

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Osawa, Hodaka; Hatano, Yuji; Torikai, Yuji; Murata, Isao; Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki; Masuda, Kai; Hotta, Eiki

    2016-01-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10"8(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  10. Inertial confinement fusion and fast ignitor studies

    International Nuclear Information System (INIS)

    Willi, O.; Barringer, L.; Bell, A.

    1999-01-01

    The paper discusses inertial confinement fusion research carried out at several different laser facilities including the VULCAN laser at the Rutherford Appleton Laboratory, the TRIDENT laser at the Los Alamos National Laboratory and the PHEBUS laser at Limeil. Low density foam targets were irradiated either with nanosecond laser or soft x-ray pulses. Laser imprinting was studied and in particular saturation of areal density perturbations induced by near-single mode laser imprinting has been observed. Several issues important for the foam buffered direct drive scheme were investigated. These studies included measurements of the absolute levels of Stimulated Brillouin and Raman Scattering observed from laser irradiated low density foam targets either bare or overcoated with a thin layer of gold. A novel scheme is proposed to increase the pressure in indirectly driven targets. Low density foams that are mounted onto a foil target are heated with an intense pulse of soft x-ray radiation. If the foam is heated supersonically the pressure generated is not only the ablation pressure but the combined pressure due to ablation at the foam/foil interface and the heated foam material. The scheme was confirmed on planar targets. Brominated foil targets overcoated with a low density foam were irradiated by a soft x-ray pulse emitted from a hohlraum. The pressure was obtained by comparing the rear side trajectory of the driven target observed by soft x-ray radiography to one dimensional radiation hydrodynamic simulations. Further, measurements were carried out to observe the transition from super- to subsonic propagation of an ionisation front in low density chlorinated foam targets irradiated by an intense soft x-ray pulse both in open and confined geometry. The diagnostic for these measurements was K-shell point projection absorption spectroscopy. In the fast ignitor area the channeling and guiding of picosecond laser pulses through underdense plasmas, preformed density

  11. Inertial confinement fusion and fast ignitor studies

    International Nuclear Information System (INIS)

    Willi, O.; Barringer, L.; Bell, A.

    2001-01-01

    The paper discusses inertial confinement fusion research carried out at several different laser facilities including the VULCAN laser at the Rutherford Appleton Laboratory, the TRIDENT laser at the Los Alamos National Laboratory and the PHEBUS laser at Limeil. Low density foam targets were irradiated either with nanosecond laser or soft x-ray pulses. Laser imprinting was studied and in particular saturation of areal density perturbations induced by near-single mode laser imprinting has been observed. Several issues important for the foam buffered direct drive scheme were investigated. These studies included measurements of the absolute levels of Stimulated Brillouin and Raman Scattering observed from laser irradiated low density foam targets either bare or overcoated with a thin layer of gold. A novel scheme is proposed to increase the pressure in indirectly driven targets. Low density foams that are mounted onto a foil target are heated with an intense pulse of soft x-ray radiation. If the foam is heated supersonically the pressure generated is not only the ablation pressure but the combined pressure due to ablation at the foam/foil interface and the heated foam material. The scheme was confirmed on planar targets. Brominated foil targets overcoated with a low density foam were irradiated by a soft x-ray pulse emitted from a hohlraum. The pressure was obtained by comparing the rear side trajectory of the driven target observed by soft x-ray radiography to one dimensional radiation hydrodynamic simulations. Further, measurements were carried out to observe the transition from super- to subsonic propagation of an ionisation front in low density chlorinated foam targets irradiated by an intense soft x-ray pulse both in open and confined geometry. The diagnostic for these measurements was K-shell point projection absorption spectroscopy. In the fast ignitor area the channeling and guiding of picosecond laser pulses through underdense plasmas, preformed density

  12. Plan for the development and commercialization of inertial confinement fusion

    International Nuclear Information System (INIS)

    Willke, T.; Dingee, D.; Ault, L.; Bampton, M.; Bickford, W.; Hartman, J.; Rockwood, A.; Simonen, E.; Teofilo, V.; Frank, T.

    1978-01-01

    An engineering development program strategy to take inertial confinement fusion (ICF) from the milestone of scientific feasibility to a point where its commercial viability can be determined is described. The ICF program objectives and basic program strategy are discussed

  13. Polarization beam smoothing for inertial confinement fusion

    International Nuclear Information System (INIS)

    Rothenberg, Joshua E.

    2000-01-01

    For both direct and indirect drive approaches to inertial confinement fusion (ICF) it is imperative to obtain the best possible drive beam uniformity. The approach chosen for the National Ignition Facility uses a random-phase plate to generate a speckle pattern with a precisely controlled envelope on target. A number of temporal smoothing techniques can then be employed to utilize bandwidth to rapidly change the speckle pattern, and thus average out the small-scale speckle structure. One technique which generally can supplement other smoothing methods is polarization smoothing (PS): the illumination of the target with two distinct and orthogonally polarized speckle patterns. Since these two polarizations do not interfere, the intensity patterns add incoherently, and the rms nonuniformity can be reduced by a factor of (√2). A number of PS schemes are described and compared on the basis of the aggregate rms and the spatial spectrum of the focused illumination distribution. The (√2) rms nonuniformity reduction of PS is present on an instantaneous basis and is, therefore, of particular interest for the suppression of laser plasma instabilities, which have a very rapid response time. When combining PS and temporal methods, such as smoothing by spectral dispersion (SSD), PS can reduce the rms of the temporally smoothed illumination by an additional factor of (√2). However, it has generally been thought that in order to achieve this reduction of (√2), the increased divergence of the beam from PS must exceed the divergence of SSD. It is also shown here that, over the time scales of interest to direct or indirect drive ICF, under some conditions PS can reduce the smoothed illumination rms by nearly (√2) even when the PS divergence is much smaller than that of SSD. (c) 2000 American Institute of Physics

  14. Antiproton fast ignition for inertial confinement fusion

    International Nuclear Information System (INIS)

    Perkins, L.J.

    1999-01-01

    With 180 MJ/microg, antiprotons offer the highest stored energy per unit mass of any known entity. The use of antiprotons to promote fast ignition in an inertial confinement fusion (ICF) capsule and produce high target gains with only modest compression of the main fuel is investigated. Unlike standard fast ignition where the ignition energy is supplied by energetic, short pulse laser, the energy here is supplied through the ionization energy deposited when antiprotons annihilate at the center of a compressed fuel capsule. This can be considered in-situ fast ignition as it obviates the need for the external injection of the ignition energy. In the first of two candidate schemes, the antiproton package is delivered by a low-energy ion beam. In the second, autocatalytic scheme, the antiprotons are preemplaced at the center of the capsule prior to compression. In both schemes, the author estimates that ∼10 12 antiprotons are required to initiate fast ignition in a typical ICF capsule and show that incorporation of a thin, heavy metal shell is desirable to enhance energy deposition within the ignitor zone. In addition to eliminating the need for a second, energetic fast laser and vulnerable final optics, this scheme would achieve central ignition without reliance on laser channeling through halo plasma or Hohlraum debris. However, in addition to the practical difficulties of storage and manipulation of antiprotons at low energy, the other large uncertainty for the practicality of such a speculative scheme is the ultimate efficiency of antiproton production in an external, optimized facility. Estimates suggest that the electrical wall plug energy per pulse required for the separate production of the antiprotons is of the same order as that required for the conventional slow compression driver

  15. Comparative assessment of world research efforts on magnetic confinement fusion

    International Nuclear Information System (INIS)

    McKenney, B.L.; McGrain, M.; Rutherford, P.H.

    1990-02-01

    This report presents a comparative assessment of the world's four major research efforts on magnetic confinement fusion, including a comparison of the capabilities in the Soviet Union, the European Community (Western Europe), Japan, and the United States. A comparative evaluation is provided in six areas: tokamak confinement; alternate confinement approaches; plasma technology and engineering; and fusion computations. The panel members are involved actively in fusion-related research, and have extensive experience in previous assessments and reviews of the world's four major fusion programs. Although the world's four major fusion efforts are roughly comparable in overall capabilities, two conclusions of this report are inescapable. First, the Soviet fusion effort is presently the weakest of the four programs in most areas of the assessment. Second, if present trends continue, the United States, once unambiguously the world leader in fusion research, will soon lose its position of leadership to the West European and Japanese fusion programs. Indeed, before the middle 1990s, the upgraded large-tokamak facilities, JT-60U (Japan) and JET (Western Europe), are likely to explore plasma conditions and operating regimes well beyond the capabilities of the TFTR tokamak (United States). In addition, if present trends continue in the areas of fusion nuclear technology and materials, and plasma technology and materials, and plasma technology development, the capabilities of Japan and Western Europe in these areas (both with regard to test facilities and fusion-specific industrial capabilities) will surpass those of the United States by a substantial margin before the middle 1990s

  16. High-density-plasma diagnostics in magnetic-confinement fusion

    International Nuclear Information System (INIS)

    Jahoda, F.C.

    1982-01-01

    The lectures will begin by defining high density in the context of magnetic confinement fusion research and listing some alternative reactor concepts, ranging from n/sub e/ approx. 2 x 10 14 cm -3 to several orders of magnitude greater, that offer potential advantages over the main-line, n/sub e/ approx. 1 x 10 14 cm -3 , Tokamak reactor designs. The high density scalings of several major diagnostic techniques, some favorable and some disadvantageous, will be discussed. Special emphasis will be given to interferometric methods, both electronic and photographic, for which integral n/sub e/dl measurements and associated techniques are accessible with low wavelength lasers. Reactor relevant experience from higher density, smaller dimension devices exists. High density implies high β, which implies economies of scale. The specialized features of high β diagnostics will be discussed

  17. The sensitivity theory for inertial confinement pellet fusion system

    International Nuclear Information System (INIS)

    Cai, Shaohui; Zhang, Yuquan.

    1986-01-01

    A sensitivity theory for inertial confinement pellet fusion system is developed based on a physical model similar to that embodied in the laser fusion code MEDUSA. The theory presented here can be an efficient tool for estimating the effects of many alternations in the data field. Our result is different from Greenspan's work in 1980. (author)

  18. Microencapsulation and fabrication of fuel pellets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Nolen, R.L. Jr.; Kool, L.B.

    1981-01-01

    Various microencapsulation techniques were evaluated for fabrication of thermonuclear fuel pellets for use in existing experimental facilities studying inertial confinement fusion and in future fusion-power reactors. Coacervation, spray drying, in situ polymerization, and physical microencapsulation methods were employed. Highly spherical, hollow polymeric shells were fabricated ranging in size from 20 to 7000 micron. In situ polymerization microencapsulation with poly(methyl methacrylate) provided large shells, but problems with local wall defects still must be solved. Extension to other polymeric systems met with limited success. Requirements for inertial confinement fusion targets are described, as are the methods that were used

  19. Fusion and technology: An introduction to the physics and technology of magnetic confinment fusion

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1984-01-01

    This book is an introduction covering all aspects of magnetic fusion and magnetic fusion technology. Physical property data relevant to fusion technology and a summary of fusion reactor design parameters are provided. Topics covered include: basic properties; equilibrium and transport confinement concepts; plasma heating; plasma wall interaction; magnetics; energy storage and transfer; interaction of radiation with matter; primary energy conversion and tritium breeding blanket; tritium and vacuum; and Fusion Reactor Design

  20. Scaling laws for inertial confinement fusion

    International Nuclear Information System (INIS)

    Brueckner, K.A.

    1978-01-01

    The fusion yield due to a spherically expanding burning front in a compressed fuel pellet is obtained. The pellet gain and beam energy for a laser system and an ion or electron beam driven system are compared. The results suggest an interesting possibility for heavy-ion fusion with driver parameters far below those usually considered

  1. Confinement inertial fusion. Power reactors of nuclear fusion by lasers

    International Nuclear Information System (INIS)

    Velarde, G.; Ahnert, C.; Aragones, J.M.; Leira, G; Martinez-Val, J.M.

    1980-01-01

    The energy crisis and the need of the nuclear fusion energy are analized. The nuclear processes in the laser interation with the ablator material are studied, as well as the thermohydrodinamic processes in the implossion, and the neutronics of the fusion. The fusion reactor components are described and the economic and social impact of its introduction in the future energetic strategies.(author)

  2. Inertial-confinement fusion with lasers

    International Nuclear Information System (INIS)

    Betti, R.; Hurricane, O. A.

    2016-01-01

    The quest for controlled fusion energy has been ongoing for over a half century. The demonstration of ignition and energy gain from thermonuclear fuels in the laboratory has been a major goal of fusion research for decades. Thermonuclear ignition is widely considered a milestone in the development of fusion energy, as well as a major scientific achievement with important applications to national security and basic sciences. The U.S. is arguably the world leader in the inertial con fment approach to fusion and has invested in large facilities to pursue it with the objective of establishing the science related to the safety and reliability of the stockpile of nuclear weapons. Even though significant progress has been made in recent years, major challenges still remain in the quest for thermonuclear ignition via laser fusion

  3. Toroidal nuclear fusion device

    International Nuclear Information System (INIS)

    Ito, Yutaka; Kasahara, Tatsuo; Takizawa, Teruhiro.

    1975-01-01

    Object: To design a device so as to be formed into a large-size and to arrange ports, through which neutral particles enter, in inclined fashion. Structure: Toroidal coils are wound about vacuum vessels which are divided into plural number. In the outer periphery of the vacuum vessels, ports are disposed inclined in the peripheral direction of the vacuum vessels and communicated with the vacuum vessels, and wall surfaces opposed to the ports of the toroidal coils adjacent at least the inclined sides of the ports are inclined substantially simularly to the port wall surfaces. (Kamimura, M.)

  4. Forty years of inertial confinement fusion research in the Shanghai Institute of Optics and Fine Mechanics

    International Nuclear Information System (INIS)

    Chen Chongbin; Wang Letian

    2010-01-01

    On the basis of China's own technology and industry, the 'Shenguang' inertial confinement fusion (ICF) research devices were built, and a series of world-class results achieved. In this paper, the history of ICF research in the Shanghai Institute of Optics and Fine Mechanics is reviewed. (authors)

  5. Thermonuclear plasma physic: inertial confinement fusion; Physique des plasmas thermonucleaires: la fusion par confinement inertiel

    Energy Technology Data Exchange (ETDEWEB)

    Bayer, Ch.; Juraszek, D

    2001-07-01

    Inertial Confinement Fusion (ICF) is an approach to thermonuclear fusion in which the fuel contained in a spherical capsule is strongly compressed and heated to achieve ignition and burn. The released thermonuclear energy can be much higher than the driver energy, making energetic applications attractive. Many complex physical phenomena are involved by the compression process, but it is possible to use simple analytical models to analyze the main critical points. We first determine the conditions to obtain fuel ignition. High thermonuclear gains are achieved if only a small fraction of the fuel called hot spot is used to trigger burn in the main fuel compressed on a low isentrope. A simple hot spot model will be described. The high pressure needed to drive the capsule compression are obtained by the ablation process. A simple Rocket model describe the main features of the implosion phase. Several parameters have to be controlled during the compression: irradiation symmetry, hydrodynamical stability and when the driver is a laser, the problems arising from interaction of the EM wave with the plasma. Two different schemes are examined: Indirect Drive which uses X-ray generated in a cavity to drive the implosion and the Fast Ignitor concept using a ultra intense laser beam to create the hot spot. At the end we present the Laser Megajoule (LMJ) project. LMJ is scaled to a thermonuclear gain of the order of ten. (authors)

  6. Review of compact, alternate concepts for magnetic confinement fusion

    International Nuclear Information System (INIS)

    Nickerson, S.B.; Shmayda, W.T.; Dinner, P.J.; Gierszewski, P.

    1984-06-01

    This report documents a study of compact alternate magnetic confinement fusion experiments and conceptual reactor designs. The purpose of this study is to identify those devices with a potential to burn tritium in the near future. The bulk of the report is made up of a review of the following compact alternates: compact toroids, high power density tokamaks, linear magnetic systems, compact mirrors, reversed field pinches and some miscellaneous concepts. Bumpy toruses and stellarators were initially reviewed but were not pursued since no compact variations were found. Several of the concepts show promise of either burning tritium or evolving into tritium burning devices by the early 1990's: RIGGATRON, Ignitor, OHTE, Frascati Tokamak upgrade, several driven (low or negative net power) mirror experiments and several Reversed Field Pinch experiments that may begin operation around 1990. Of the above only the Frascati Tokamak Upgrade has had funds allocated. Also identified in this report are groups who may have tritium burning experiments in the mid to late 1990's. There is a discussion of the differences between the reviewed devices and the mainline tokamak experiments. This discussion forms the basis of recommendations for R and D aimed at the compact alternates and the applicability of the present CFFTP program to the needs of the compact alternates. These recommendations will be presented in a subsequent report

  7. Compact magnetic confinement fusion: Spherical torus and compact torus

    Directory of Open Access Journals (Sweden)

    Zhe Gao

    2016-05-01

    Full Text Available The spherical torus (ST and compact torus (CT are two kinds of alternative magnetic confinement fusion concepts with compact geometry. The ST is actually a sub-category of tokamak with a low aspect ratio; while the CT is a toroidal magnetic configuration with a simply-connected geometry including spheromak and field reversed pinch. The ST and CT have potential advantages for ultimate fusion reactor; while at present they can also provide unique fusion science and technology contributions for mainstream fusion research. However, some critical scientific and technology issues should be extensively investigated.

  8. Inertial Confinement Fusion R and D and Nuclear Proliferation

    International Nuclear Information System (INIS)

    Goldston, Robert J.

    2011-01-01

    In a few months, or a few years, the National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory may achieve fusion gain using 192 powerful lasers to generate x-rays that will compress and heat a small target containing isotopes of hydrogen. This event would mark a major milestone after decades of research on inertial confinement fusion (ICF). It might also mark the beginning of an accelerated global effort to harness fusion energy based on this science and technology. Unlike magnetic confinement fusion (ITER, 2011), in which hot fusion fuel is confined continuously by strong magnetic fields, inertial confinement fusion involves repetitive fusion explosions, taking advantage of some aspects of the science learned from the design and testing of hydrogen bombs. The NIF was built primarily because of the information it would provide on weapons physics, helping the United States to steward its stockpile of nuclear weapons without further underground testing. The U.S. National Academies' National Research Council is now hosting a study to assess the prospects for energy from inertial confinement fusion. While this study has a classified sub-panel on target physics, it has not been charged with examining the potential nuclear proliferation risks associated with ICF R and D. We argue here that this question urgently requires direct and transparent examination, so that means to mitigate risks can be assessed, and the potential residual risks can be balanced against the potential benefits, now being assessed by the NRC. This concern is not new (Holdren, 1978), but its urgency is now higher than ever before.

  9. Continuous Electrode Inertial Electrostatic Confinement Fusion

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA recognizes within its roadmaps (specifically TA 3.1.6) that development of aneutronic fusion (such as p-11B) reactors with direct energy conversion (>80%)...

  10. System and method of operating toroidal magnetic confinement devices

    Science.gov (United States)

    Chance, M.S.; Jardin, S.C.; Stix, T.H.; Grimm, R.C.; Manickam, J.; Okabayashi, M.

    1984-08-30

    This invention pertains to methods and arrangements for attaining high beta values in plasma confinement devices. More specifically, this invention pertains to methods for accessing the second stability region of operation in toroidal magnetic confinement devices.

  11. Progress in toroidal confinement and fusion research

    International Nuclear Information System (INIS)

    Furth, H.P.

    1987-10-01

    During the past 30 years, the characteristic T/sub i/n tau/sub E/-value of toroidal-confinement experiments has advanced by more than seven orders of magnitude. Part of this advance has been due to an increase of gross machine parameters. Most of this advance has been due to an increase of gross machine parameters. Most of the advance is associated with improvements in the ''quality of plasma confinement.'' The combined evidence of spherator and tokamak research clarifies the role of magnetic-field geometry in determining confinement and points to the importance of shielding out plasma edge effects. A true physical understanding of anomalous transport remains to be achieved. 39 refs., 11 figs., 1 tab

  12. Potential well formation in electrostatic confinement devices. Technical progress report

    International Nuclear Information System (INIS)

    Cherrington, B.E.; Verdeyen, J.T.

    1975-01-01

    A large (2' diameter) spherical electrostatic confinement device has been constructed to test the feasibility of using inertial electrostatic forces to confine energetic plasmas capable of sustaining fusion reactions. Electron injection under high vacuum has produced negative wells that completely depress the potential in the center and approach the classical Langmuir virtual cathode. Electron injection into low pressure deuterium reproduces our previous results of an ion rich region within the negative well. Additional theoretical studies incorporating electrons with very narrow angular momentum (corresponding to trapped electrons in the center) has shown that an additional electron rich region (or ion rich if the polarities are reversed) can be produced within the ion rich region for presumably realistic ranges of parameters

  13. Inertial confinement fusion: present status and future potential

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1984-01-01

    Power from inertial confinement fusion holds much promise for society. This paper points out many of the benefits relative to combustion of hydrocarbon fuels and fission power. Potential problems are also identified and put in perspective. The progress toward achieving inertial fusion power is described and results of recent work at the Lawrence Livermore National Laboratory are presented. Key phenomenological uncertainties are described and experimental goals for the Nova laser system are given. Several ICF reactor designs are discussed

  14. Neutral particle kinetics in fusion devices

    International Nuclear Information System (INIS)

    Tendler, M.; Heifetz, D.

    1986-05-01

    The theory of neutral particle kinetics treats the transport of mass, momentum, and energy in a plasma due to neutral particles which themselves are unaffected by magnetic fields. This transport affects the global power and particle balances in fusion devices, as well as profile control and plasma confinement quality, particle and energy fluxes onto device components, performance of pumping systems, and the design of diagnostics and the interpretation of their measurements. This paper reviews the development of analytic, numerical, and Monte Carlo methods of solving the time-independent Boltzmann equation describing neutral kinetics. These models for neutral particle behavior typically use adaptations of techniques developed originally for computing neutron transport, due to the analogy between the two phenomena, where charge-exchange corresponds to scattering and ionization to absorption. Progress in the field depends on developing multidimensional analytic methods, and obtaining experimental data for the physical processes of wall reflection, the neutral/plasma interaction, and for processes in fusion devices which are directly related to neutral transport, such as H/sub α/ emission rates, plenum pressures, and charge-exchange emission spectra

  15. Neutral particle kinetics in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Tendler, M.; Heifetz, D.

    1986-05-01

    The theory of neutral particle kinetics treats the transport of mass, momentum, and energy in a plasma due to neutral particles which themselves are unaffected by magnetic fields. This transport affects the global power and particle balances in fusion devices, as well as profile control and plasma confinement quality, particle and energy fluxes onto device components, performance of pumping systems, and the design of diagnostics and the interpretation of their measurements. This paper reviews the development of analytic, numerical, and Monte Carlo methods of solving the time-independent Boltzmann equation describing neutral kinetics. These models for neutral particle behavior typically use adaptations of techniques developed originally for computing neutron transport, due to the analogy between the two phenomena, where charge-exchange corresponds to scattering and ionization to absorption. Progress in the field depends on developing multidimensional analytic methods, and obtaining experimental data for the physical processes of wall reflection, the neutral/plasma interaction, and for processes in fusion devices which are directly related to neutral transport, such as H/sub ..cap alpha../ emission rates, plenum pressures, and charge-exchange emission spectra.

  16. Laser thermonuclear fusion with force confinement of hot plasma

    International Nuclear Information System (INIS)

    Korobkin, V.V.; Romanovsky, M.Y.

    1994-01-01

    The possibility of the utilization of laser radiation for plasma heating up to thermonuclear temperatures with its simultaneous confinement by ponderomotive force is investigated. The plasma is located inside a powerful laser beam with a tubelike section or inside a cavity of duct section, formed by several intersecting beams focused by cylindrical lenses. The impact of various physical processes upon plasma confinement is studied and the criteria of plasma confinement and maintaining of plasma temperature are derived. Plasma and laser beam stability is considered. Estimates of laser radiation energy necessary for thermonuclear fusion are presented

  17. Fuel and helium confinement in fusion reactors

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Attenberger, S.E.

    1993-01-01

    An expanded macroscopic model for particle confinement is used to investigate both fuel and helium confinement in reactor plasmas. The authors illustrate the relative effects of external sources of fuel, divertor pumping, and wall and divertory recycle on core, edge and scrape-off layer densities by using separate particle confinement times for open-quote core close-quote fueling (deep pellet or beam penetration, τ c ), open-quote shallow close-quote fueling (shallow pellet penetration or neutral atoms that penetrate the scrape-off layer, τ s ) and fueling in the scrape-off layer (τ sol ). Because τ s is determined by the parallel flow velocity and characteristic distance to the divertor plate, it can be orders of magnitude lower than either τ c or τ sol . A dense scrape-off region, desirable for reduced divertor erosion, leads to a high fraction of the recycled neutrals being ionized in the scrape-off region and poor core fueling efficiency. The overall fueling efficiency can then be dramatically improved with either shallow or deep auxillary fueling. Helium recycle is nearly always coupled to the scrape-off region and does not lead to strong core accumulation unless the helium pumping efficiency is much less than the fuel pumping efficiency, or the plasma preferentially retains helium over hydrogenic ions. Differences between the results of this model, single-τ p macroscopic models, and 1-D and 2-D models are discussed in terms of assumptions and boundary conditions

  18. Trends and developments in magnetic confinement fusion reactor concepts

    International Nuclear Information System (INIS)

    Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

    1981-01-01

    An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts. Emphasis is placed on reactors that operate on the deuterium/tritium/lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts. The paper emphasizes recent developments of these concepts within the last two to three years

  19. 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)

  20. Magnetic and inertial confinement fusion - an overview

    International Nuclear Information System (INIS)

    Murtaza, G.

    1993-01-01

    This paper describes the status of ICF which output is given in terms of number of emitted neutrons and fusion energy as a function of the estimated input energy from the implosion driver. Results from the highest energy laser experiments are summarised. The theoretical targets and the proposed upgrades of NOVA, GEKKO and OMEGA are also shown. The promised ICF paradise will therefore be approached through a process of interpolation between two known extremes. The objectives of this experiment was to produce more than one MW of fusion power in a controlled way, to demonstrate the technology related to tritium usage tritium injection, its tracking, monitoring/recovery and to establish safe procedures for handling tritium in compliance with the regulatory requirements. (A.B)

  1. Heavy ion drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Keefe, D.

    1983-01-01

    The advantages of heavy ion beams as a way of delivering the needed energy and power to an inertial fusion target are surveyed. The existing broad technology base of particle accelerators provides an important foundation for designing, costing, and evaluating proposed systems. The sequence of steps needed for the verification of the heavy ion approach is described; recent research results are even more encouraging than had been assumed hitherto

  2. Heavy ion drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Keefe, D.

    1983-12-01

    The advantages of heavy ion beams as a way of delivering the needed energy and power to an inertial fusion target are surveyed. The existing broad technology base of particle accelerators provides an important foundation for designing, costing, and evaluating proposed systems. The sequence of steps needed for the verification of the heavy ion approach is described; recent research results are even more encouraging than had been assumed hitherto

  3. Using Quantum Confinement to Uniquely Identify Devices

    Science.gov (United States)

    Roberts, J.; Bagci, I. E.; Zawawi, M. A. M.; Sexton, J.; Hulbert, N.; Noori, Y. J.; Young, M. P.; Woodhead, C. S.; Missous, M.; Migliorato, M. A.; Roedig, U.; Young, R. J.

    2015-11-01

    Modern technology unintentionally provides resources that enable the trust of everyday interactions to be undermined. Some authentication schemes address this issue using devices that give a unique output in response to a challenge. These signatures are generated by hard-to-predict physical responses derived from structural characteristics, which lend themselves to two different architectures, known as unique objects (UNOs) and physically unclonable functions (PUFs). The classical design of UNOs and PUFs limits their size and, in some cases, their security. Here we show that quantum confinement lends itself to the provision of unique identities at the nanoscale, by using fluctuations in tunnelling measurements through quantum wells in resonant tunnelling diodes (RTDs). This provides an uncomplicated measurement of identity without conventional resource limitations whilst providing robust security. The confined energy levels are highly sensitive to the specific nanostructure within each RTD, resulting in a distinct tunnelling spectrum for every device, as they contain a unique and unpredictable structure that is presently impossible to clone. This new class of authentication device operates with minimal resources in simple electronic structures above room temperature.

  4. Inertial confinement fusion at the Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Lindman, E.; Baker, D.; Barnes, C.; Bauer, B.; Beck, J.B.

    1997-01-01

    The Los Alamos National Laboratory is contributing to the resolution of key issues in the US Inertial-Confinement-Fusion Program and plans to play a strong role in the experimental program at the National Ignition Facility when it is completed

  5. Inertial Confinement Fusion Annual Report 1997

    International Nuclear Information System (INIS)

    Correll, D

    1998-01-01

    The ICF Annual Report provides documentation of the achievements of the LLNL ICF Program during the fiscal year by the use of two formats: (1) an Overview that is a narrative summary of important results for the fiscal year and (2) a compilation of the articles that previously appeared in the ICF Quarterly Report that year. Both the Overview and Quarterly Report are also on the Web at http://lasers.llnl.gov/lasers/pubs/icfq.html. Beginning in Fiscal Year 1997, the fourth quarter issue of the ICF Quarterly was no longer printed as a separate document but rather included in the ICF Annual. This change provided a more efficient process of documenting our accomplishments with-out unnecessary duplication of printing. In addition we introduced a new document, the ICF Program Monthly Highlights. Starting with the September 1997 issue and each month following, the Monthly Highlights will provide a brief description of noteworthy activities of interest to our DOE sponsors and our stakeholders. The underlying theme for LLNL's ICF Program research continues to be defined within DOE's Defense Programs missions and goals. In support of these missions and goals, the ICF Program advances research and technology development in major interrelated areas that include fusion target theory and design, target fabrication, target experiments, and laser and optical science and technology. While in pursuit of its goal of demonstrating thermonuclear fusion ignition and energy gain in the laboratory, the ICF Program provides research and development opportunities in fundamental high-energy-density physics and supports the necessary research base for the possible long-term application of inertial fusion energy for civilian power production. ICF technologies continue to have spin-off applications for additional government and industrial use. In addition to these topics, the ICF Annual Report covers non-ICF funded, but related, laser research and development and associated applications. We also

  6. Dynamic behaviour of the high confinement mode of fusion plasmas

    International Nuclear Information System (INIS)

    Zohm, H.

    1995-05-01

    This paper describes the dynamic behaviour of the High Confinement mode (H-mode) of fusion plasmas, which is one of the most promising regimes of enhanced energy confinement in magnetic fusion research. The physics of the H-mode is not yet fully understood, and the detailed behaviour is complex. However, we establish a simple physics picture of the phenomenon. Although a first principles theory of the anomalous transport processes in a fusion plasma has not yet been given, we show that within the picture developed here, it is possible to describe the dynamic behaviour of the H-mode, namely the dynamics of the L-H transition and the occurrence of edge localized modes (ELMs). (orig.)

  7. Fueling of magnetic-confinement devices

    International Nuclear Information System (INIS)

    Milora, S.L.

    1981-01-01

    A general overview of the fueling of magnetic confinement devices is presented, with particular emphasis on recent experimental results. Various practical fueling mechanisms are considered, such as cold gas inlet (or plasma edge fueling), neutral beam injection, and injection of high speed cryogenic hydrogen pellets. The central role played by charged particle transport and recycle of plasma particles from material surfaces in contact with the plasma is discussed briefly. The various aspects of hydrogen pellet injection are treated in detail, including applications to the production of high purity startup plasmas for stellarators and other devices, refueling of tokamak plasmas, pellet ablation theory, and the technology and performance characteristics of low and high speed pellet injectors

  8. Data acquisition systems for fusion devices

    International Nuclear Information System (INIS)

    Van Haren, P.C.; Oomens, N.A.

    1993-01-01

    During the last two decades, computerized data acquisition systems (DASs) have been applied at magnetic confinement fusion devices. Present-day data acquisition is done by means of distributed computer systems and transient recorders in CAMAC systems. The development of DASs has been technology driven; the emphasis has been on the development of computer hardware and system software. For future DASs, challenging problems are to be solved: The DASs have to be better optimized with respect to the needs of the users. Existing bottlenecks, such as CAMAC-computer coupling or pulse file merging, need to be eliminated. Continuous or long-pulse operation will require the introduction of event abstraction in DAS design. 59 refs., 4 figs., 1 tab

  9. Pulse Star inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Blink, J.A.; Hogan, W.J.

    1985-01-01

    Pulse Star is a pool-type ICF reactor that emphasizes low cost and high safety levels. The reactor consists of a vacuum chamber (belljar) submerged in a compact liquid metal (Li 17 Pb 83 or lithium) pool which also contains the heat exchangers and liquid metal pumps. The shielding efficiency of the liquid metal pool is high enough to allow hands-on maintenance of (removed) pumps and heat exchangers. Liquid metal is allowed to spray through the 5.5 m radius belljar at a controlled rate, but is prohibited from the target region by a 4 m radius mesh first wall. The wetted first wall absorbs the fusion x-rays and debris while the spray region absorbs the fusion neutrons. The mesh allows vaporized liquid metal to blow through to the spray region where it can quickly cool and condense. Preliminary calculations show that a 2 m thick first wall could handle the mechanical (support, buckling, and x-ray-induced hoop) loads. Wetting and gas flow issues are in an initial investigation stage

  10. 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

  11. Radiation control in fusion plasmas by magnetic confinement

    International Nuclear Information System (INIS)

    Dachicourt, R.

    2012-10-01

    The present work addresses two important issues for the industrial use of fusion: plasma radiation control, as a part of the more general power handling issue, and high density tokamak operation. These two issues will be most critical in the demonstration reactor, called DEMO, intermediate step between ITER and a future commercial reactor. For DEMO, the need to radiate a large fraction of the power so as to limit the peak power load on the divertor will be a key constraint. High confinement will have to be combined with high radiated power fraction, and the required level of plasma purity. The main achievement of this thesis is to have shown experimental evidence of the existence of a stable plasma regime meeting the most critical requirements of a DEMO scenario: an electron density up to 40% above the Greenwald value, together with a fraction of radiated power close to 80%, with a good energy confinement and limited dilution. The plasma is additionally heated with ion cyclotron waves in a central electron heating scenario, featuring alpha particle heating. The original observations reported in this work bring highly valuable new pieces of information both to the physics of the tokamak edge layer and to the construction of an 'integrated operational scenario' required to successfully operate fusion devices. In the way for getting high density plasmas, the new observations involve the following topics. First, the formation of a poloidal asymmetry in the edge electron density profile, with a maximum density located close to toroidal pumped limiter. This asymmetry occurs inside the separatrix, with a constant plasma pressure on magnetic surfaces. Secondly, a correlative decrease of the electron temperature in the same edge region. Thirdly, the excellent coupling capabilities of the ICRH waves, up to a central line averaged electron density of 1.4 times the Greenwald density. Fourthly, a poloidally asymmetric edge radiation region, providing the dissipation of 80% of

  12. General principles of magnetic fusion confinement

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1980-01-01

    A few of the areas are described in which there is close interaction between atomic/molecular (A and M) and magnetic fusion physics. The comparisons between predictions of neoclassical transport theory and experiment depend on knowledge of ionization and recombination rate coefficients. Modeling of divertor/scrapeoff plasmas requires better low energy charge exchange cross sections for H + A/sup n+/ collisions. The range of validity of neutral beam trapping cross sections must be broadened, both to encompass the energies typical of present injection experiments and to deal with the problem of prompt trapping of highly excited beam atoms at high energy. Plasma fueling models present certain anomalies that could be resolved by calculation and measurement of low energy (<1 keV) charge exchange cross sections

  13. Fusion radioactivity confinement and application to postulated ITER accidents

    International Nuclear Information System (INIS)

    Piet, S.J.; Brereton, S.J.

    1991-01-01

    An assessment of the ITER radioactivity confinement shows reduction of potential accidental releases to the environment by two orders of magnitude. Important credits are the 1% volume/day confinement leakage rate, radioactivity decay for short-lived isotopes, resumption of detritiation/negative pressure within seven days of the accident, and wind meander during the slow confinement leakage. Achieving this two order of magnitude credit in practice requires appropriate design details, especially the leakage rate and detritiation/negative pressure equipment, and research to validate some key assumptions. The confinement maximizes dependence on passive safety features, thereby working toward using fusion's potential safety advantages. The confinement includes several confinement zones with varying human access allowances. Some confinement areas are normally isolated from the environment, the closed ventilation zone. Some areas have an inert cover gas to inhibit combustion. If future assessments of accidental overpressure show the need, we propose a filter/vent system. This report documents our work for the ITER Conceptual Design Activity (CDA). The report is consistent with the final CDA design reports and descriptions, except that our analysis includes a filter/vent. For gaseous or vapor tritium and for most activated aerosols, the reference release fraction is about 2%. For short-lived tungsten-rhenium aerosols, the reference release fraction is somewhat lower, as low as 0.5% for some accident scenarios. Even without resumption of detritiation/decontamination or negative pressure within seven days of the accident, the release fraction for stays below 4%

  14. Inertially confined fusion using heavy ion drivers

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.; Bangerter, R.O.; Bock, R.; Hogan, W.J.; Lindl, J.D.

    1991-10-01

    The various technical issues of HIF will be briefly reviewed in this paper. It will be seen that there are numerous areas in common in all the approaches to HIF. In the recent International Symposium on Heavy Ion Inertial Fusion, the attendees met in specialized workshop sessions to consider the needs for research in each area. Each of the workshop groups considered the key questions of this report: (1) Is this an appropriate time for international collaboration in HIF? (2) Which problems are most appropriate for such collaboration? (3) Can the sharing of target design information be set aside until other driver and systems issues are better resolved, by which time it might be supposed that there could be a relaxation of classification of target issues? (4) What form(s) of collaboration are most appropriate, e.g., bilateral or multilateral? (5) Can international collaboration be sensibly attempted without significant increases in funding for HIF? The authors of this report share the conviction that collaboration on a broad scale is mandatory for HIF to have the resources, both financial and personnel, to progress to a demonstration experiment. Ultimately it may be possible for a single driver with the energy, power, focusibility, and pulse shape to satisfy the needs of the international community for target physics research. Such a facility could service multiple experimental chambers with a variety of beam geometries and target concepts

  15. Future directions in inertial confinement fusion

    International Nuclear Information System (INIS)

    Bodner, S.E.

    1992-01-01

    The author discusses future directions for the ICF program. At this time there is still uncertainty on a number of key issues necessary to decide on what type of a National Ignition Facility should be constructed. Mechanisms are in place to answer these questions. The author offers his opinions of where the program is likely to proceed. Technology wise indications are that direct drive heating has the best chance of reaching ignition and high gain. This has the advantage of making all three major user programs happy, namely weapons physics, weapons effects, and electrical energy. The demand for and price of energy in the country will have a major impact on the way the program is developed. From the laser fusion side the most promising drivers at present seem to be KrF lasers, and a major concern for these systems is whether the peak to valley nonuniformities can be reduced to the 1 to 2% level when delivered to the target in order to avoid driving instabilities

  16. Deceleration phase of inertial confinement fusion implosions

    International Nuclear Information System (INIS)

    Betti, R.; Anderson, K.; Goncharov, V.N.; McCrory, R.L.; Meyerhofer, D.D.; Skupsky, S.; Town, R.P.J.

    2002-01-01

    A model for the deceleration phase and marginal ignition of imploding capsules is derived by solving a set of ordinary differential equations describing the hot-spot energy balance and the shell dynamics including the return shock propagation. It is found that heat flux leaving the hot spot goes back in the form of internal energy and PdV work of the material ablated off the inner-shell surface. Though the hot-spot temperature is reduced by the heat conduction losses, the hot-spot density increases due to the ablated material in such a way that the hot-spot pressure is approximately independent of heat conduction. For hot-spot temperatures exceeding approximately 7 keV, the ignition conditions are not affected by heat conduction losses that are recycled into the hot spot by ablation. Instead, the only significant internal energy loss is due to the hot-spot expansion tamped by the surrounding shell. The change of adiabat induced by the shock is also calculated for marginally igniting shells, and the relation between the in-flight and stagnation adiabats is in general agreement with the numerical fit of LASNEX simulations by Herrmann et al. [Nucl. Fusion 41, 99 (2001)] and the self-similar solution of Kemp et al. [Phys. Rev. Lett. 15, 3336 (2001)]. The minimum kinetic energy required for ignition is also calculated from the same model and shown to be in good agreement with the numerical fit of LASNEX simulations. It is also found that mass ablation leads to a significant reduction of the deceleration phase Rayleigh-Taylor instability growth rates and to the suppression of short wavelength modes

  17. Diagnostic measurements related to laser driven inertial confinement fusion

    International Nuclear Information System (INIS)

    Campbell, D.E.

    1979-01-01

    Scientists at the Lawrence Livermore Laboratory have been conducting laser driven inertial confinement fusion experiments for over five years. The first proof of the thermonuclear burn came at the Janus target irradiation facility in the spring of 1975. Since that time three succeedingly higher energy facilities have been constructed at Livermore, Cyclops, Argus and Shiva, where increased fusion efficiency has been demonstrated. A new facility, called Nova, is now in the construction phase and we are hopeful that scientific break even (energy released compared to incident laser energy on target) will be demonstrated here in early 1980's. Projected progress of the Livermore program is shown

  18. Nuclear fusion power supply device

    International Nuclear Information System (INIS)

    Nakagawa, Satoshi.

    1975-01-01

    Object: To use a hybrid power supply device, which comprises a thyristor power supply and a diode power supply, to decrease cost of a nuclear fusion power supply device. Structure: The device comprises a thyristor power supply connected through a closing unit and a diode power supply connected in parallel through a breaker, input of each power supply being applied with an output voltage of a flywheel AC generator. When a current transformer is excited, a disconnecting switch is turned on to close the diode power supply and a current of the current transformer is increased by an automatic voltage regulator to a set value within a predetermined period of time. Next, the current is cut off by a breaker, and when the breaker is in on position, the disconnecting switch is opened to turn on the closing unit. Thus, when a plasma electric current reaches a predetermined value, the breaker is turned on, and the current of the current transformer is controlled by the thyristor power supply. (Kamimura, M.)

  19. Transport of heavy ions in inertial confinement fusion

    International Nuclear Information System (INIS)

    Parvazian, A.; Shahbandari Gouchani, A.

    2007-01-01

    In this article we have investigated the interaction of heavy ions (U) with a target (Au). In inertial confinement fusion method Interaction between heavy ion beam and target was simulated, Numerical analysis of the Boltzmann Fokker Planck equation used in order to optimize the material of the target and Energy deposition of ion beam to electrons and ions of target and The thickness of the target were calculated.

  20. The Swedish fusion research programme on magnetic confinement 1978

    International Nuclear Information System (INIS)

    Lehnert, B.

    1978-02-01

    A review is given on the activities and plans for research on plasma physics and controlled fusion at the Royal Institute of Technology in Stockholm, with descriptions and motivations of the research lines being conducted. These activities include investigations on plasma-neutral gas interaction, development of special principles for plasma stabilization, magnetic confinement schemes being based mainly on poloidal fields, as well as the generation, heating, and diagnostics of plasmas being ''impermeable'' to neutral gas. (author)

  1. Inertial confinement fusion systems using heavy ion accelerators as drivers

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.; Godlove, T.F.; Keefe, D.

    1980-03-01

    Heavy ion accelerators are the most recent entrants in the effort to identify a practical driver for inertial confinement fusion. They are of interest because of the expected efficient coupling of ion kinetic energy to the thermal energy needed to implode the pellet and because of the good electrical efficiency of high intensity particle accelerators. The beam intensities required, while formidable, lie within the range that can be studied by extensions of the theories and the technology of modern high energy accelerators

  2. A comprehensive alpha-heating model for inertial confinement fusion

    Science.gov (United States)

    Christopherson, A. R.; Betti, R.; Bose, A.; Howard, J.; Woo, K. M.; Campbell, E. M.; Sanz, J.; Spears, B. K.

    2018-01-01

    A comprehensive model is developed to study alpha-heating in inertially confined plasmas. It describes the time evolution of a central low-density hot spot confined by a compressible shell, heated by fusion alphas, and cooled by radiation and thermal losses. The model includes the deceleration, stagnation, and burn phases of inertial confinement fusion implosions, and is valid for sub-ignited targets with ≤10 × amplification of the fusion yield from alpha-heating. The results of radiation-hydrodynamic simulations are used to derive realistic initial conditions and dimensionless parameters for the model. It is found that most of the alpha energy (˜90%) produced before bang time is deposited within the hot spot mass, while a small fraction (˜10%) drives mass ablation off the inner shell surface and its energy is recycled back into the hot spot. Of the bremsstrahlung radiation emission, ˜40% is deposited in the hot spot, ˜40% is recycled back in the hot spot by ablation off the shell, and ˜20% leaves the hot spot. We show here that the hot spot, shocked shell, and outer shell trajectories from this analytical model are in good agreement with simulations. A detailed discussion of the effect of alpha-heating on the hydrodynamics is also presented.

  3. A remote monitoring system of environmental electromagnetic field in magnetic confinement fusion test facilities

    International Nuclear Information System (INIS)

    Tanaka, Masahiro; Uda, Tatsuhiko; Takami, Shigeyuki; Wang, Jianqing; Fujiwara, Osamu

    2010-01-01

    A remote, continuous environmental electromagnetic field monitoring system for use in magnetic confinement fusion test facilities is developed. Using this system, both the static magnetic field and the high frequency electromagnetic field could be measured. The required frequency range of the measurement system is from 25 to 100 MHz for the ICRF (Ion Cyclotron Range of Frequencies) heating system. The outputs from the measurement instruments are measured simultaneously by custom-built software using a laptop-type personal computer connected to a local area network. In this way, the electromagnetic field strength could be monitored from a control room located about 200 m from the fusion device building. Examples of measurement data from the vicinity of a high-frequency generator and amplifier and the leakage static magnetic field from a fusion test device are presented. (author)

  4. Analysis of Confinement Strategies for a Tokamak Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-15

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

  5. Analysis of Confinement Strategies for a Tokamak Fusion Reactor

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. Fusion an introduction to the physics and technology of magnetic confinement fusion

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    This second edition of a popular textbook is thoroughly revised with around 25% new and updated content.It provides an introduction to both plasma physics and fusion technology at a level that can be understood by advanced undergraduates and graduate students in the physical sciences and related engineering disciplines.As such, the contents cover various plasma confinement concepts, the support technologies needed to confine the plasma, and the designs of ITER as well as future fusion reactors.With end of chapter problems for use in courses.

  7. Glass solidification material confinement test device

    International Nuclear Information System (INIS)

    Namiki, Shigekazu.

    1997-01-01

    In a device for confining glass solidification materials, a pipeline connecting a detection vessel and a detector is formed to have a double walled structure, and air blowing holes are formed on the wall of the inner pipe, and an air supply mechanism is connected to inner and outer pipes for supplying blowing air thereby preventing deposition on the inner pipe wall. The air blowing holes are formed by constituting the pipe by using a porous sintered material and porous portions thereof are defined as the air blowing holes, or holes are formed on the pipe wall made of a metal by machining. A blowing boundary layer is formed by blowing the supplied air along the pipe wall of the inner pipe, by which deposition of the sucked materials to the inner wall of the inner pipe is prevented, and all of the materials sucked from the detection vessel are collected to the detector. In addition, an air exit pipe is formed into a double walled structure so as to be supplied blowing air from the air supply mechanism thereby enabling to prevent deposition of sucked materials more reliably. (N.H.)

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

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

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

  9. Institute for Fusion Research and Large Helical Device program

    International Nuclear Information System (INIS)

    Iiyoshi, Atsuo

    1989-01-01

    In the research on nuclear fusion, the final objective is to materialize nuclear fusion reactors, and for the purpose, it is necessary to cause nuclear combustion by making the plasma of higher than 100 million deg and confine it for a certain time. So far in various universities, the researches on diversified fusion processes have been advanced, but in February, 1986, the Science Council issued the report 'Nuclear fusion research in universities hereafter'. As the next large scale device, an external conductor system helical device was decided, and it is desirable to found the organization for joint utilization by national universities to promote the project. The researches on the other processes are continued by utilizing the existing facilitie. The reason of selecting a helical device is the data base of the researches carried out so far can be utilized sufficiently, it is sufficiently novel even after 10 years from now, and many researchers can be collected. The place of the research is Toki City, Gifu Prefecture, where the Institute of Plasma Physics, Nagoya University, is to be moved. The basic concept of the superconducting helical device project, the trend of nuclear fusion development in the world, the physical research using a helical system and so on are reported. (Kako, I.)

  10. Inertial confinement fusion systems using heavy ion accelerators as drivers

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.; Godlove, T.F.; Keefe, D.

    1980-01-01

    Heavy ion accelerators are the most recent entrants in the effort to identify a practical driver for inertial confinement fusion. They are of interest because of the expected efficient coupling of ion kinetic energy to the thermal energy needed to implode the pellet and because of the good electrical efficiency of high intensity particle accelerators. The beam intensities required, while formidable, lie within the range that can be studied by extensions of the theories and the technology of modern high energy accelerators. (orig.) [de

  11. Open-ended magnetic confinement systems for fusion

    International Nuclear Information System (INIS)

    Post, R.F.; Ryutov, D.D.

    1995-05-01

    Magnetic confinement systems that use externally generated magnetic fields can be divided topologically into two classes: ''closed'' and 'open''. The tokamak, the stellarator, and the reversed-field-pinch approaches are representatives of the first category, while mirror-based systems and their variants are of the second category. While the recent thrust of magnetic fusion research, with its emphasis on the tokamak, has been concentrated on closed geometry, there are significant reasons for the continued pursuit of research into open-ended systems. The paper discusses these reasons, reviews the history and the present status of open-ended systems, and suggests some future directions for the research

  12. Repetitive pulsed power technology for inertial-confinement fusion

    International Nuclear Information System (INIS)

    Prestwich, K.R.; Buttram, M.T.

    1983-01-01

    The pulsed power requirements for inertial-confinement fusion reactors are defined for ion-beam and laser drivers. Several megajoule beams with 100's of terrawatt peak powers must be delivered to the reactor chamber 1 to 10 times per second. Ion-beam drivers are relatively efficient requiring less energy storage in the pulsed-power system but more time compression in the power flow chain than gas lasers. These high peak powers imply very large numbers of components for conventional pulse-power systems. A new design that significantly reduces the number of components is presented

  13. Progress of research and development of nuclear fusion and development of large nuclear fusion device technology

    International Nuclear Information System (INIS)

    1994-01-01

    In the last several years, the results of tokamak experiments were conspicuous, and the progress of plasma confinement performance, transport mechanism, divertors and impurities, helium transport and exhaust, electric current drive, magnetic field ripple effect and high speed particle transport and DT experiment are reported. The other confinement methods than tokamak, the related theories and reactor technology are described. The conceptual design of ITER was carried out by the cooperation of Japan, USA, EC and the former USSR. The projects of developing nuclear fusion in various countries, the design and the required research and development of ITER, the reconstruction and the required research and development of JT-60, JET and TFTR, the design and the required research and development of large helical device, the state of research and development of laser nuclear fusion and inversion magnetic field pinch nuclear fusion, the activities and roles of industrial circles in large nuclear fusion device technology, and the long term perspective of the technical development of nuclear fusion are described. (K.I.)

  14. Heat transfer in inertial confinement fusion reactor systems

    International Nuclear Information System (INIS)

    Hovingh, J.

    1979-01-01

    The transfer of energy produced by the interaction of the intense pulses of short-ranged fusion microexplosion products with materials is one of the most difficult problems in inertially-confined fusion (ICF) reactor design. The short time and deposition distance for the energy results in local peak power densities on the order of 10 18 watts/m 3 . High local power densities may cause change of state or spall in the reactor materials. This will limit the structure lifetimes for ICF reactors of economic physical sizes, increasing operating costs including structure replacement and radioactive waste management. Four basic first wall protection methods have evolved: a dry-wall, a wet-wall, a magnetically shielded wall, and a fluid wall. These approaches are distinguished by the way the reactor wall interfaces with fusion debris as well as the way the ambient cavity conditions modify the fusion energy forms and spectra at the first wall. Each of these approaches requires different heat transfer considerations

  15. A Parallel Boltzmann Simulation for Multi-grid Inertial Electrostatic Confinement Fusion

    Data.gov (United States)

    National Aeronautics and Space Administration — Inertial electrostatic confinement (IEC) is a means of confining a non-neutral, non-Maxwellian plasma with an electric field, with the goal of creating fusion...

  16. Fourth annual progress report on special-purpose materials for magnetically confined fusion reactors

    International Nuclear Information System (INIS)

    1982-08-01

    The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. The Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits

  17. The VISTA spacecraft: Advantages of ICF [Inertial Confinement Fusion] for interplanetary fusion propulsion applications

    International Nuclear Information System (INIS)

    Orth, C.D.; Klein, G.; Sercel, J.; Hoffman, N.; Murray, K.; Chang-Diaz, F.

    1987-01-01

    Inertial Confinement Fusion (ICF) is an attractive engine power source for interplanetary manned spacecraft, especially for near-term missions requiring minimum flight duration, because ICF has inherent high power-to-mass ratios and high specific impulses. We have developed a new vehicle concept called VISTA that uses ICF and is capable of round-trip manned missions to Mars in 100 days using A.D. 2020 technology. We describe VISTA's engine operation, discuss associated plasma issues, and describe the advantages of DT fuel for near-term applications. Although ICF is potentially superior to non-fusion technologies for near-term interplanetary transport, the performance capabilities of VISTA cannot be meaningfully compared with those of magnetic-fusion systems because of the lack of a comparable study of the magnetic-fusion systems. We urge that such a study be conducted

  18. Progress in the pulsed power Inertial Confinement Fusion program

    International Nuclear Information System (INIS)

    Quintenz, J.P.; Matzen, M.K.; Mehlhorn, T.A.

    1996-01-01

    Pulsed power accelerators are being used in Inertial Confinement Fusion (ICF) research. In order to achieve our goal of a fusion yield in the range of 200 - 1000 MJ from radiation-driven fusion capsules, it is generally believed that ∼10 MJ of driver energy must be deposited within the ICF target in order to deposit ∼1 MJ of radiation energy in the fusion capsule. Pulsed power represents an efficient technology for producing both these energies and these radiation environments in the required short pulses (few tens of ns). Two possible approaches are being developed to utilize pulsed power accelerators in this effort: intense beams of light ions and z- pinches. This paper describes recent progress in both approaches. Over the past several years, experiments have successfully answered many questions critical to ion target design. Increasing the ion beam power and intensity are our next objectives. Last year, the Particle Beam Fusion Accelerator H (PBFA II) was modified to generate ion beams in a geometry that will be required for high yield applications. This 2048 modification has resulted in the production of the highest power ion beam to be accelerated from an extraction ion diode. We are also evaluating fast magnetically-driven implosions (z-pinches) as platforms for ICF ablator physics and EOS experiments. Z-pinch implosions driven by the 20 TW Saturn accelerator have efficiently produced high x- ray power (> 75 TW) and energy (> 400 kJ). Containing these x-ray sources within a hohlraum produces a unique large volume (> 6000 mm 3 ), long lived (>20 ns) radiation environment. In addition to studying fundamental ICF capsule physics, there are several concepts for driving ICF capsules with these x-ray sources. Progress in increasing the x-ray power on the Saturn accelerator and promise of further increases on the higher power PBFA II accelerator will be described

  19. Ignition and Inertial Confinement Fusion at The National Ignition Facility

    International Nuclear Information System (INIS)

    Moses, E.

    2009-01-01

    The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and for studying high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). The NIF is now conducting experiments to commission the laser drive, the hohlraum and the capsule and to develop the infrastructure needed to begin the first ignition experiments in FY 2010. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. NIF will achieve this by concentrating the energy from the 192 beams into a mm 3 -sized target and igniting a deuterium-tritium mix, liberating more energy than is required to initiate the fusion reaction. NIF's ignition program is a national effort managed via the National Ignition Campaign (NIC). The NIC has two major goals: execution of DT ignition experiments starting in FY2010 with the goal of demonstrating ignition and a reliable, repeatable ignition platform by the conclusion of the NIC at the end of FY2012. The NIC will also develop the infrastructure and the processes required to operate NIF as a national user facility. The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on laser fusion as a viable energy option. A laser fusion-based energy concept that builds on NIF, known as LIFE (Laser Inertial Fusion Energy), is currently under development. LIFE is inherently safe and can provide a global carbon-free energy generation solution in the 21st century. This paper describes recent progress on NIF, NIC, and the LIFE concept.

  20. 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

  1. Plasma fluctuations and confinement of fusion reaction products

    International Nuclear Information System (INIS)

    Coppi, B.; Pegoraro, F.

    1981-01-01

    The interaction between the fluctuations that can be excited in a magnetically confined plasma and the high-energy-particle population produced by fusion reactions is analyzed in view of its relevance to the process of thermonuclear ignition. The spectrum of the perturbations that, in the absence of fusion reaction products, would be described by the incompressible ideal magnetohydrodynamic approximation is studied considering finite value of the plasma pressure relative ot the magnetic pressure. The combined effects of the magnetic field curvature and shear are taken into account and the relevant spectrum is shown to consist of a continuous portion, that could be identified as a mixture of shear-Alfven and interchange oscillations, and a discrete unstable part corresponding to the so-called ballooning modes. The rate of diffusion of the fusion reaction products induced by oscillations in the continuous part of the spectrum, as estimated from the appropriate quasi-linear theory, is found to be significantly smaller than could be expected if normal modes (i.e., nonconvective solutions) were excited. However, a relatively wide intermediate region is identified where opalescent fluctuations, capable of achieving significant amplitudes and corresponding to a quasi-discrete spectrum, can be excited

  2. 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

  3. Electromagnetic pumping of liquid lithium in inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Baker, R.S.; Blink, J.A.; Tessier, M.J.

    1983-01-01

    The basic operating principles and geometries of ten electromagnetic pumps are described. Two candidate pumps, the annular-linear-induction pump and the helical-rotor electromagnetic pump, are compared for possible use in a full-scale liquid-lithium inertial confinement fusion reactor. A parametric design study completed for the helical-rotor pump is shown to be valid when applied to an experimental sodium pump. Based upon the preliminary HYLIFE requirements for a lithium flow rate per pump of 8.08 m 3 /s at a head of 82.5 kPa, a complete set of 70 variables are specified for a helical-rotor pump with either a normally conducting or a superconducting winding. The two alternative designs are expected to perform with efficiencies of 50 and 60%, respectively

  4. Tertiary proton diagnostics in future inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Cremer, S.; Verdon, C.P.; Petrasso, R.D.

    1998-01-01

    Recently, it was proposed to use energetic (up to 31 MeV) tertiary protons produced during the final stage of inertial confinement fusion implosions to measure the fuel areal density of compressed deuterium endash tritium (DT). The method is based on seeding the fuel with 3 He. The reaction of 3 He ions with the energetic knock-on deuterons, produced via the elastic scattering of 14.1 MeV neutrons, is a source of very energetic protons capable of escaping from very large areal density targets. This work presents results of detailed time-dependent Monte Carlo simulations of the nuclear processes involved in producing and transporting these protons through imploding targets proposed for direct-drive experiments on OMEGA [D. K. Bradley et al., Phys. Plasmas 5, 1870 (1998)] and the National Ignition Facility [S. W. Haan et al., Phys. Plasmas 2, 2480 (1995)]. copyright 1998 American Institute of Physics

  5. 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

  6. Passive Spectroscopic Diagnostics for Magnetically-confined Fusion Plasmas

    International Nuclear Information System (INIS)

    Stratton, B.C.; Bitter, M.; Hill, K.W.; Hillis, D.L.; Hogan, J.T.

    2007-01-01

    Spectroscopy of radiation emitted by impurities and hydrogen isotopes plays an important role in the study of magnetically-confined fusion plasmas, both in determining the effects of impurities on plasma behavior and in measurements of plasma parameters such as electron and ion temperatures and densities, particle transport, and particle influx rates. This paper reviews spectroscopic diagnostics of plasma radiation that are excited by collisional processes in the plasma, which are termed 'passive' spectroscopic diagnostics to distinguish them from 'active' spectroscopic diagnostics involving injected particle and laser beams. A brief overview of the ionization balance in hot plasmas and the relevant line and continuum radiation excitation mechanisms is given. Instrumentation in the soft X-ray, vacuum ultraviolet, ultraviolet, visible, and near-infrared regions of the spectrum is described and examples of measurements are given. Paths for further development of these measurements and issues for their implementation in a burning plasma environment are discussed.

  7. Passive Spectroscopic Diagnostics for Magnetically-confined Fusion Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Stratton, B. C.; Biter, M.; Hill, K. W.; Hillis, D. L.; Hogan, J. T.

    2007-07-18

    Spectroscopy of radiation emitted by impurities and hydrogen isotopes plays an important role in the study of magnetically-confined fusion plasmas, both in determining the effects of impurities on plasma behavior and in measurements of plasma parameters such as electron and ion temperatures and densities, particle transport, and particle influx rates. This paper reviews spectroscopic diagnostics of plasma radiation that are excited by collisional processes in the plasma, which are termed 'passive' spectroscopic diagnostics to distinguish them from 'active' spectroscopic diagnostics involving injected particle and laser beams. A brief overview of the ionization balance in hot plasmas and the relevant line and continuum radiation excitation mechanisms is given. Instrumentation in the soft X-ray, vacuum ultraviolet, ultraviolet, visible, and near-infrared regions of the spectrum is described and examples of measurements are given. Paths for further development of these measurements and issues for their implementation in a burning plasma environment are discussed.

  8. Free Electron Laser as Energy Driver for Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Saldin, E.L.; Shnejdmiller, E.A.; Ul'yanov, Yu.N.; Sarantsev, V.P.; Yurkov, M.V.

    1994-01-01

    A FEL based energy driver for Inertial Confinement Fusion (ICF) is proposed. The key element of the scheme is free electron laser system. Novel technical solutions reveal a possibility to construct the FEL system operating at radiation wavelength λ = 0.5 μm and providing flash energy E = 1 MJ and brightness 4 x 10 22 W cm -2 sr -1 within steering pulse duration 0.1-2 ns. Total energy efficiency of the proposed ICF energy driver is about of 11% and repetition rate is 40 Hz. Dimensions of such an ICF driver are comparable with those of heavy-ion ICF driver, while the problem of technical realization seems to be more realistic. It is shown that the FEL based ICF energy driver may be constructed at the present level of accelerator technique R and D. 27 refs., 10 figs., 3 tabs

  9. Application of inertial confinement fusion to weapon technology

    International Nuclear Information System (INIS)

    Toepfer, A.J.; Posey, L.D.

    1978-12-01

    This report reviews aspects of the military applications of the inertial confinement fusion (ICF) program at Sandia Laboratories. These applications exist in the areas of: (1) weapon physics research, and (2) weapon effects simulation. In the area of weapon physics research, ICF source technology can be used to study: (1) equations-of-state at high energy densities, (2) implosion dynamics, and (3) laboratory simulation of exoatmospheric burst phenomena. In the area of weapon effects simulation, ICF technology and facilities have direct near, intermediate, and far term applications. In the near term, short pulse x-ray simulation capabilities exist for electronic component effects testing. In the intermediate term, capabilities can be developed for high energy neutron exposures and bremsstrahlung x-ray exposures of components. In the far term, system level exposures of full reentry vehicles will be possible if sufficiently high pellet gains are achieved

  10. Computer modeling and simulation in inertial confinement fusion

    International Nuclear Information System (INIS)

    McCrory, R.L.; Verdon, C.P.

    1989-03-01

    The complex hydrodynamic and transport processes associated with the implosion of an inertial confinement fusion (ICF) pellet place considerable demands on numerical simulation programs. Processes associated with implosion can usually be described using relatively simple models, but their complex interplay requires that programs model most of the relevant physical phenomena accurately. Most hydrodynamic codes used in ICF incorporate a one-fluid, two-temperature model. Electrons and ions are assumed to flow as one fluid (no charge separation). Due to the relatively weak coupling between the ions and electrons, each species is treated separately in terms of its temperature. In this paper we describe some of the major components associated with an ICF hydrodynamics simulation code. To serve as an example we draw heavily on a two-dimensional Lagrangian hydrodynamic code (ORCHID) written at the University of Rochester's Laboratory for Laser Energetics. 46 refs., 19 figs., 1 tab

  11. 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

  12. Control of a laser inertial confinement fusion-fission power plant

    Science.gov (United States)

    Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

    2015-10-27

    A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

  13. Geodesic least squares regression for scaling studies in magnetic confinement fusion

    International Nuclear Information System (INIS)

    Verdoolaege, Geert

    2015-01-01

    In regression analyses for deriving scaling laws that occur in various scientific disciplines, usually standard regression methods have been applied, of which ordinary least squares (OLS) is the most popular. However, concerns have been raised with respect to several assumptions underlying OLS in its application to scaling laws. We here discuss a new regression method that is robust in the presence of significant uncertainty on both the data and the regression model. The method, which we call geodesic least squares regression (GLS), is based on minimization of the Rao geodesic distance on a probabilistic manifold. We demonstrate the superiority of the method using synthetic data and we present an application to the scaling law for the power threshold for the transition to the high confinement regime in magnetic confinement fusion devices

  14. Innovative approaches to inertial confinement fusion reactors: Final report

    International Nuclear Information System (INIS)

    Bourque, R.F.; Schultz, K.R.

    1986-11-01

    Three areas of innovative approaches to inertial confinement fusion (ICF) reactor design are given. First, issues pertaining to the Cascade reactor concept are discussed. Then, several innovative concepts are presented which attempt to directly recover the blast energy from a fusion target. Finally, the Turbostar concept for direct recovery of that energy is evaluated. The Cascade issues discussed are combustion of the carbon granules in the event of air ingress, the use of alternate granule materials, and the effect of changes in carbon flow on details of the heat exchanger. Carbon combustion turns out to be a minor problem. Four ICF innovative concepts were considered: a turbine with ablating surfaces, a liquid piston system, a wave generator, and a resonating pump. In the final analysis, none show any real promise. The Turbostar concept of direct recovery is a very interesting idea and appeared technically viable. However, it shows no efficiency gain or any decrease in capital cost compared to reactors with conventional thermal conversion systems. Attempts to improve it by placing a close-in lithium sphere around the target to increase gas generation increased efficiency only slightly. It is concluded that these direct conversion techniques require thermalization of the x-ray and debris energy, and are Carnot limited. They therefore offer no advantage over existing and proposed methods of thermal energy conversion or direct electrical conversion

  15. Addressing Common Technical challenges in Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Haynes, Donald A.

    2016-01-01

    The implosion phase for Inertial Confinement Fusion (ICF) occurs from initiation of the drive until just before stagnation. Evolution of the shell and fusion fuel during the implosion phase is affected by the initial conditions of the target, the drive history. Poor performing implosions are a result of the behavior that occurs during the implosion phase such as low mode asymmetries, mixing of the ablator into the fuel, and the hydrodynamic evolution of initial target features and defects such as the shell mounting hardware. The ultimate results of these effects can only be measured at stagnation. However, studying the implosion phase can be effective for understanding and mitigating these effects and for of ultimately improving the performance of ICF implosions. As the ICF program moves towards the 2020 milestone to ''determine the efficacy of ignition'', it will be important to understand the physics that occurs during the implosion phase. This will require both focused and integrated experiments. Focused experiments will provide the understanding and the evidence needed to support any determination concerning the efficacy of ignition.

  16. Addressing Common Technical challenges in Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Haynes, Donald A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-22

    The implosion phase for Inertial Confinement Fusion (ICF) occurs from initiation of the drive until just before stagnation. Evolution of the shell and fusion fuel during the implosion phase is affected by the initial conditions of the target, the drive history. Poor performing implosions are a result of the behavior that occurs during the implosion phase such as low mode asymmetries, mixing of the ablator into the fuel, and the hydrodynamic evolution of initial target features and defects such as the shell mounting hardware. The ultimate results of these effects can only be measured at stagnation. However, studying the implosion phase can be effective for understanding and mitigating these effects and for of ultimately improving the performance of ICF implosions. As the ICF program moves towards the 2020 milestone to “determine the efficacy of ignition”, it will be important to understand the physics that occurs during the implosion phase. This will require both focused and integrated experiments. Focused experiments will provide the understanding and the evidence needed to support any determination concerning the efficacy of ignition.

  17. Technology requirements for commercial applications of inertial confinement fusion

    International Nuclear Information System (INIS)

    Frank, T.G.; Rossi, C.E.

    1981-01-01

    Current inertial confinement fusion (ICF) research is directed primarily at physics experiments intended to provide confidence in the scientific feasibility of the basic concept. In conjunction with these experiments, a variety of laser and particle beam drivers having potential for eventual use in fusion power plants is being developed. Expectations are that the scientific feasibility of ICF will be demonstrated in the latter part of the 1980s. At that time, the emphasis of the program will shift to engineering, economic, environmental, and licensing issues with the necessary technology development effort continuing into the early part of the next century. This paper discusses the technology requirements for the successive phases of engineering development leading to commercial application of ICF. The engineering areas requiring significant effort for ICF application include high average power driver development; pulsed high-energy power supply development; reactor cavity and heat transport system design; tritium extraction and control; commercial pellet development; pellet injection, tracking, and targeting systems design; materials radiation, fatigue, and corrosion behavior; and reactor plant systems integration and demonstration

  18. Magnetic stochasticity in magnetically confined fusion plasmas chaos of field lines and charged particle dynamics

    CERN Document Server

    Abdullaev, Sadrilla

    2014-01-01

    This is the first book to systematically consider the modern aspects of chaotic dynamics of magnetic field lines and charged particles in magnetically confined fusion plasmas.  The analytical models describing the generic features of equilibrium magnetic fields and  magnetic perturbations in modern fusion devices are presented. It describes mathematical and physical aspects of onset of chaos, generic properties of the structure of stochastic magnetic fields, transport of charged particles in tokamaks induced by magnetic perturbations, new aspects of particle turbulent transport, etc. The presentation is based on the classical and new unique mathematical tools of Hamiltonian dynamics, like the action--angle formalism, classical perturbation theory, canonical transformations of variables, symplectic mappings, the Poincaré-Melnikov integrals. They are extensively used for analytical studies as well as for numerical simulations of magnetic field lines, particle dynamics, their spatial structures and  statisti...

  19. Performance test of personal RF monitor for area monitoring at magnetic confinement fusion facility

    International Nuclear Information System (INIS)

    Tanaka, M.; Uda, T.; Wang, J.; Fujiwara, O.

    2012-01-01

    For safety management at a magnetic confinement fusion-test facility, protection from not only ionising radiation, but also non-ionising radiation such as the leakage of static magnetic and electromagnetic fields is an important issue. Accordingly, the use of a commercially available personal RF monitor for multipoint area monitoring is proposed. In this study, the performance of both fast- and slow-type personal RF monitors was investigated by using a transverse electromagnetic cell system. The range of target frequencies was between 10 and 300 MHz, corresponding to the ion cyclotron range of frequency in a fusion device. The personal RF monitor was found to have good linearity, frequency dependence and isotropic response. However, the time constant for the electric field sensor of the slow-type monitor was much longer than that for the fast-type monitor. Considering the time-varying field at the facility, it is found that the fast-type monitor is suitable for multipoint monitoring at magnetic confinement fusion test facilities. (authors)

  20. Improvement of proton source based on cylindrical inertial electrostatic confinement fusion with ion source

    International Nuclear Information System (INIS)

    Yamauchi, Kunihito; Ohura, Sonoe; Tashiro, Atsushi; Watanabe, Masato; Okino, Akitoshi; Kohno, Toshiyuki; Hotta, Eiki; Yuura, Morimasa

    2005-01-01

    Inertial Electrostatic Confinement Fusion (IECF) device is a compact fusion proton/neutron source with an extremely simple configuration, high controllability, and hence high safety. Therefore, it has been studied for practical use as a portable neutron/proton source for various applications such as landmine detection and medical positron emission tomography. However, some problems remain for the practical use, and the most critical one is the insufficiency of absolute neutron/proton yields. In this study, a new IECF device was designed and tested to obtain high neutron/proton yields. The key features of the new device are the cylindrical electrode configuration in consideration of better electrostatic confinement of ions and extraction of protons, and an integrated ion source that consists of sixteen ferrite magnets and biasing the grid anode. To investigate the performance characteristics of the device and the effect of the ion source, three kinds of experimental setup were used for comparison. At first, the device was operated with the basic setup. Then a cusp magnetic field was applied by using ferrite magnets, and the grid anode was negatively biased. As a result, it was confirmed that the ion source works effectively. At the same voltage and current, the obtained neutron production rate was about one order of magnitude higher than that of the conventional spherical IECF device. The maximum neutron production rate of 6.8x10 9 n/s was obtained at a pulsed discharge of -70 kV and 10 A with an anode bias voltage of -1.0 kV. (author)

  1. Special-purpose materials for magnetically confined fusion reactors. Third annual progress report

    International Nuclear Information System (INIS)

    1981-11-01

    The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. It is recognized that there will be numerous materials problems that will arise during the design and construction of large magnetic-fusion energy devices such as the Engineering Test Facility (ETF) and Demonstration Reactor (DEMO). Most of these problems will be specific to a particular design or project and are the responsibility of the project, not the Materials and Radiation Effects Branch. Consequently, the Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits

  2. 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

  3. Ion accelerators as drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Faltens, A.; Keefe, D.; Rosenblum, S.S.

    1980-11-01

    During the past few years the possibility of using intense ion beams to ignite a pellet of fusion fuel has looked increasingly promising. Ion beams ranging in mass from protons up to uranium have been investigated and several machines have been built at different laboratories to investigate the required technology. Light ion drivers are based on the use of high current, high voltage diodes arranged around a central target. These devices have the necessary power and energy to initiate fusion burn but suffer from the inability to transport stably the necessary huge beam currents over long distances to a small target. Heavy ion drivers are based either on the radio-frequency linac or the induction linac. Because heavy ions have a much shorter range than light ions of the same energy, one is able to raise the beam voltage by a factor of one-thousand and lower the current correspondingly. The expected parameters for a fusion driver will be delineated and the present state of development of the technology for the candidate ion beam drivers will be described in light of these desiderata

  4. Magnetic systems for fusion devices

    International Nuclear Information System (INIS)

    Henning, C.D.

    1985-02-01

    Mirror experiments have led the way in applying superconductivity to fusion research because of unique requirements for high and steady magnetic fields. The first significant applications were Baseball II at LLNL and IMP at ORNL. More recently, the MFTF-B yin-yang coil was successfully tested and the entire tandem configuration is nearing completion. Tokamak magnets have also enjoyed recent success with the large coil project tests at ORNL, preceded by single coil tests in Japan and Germany. In the USSR, the T-7 Tokamak has been operational for many years and the T-15 Tokamak is under construction, with the TF coils nearing completion. Also the Tore Supra is being built in France

  5. Numerical modelling of electromagnetic loads on fusion device structures

    International Nuclear Information System (INIS)

    Bettini, Paolo; Palumbo, Maurizio Furno; Specogna, Ruben

    2014-01-01

    In magnetic confinement fusion devices, during abnormal operations (disruptions) the plasma begins to move rapidly towards the vessel wall in a vertical displacement event (VDE), producing plasma current asymmetries, vessel eddy currents and open field line halo currents, each of which can exert potentially damaging forces upon the vessel and in-vessel components. This paper presents a methodology to estimate electromagnetic loads, on three-dimensional conductive structures surrounding the plasma, which arise from the interaction of halo-currents associated to VDEs with a magnetic field of the order of some Tesla needed for plasma confinement. Lorentz forces, calculated by complementary formulations, are used as constraining loads in a linear static structural analysis carried out on a detailed model of the mechanical structures of a representative machine

  6. Numerical modelling of electromagnetic loads on fusion device structures

    Science.gov (United States)

    Bettini, Paolo; Furno Palumbo, Maurizio; Specogna, Ruben

    2014-03-01

    In magnetic confinement fusion devices, during abnormal operations (disruptions) the plasma begins to move rapidly towards the vessel wall in a vertical displacement event (VDE), producing plasma current asymmetries, vessel eddy currents and open field line halo currents, each of which can exert potentially damaging forces upon the vessel and in-vessel components. This paper presents a methodology to estimate electromagnetic loads, on three-dimensional conductive structures surrounding the plasma, which arise from the interaction of halo-currents associated to VDEs with a magnetic field of the order of some Tesla needed for plasma confinement. Lorentz forces, calculated by complementary formulations, are used as constraining loads in a linear static structural analysis carried out on a detailed model of the mechanical structures of a representative machine.

  7. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

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

  8. Plasma surface interactions in controlled fusion devices

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L.

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak

  9. Plasma surface interactions in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L. [and others

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak.

  10. Inertial electrostatic confinement fusion neutron source R ampersand D and issues

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Hasegawa, Mitsunori

    1997-01-01

    An inertial electrostatic confinement (IEC) fusion is the scheme of injecting the ions and electrons toward the spherical center, trapping both species in the electrostatic self-field and giving rise to fusion reactions in the dense core. An IEC is expected to have wide application from a small neutron source to a D- 3 He fusion reactor. Hirsch reported 10 9 n/s deuterium-tritium (D-T) neutron production in the device equipped with ion guns. Recently, Gu et al. measured 10 6 n/s using a D 2 gas discharge between the spherical wire cathode and the anode vacuum vessel, where the applied voltage is 60 kV and the current is 15 mA. We have also obtained similar neutron production at a lower voltage, ∼45 kV in a single-grid IEC device. Fusion reaction rates obtained by IEC experiments so far cannot be explained by a model of a simple potential well structure because the electrical potential peaked at the center prevents making a dense core. Hirsch proposed a multiwell structure called open-quotes poissorsclose quotes to explain the experiments. It is generally believed that there may be some correlation between the potential well structure and the neutron production rate. The scaling of neutron production on the injected ion current is a most important aspect of the problem for the prospect of utilizing IEC for fusion energy. The potential structure and its behavior are keys to the physics in understanding the principle of an IEC

  11. Study of ion flow dynamics in an inertial electrostatic confinement device through sequential grid construction

    International Nuclear Information System (INIS)

    Murali, S. Krupakar; Kulcinski, G. L.; Santarius, J. F.

    2008-01-01

    Experiments were performed to understand the dynamics of the ion flow in an inertial electrostatic confinement (IEC) device. This was done by monitoring the fusion rate as the symmetry of the grid was increased starting with a single loop all the way until the entire grid is constructed. The fusion rate was observed to increase with grid symmetry and eventually saturate. A single loop grid was observed to generate a cylindrical (∼line) fusion source. The ion flow distribution was measured by introducing fine wires across a single loop of the grid in the form of a chord of a circle (chord wires). This study revealed that with increased symmetry of the cathode grid wires the convergence of the ions improves. The chord wires provided electrons for ionization even at low pressures (∼6.67 mPa) and helped sustain the plasma. The impinging ions heat these wires locally and the temperature of the wires was measured using an infrared thermometer that was used to understand the ion flow distribution across the cathode grid. The presence of the grid wires seems to affect the fusion rate more drastically than previously thought (was assumed to be uniform around the central grid). Most of the fusion reactions were observed to occur in the ion microchannels that form in gaps between the cathode wires. This work helps understand the fusion source regimes and calibrate the IEC device.

  12. Repetitive pulse accelerator technology for light ion inertial confinement fusion

    International Nuclear Information System (INIS)

    Buttram, M.T.

    1985-01-01

    Successful ignition of an inertial confinement fusion (ICF) pellet is calculated to require that several megajoules of energy be deposited in the pellet's centimeter-sized shell within 10 ns. This implies a driver power of several hundreds of terawatts and power density around 100 TW/cm 2 . The Sandia ICF approach is to deposit the energy with beams of 30 MV lithium ions. The first accelerator capable of producing these beams (PBFA II, 100 TW) will be used to study beam formation and target physics on a single pulse basis. To utilize this technology for power production, repetitive pulsing at rates that may be as high as 10 Hz will be required. This paper will overview the technologies being studied for a repetitively pulsed ICF accelerator. As presently conceived, power is supplied by rotating machinery providing 16 MJ in 1 ms. The generator output is transformed to 3 MV, then switched into a pulse compression system using laser triggered spark gaps. These must be synchronized to about 1 ns. Pulse compression is performed with saturable inductor switches, the output being 40 ns, 1.5 MV pulses. These are transformed to 30 MV in a self-magnetically insulated cavity adder structure. Space charge limited ion beams are drawn from anode plasmas with electron counter streaming being magnetically inhibited. The ions are ballistically focused into the entrances of guiding discharge channels for transport to the pellet. The status of component development from the prime power to the ion source will be reviewed

  13. Production of hollow microspheres for inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Cook, R.

    1994-12-01

    The targets used in inertial confinement fusion (ICF) experiments at the Lawrence Livermore National Laboratory are plastic capsules roughly 0.5 mm in diameter. The capsules, which typically have wall thicknesses from 20 to 60 μm, must possess extraordinary symmetry and concentricity and must have surface finishes of less than 1000 Angstrom peak-to-valley variation over surface contours of from 10 to 100's of μm. This paper reviews the fabrication of these capsules, focusing on the production of the thin-walled polystyrene microshell mandrel around which the capsule is built. The relationship between the capsule characteristics, especially surface finish, and capsule performance is discussed, as are the methods of surface characterization and modification necessary for experiments designed to study the effects of surface roughness on implosion dynamics. Targets for the next generation of ICF facilities using more powerful laser drivers will have to be larger while meeting the same or even more stringent symmetry and surface finish requirements. Some of the technologies for meeting these needs are discussed briefly

  14. Neutron imaging for inertial confinement fusion and molecular optic imaging

    International Nuclear Information System (INIS)

    Delage, O.

    2010-01-01

    Scientific domains that require imaging of micrometric/nano-metric objects are dramatically increasing (Plasma Physics, Astrophysics, Biotechnology, Earth Sciences...). Difficulties encountered in imaging smaller and smaller objects make this research area more and more challenging and in constant evolution. The two scientific domains, through which this study has been led, are the neutron imaging in the context of the inertial confinement fusion and the fluorescence molecular imaging. Work presented in this thesis has two main objectives. The first one is to describe the instrumentation characteristics that require such imagery and, relatively to the scientific domains considered, identify parameters likely to optimize the imaging system accuracy. The second one is to present the developed data analysis and reconstruction methods able to provide spatial resolution adapted to the size of the observed object. Similarities of numerical algorithms used in these two scientific domains, which goals are quiet different, show how micrometric/nano-metric object imaging is a research area at the border of a large number of scientific disciplines. (author)

  15. Modeling Small-Amplitude Perturbations in Inertial Confinement Fusion Pellets

    Science.gov (United States)

    Zalesak, Steven; Metzler, N.; Velikovich, A. L.; Gardner, J. H.; Manheimer, W.

    2005-10-01

    Recent advances in inertial confinement fusion (ICF) technology serve to ensure that imploding laser-driven ICF pellets will spend a significantly larger portion of their time in what is regarded as the ``linear'' portion of their perturbation evolution, i.e., in the presence of small-amplitude but nonetheless evolving perturbations. Since the evolution of these linear perturbations collectively form the initial conditions for the subsequent nonlinear evolution of the pellet, which in turn determines the energy yield of the pellet, the accurate numerical modeling of these small-amplitude perturbations has taken on an increased importance. This modeling is difficult despite the expected linear evolution of the perturbations themselves, because these perturbations are embedded in a highly nonlinear, strongly-shocked, and highly complex flow field which in and of itself stresses numerical computation capabilities, and whose simulation often employs numerical techniques which were not designed with the proper treatment of small-amplitude perturbations in mind. In this paper we will review some of the techniques that we have recently found to be of use toward this end.

  16. Statistical Relations for Yield Degradation in Inertial Confinement Fusion

    Science.gov (United States)

    Woo, K. M.; Betti, R.; Patel, D.; Gopalaswamy, V.

    2017-10-01

    In inertial confinement fusion (ICF), the yield-over-clean (YOC) is a quantity commonly used to assess the performance of an implosion with respect to the degradation caused by asymmetries. The YOC also determines the Lawson parameter used to identify the onset of ignition and the level of alpha heating in ICF implosions. In this work, we show that the YOC is a unique function of the residual kinetic energy in the compressed shell (with respect to the 1-D case) regardless of the asymmetry spectrum. This result is derived using a simple model of the deceleration phase as well as through an extensive set of 3-D radiation-hydrodynamics simulations using the code DEC3D. The latter has been recently upgraded to include a 3-D spherical moving mesh, the HYPRE solver for 3-D radiation transport and piecewise-parabolic method for robust shock-capturing hydrodynamic simulations. DEC3D is used to build a synthetic single-mode database to study the behavior of yield degradation caused by Rayleigh-Taylor instabilities in the deceleration phase. The relation between YOC and residual kinetic energy is compared with the result in an adiabatic implosion model. The statistical expression of YOC is also applied to the ignition criterion in the presence of multidimensional nonuniformities. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  17. Shock-timing experiments for Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Debras, G.

    2012-01-01

    The Laser Megajoule (LMJ), which should achieve energy gain in an indirect drive inertial confinement fusion configuration, is being built in France by the CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives). To achieve thermonuclear ignition, the compression of a spherical target will have to be controlled by a series of accurately timed centripetal shocks, with a finely tuned level. A first experiment, performed in 2010 on the LIL (Ligne d'Integration Laser) facility at CEA, has allowed us to study the coalescence of two planar shocks in an indirectly-driven sample of polystyrene, within the framework of shock timing. The main objectives were to validate the experimental concept and the numerical simulations, as a proof-of-principle for future shock-timing campaigns. The main diagnostics used for this study are VISAR (Velocity Interferometer System for Any Reflection) and an optical shock breakout diagnostic, taking into account optical perturbations caused by X-rays. In another experiment, conducted on the LULI (Laboratoire pour l'Utilisation des Lasers Intenses) laser facility in 2010, we studied the timing of two planar directly-driven shocks using the same diagnostics. This latter study is related to the shock ignition concept, with the long-term perspective of energy production. This thesis presents these two experiments and their results. (author) [fr

  18. Philosophy and physics of predemonstration fusion devices

    International Nuclear Information System (INIS)

    Clarke, J.F.

    1976-01-01

    A PDFD will operate in the 1980's and must provide the plasma and plasma support technology information necessary to warrant design, construction, and operation of succeeding experimental power reactors and then the demonstration plant. The PDFD must be prototypical of economic fusion devices to justify its cost. Therefore, development of the fusion core will be the focus of the PDFD. The physics performance, power production objectives, and characteristics of the PDFD, and their relationship to the research and development needs to achieve them are outlined. The design criteria for a PDFD which satisfied these constraints will be established

  19. Possible in-lattice confinement fusion (LCF). Dynamic application of atomic and nuclear data

    International Nuclear Information System (INIS)

    Kawarasaki, Yuuki

    1995-01-01

    New scheme of a nuclear fusion reactor system is proposed, the basic concept of which comes from ingenious combination of hitherto developed techniques and verified facts; 1) so-called cold fusion (CF), 2) plasma of both magnetic confinement fusion (MCF) and inertial confinement fusion (ICF), and 3) accelerator-based D-T(D) neutron source. Details of the LCF reactor physics require dynamics of atomic data as well as nuclear data; interaction of ions with matters in solid and the problems of radiation damage. (author)

  20. The scientific benefits of inertially confined fusion research

    International Nuclear Information System (INIS)

    Key, M

    1999-01-01

    A striking feature of 25 years of research into inertially confined fusion (ICF) and inertial fusion energy (IFE) has been its significant impact in other fields of science. Most ICF facilities worldwide are now being used in part to support a wider portfolio of research than simply ICF. Reasons for this trend include the high intrinsic interest of the new science coupled with the relative ease and low marginal cost of adapting the facilities particularly lasers, to carry out experiments with goals other than ICF. The availability at ICF laboratories of sophisticated theory and modeling capability and advanced diagnostics has given added impetus. The expertise of ICF specialists has also triggered more lateral scientific spin-offs leading for example to new types of lasers and to related developments in basic science. In a generic sense, the facilities developed for ICF have made possible study of new regimes of the properties of matter at extremely high-energy density and the interaction of ultraintense light with matter. This general opportunity has been exploited in numerous and diverse specific lines of research. Examples elaborated below include laboratory simulation of astrophysical phenomena; studies of the equation of state (EOS) of matter under conditions relevant to the interior of planets and stars; development of uniquely intense sources of extreme ultraviolet (EUV) to hard x-ray emission, notably the x-ray laser; understanding of the physics of strong field interaction of light and matter; and related new phenomena such as laser-induced nuclear processes and high-field-electron accelerators. Some of these developments have potential themselves for further scientific exploitation such as the scientific use of advanced light sources. There are also avenues for commercial exploitation, for example the use of laser plasma sources in EUV lithography. Past scientific progress is summarized here and projections are made for new science that may flow from the

  1. Development of superconducting equipment for fusion device

    International Nuclear Information System (INIS)

    Konno, Masayuki; Ueda, Toshio; Hiue, Hisaaki; Ohgushi, Kouzou

    1993-01-01

    At Fuji Electric Co., Ltd., the development of superconductivity was started from 1960, and superconducting equipment for fusion device has been developed for ten years. The superconducting equipment, which is developed for fusion by Fuji Electric Co., Ltd., are able to be grouped in three categories which are current lead, superconducting coil and superconducting bus-line. The current lead is an electrical feeder between a superconducting coil and an electrical power supply. The rated current of developed current lead is 30kA at continuous use and 100kA at short time use respectively. The advanced disk type coil is developed for the toroidal field coil and some coils are developed for critical current measurement. Superconductor is applied to the superconducting bus-line between the superconducting coils and the current leads, and the bus-line is being developed for the Large Helical Device. This report describes an abstract of these equipment. (author)

  2. IAEA technical committee meeting on research using small fusion devices (abstracts)

    International Nuclear Information System (INIS)

    1999-12-01

    The thirteenth IAEA technical committee meeting on research using small fusion devices are held in Chengdu, P. R. China on 18-20 Oct. , 1999. 41 articles are received and the content includes toroidal systems, helical systems, plasma focus, diagnostic systems, theory and modeling, improving confinement, numerical simulation, innovative concepts and others

  3. Contribution to the numerical modeling of inertial confinement fusion

    International Nuclear Information System (INIS)

    Maire, P.H.

    2011-02-01

    This work was realized by writing the CHIC code, which is a software for designing and restoring experience in the field of inertial confinement fusion. The theoretical model describing the implosion of a laser target is a system of partial differential equations in the center of which is the Euler equations written in Lagrangian formalism, coupled with diffusion equations modeling the nonlinear transport of energy by electrons and photons. After a brief review of the physical context, we describe two novel methods which constitute the backbone of the CHIC code. These are 2 high-order finite volume schemes respectively dedicated to solving the equations of Lagrangian hydrodynamics and the anisotropic diffusion equations on bi-dimensional unstructured grids. The first scheme, called EUCCLHYD (Explicit Unstructured Lagrangian Hydrodynamics), solves the equations of gas dynamics on a moving mesh that moves at the speed of light. It is obtained from a general formalism based on the concept of sub-cell forces. In this context, the numerical fluxes are expressed in terms of the sub-cell force and the nodal velocity. Their determination is based on 3 basic principles: geometric compatibility between the movement of nodes and the volume change of mesh (geometric conservation law), compatibility with the second law of thermodynamics and conservation of total energy and momentum. The high-order extension is performed using a method based on solving a generalized Riemann problem in the acoustic approximation. The second scheme, called CCLAD (Cell-Centered Lagrangian Diffusion), solves the anisotropic heat equation. The corresponding discretization relies on a discrete variational formulation based on the sub-cell that allows to build a multipoint approximation of heat flux. This high-order discretization makes possible the resolution of the equations of anisotropic diffusion with satisfactory accuracy on highly distorted Lagrangian meshes. (author)

  4. Mixing with applications to inertial-confinement-fusion implosions

    Science.gov (United States)

    Rana, V.; Lim, H.; Melvin, J.; Glimm, J.; Cheng, B.; Sharp, D. H.

    2017-01-01

    Approximate one-dimensional (1D) as well as 2D and 3D simulations are playing an important supporting role in the design and analysis of future experiments at National Ignition Facility. This paper is mainly concerned with 1D simulations, used extensively in design and optimization. We couple a 1D buoyancy-drag mix model for the mixing zone edges with a 1D inertial confinement fusion simulation code. This analysis predicts that National Ignition Campaign (NIC) designs are located close to a performance cliff, so modeling errors, design features (fill tube and tent) and additional, unmodeled instabilities could lead to significant levels of mix. The performance cliff we identify is associated with multimode plastic ablator (CH) mix into the hot-spot deuterium and tritium (DT). The buoyancy-drag mix model is mode number independent and selects implicitly a range of maximum growth modes. Our main conclusion is that single effect instabilities are predicted not to lead to hot-spot mix, while combined mode mixing effects are predicted to affect hot-spot thermodynamics and possibly hot-spot mix. Combined with the stagnation Rayleigh-Taylor instability, we find the potential for mix effects in combination with the ice-to-gas DT boundary, numerical effects of Eulerian species CH concentration diffusion, and ablation-driven instabilities. With the help of a convenient package of plasma transport parameters developed here, we give an approximate determination of these quantities in the regime relevant to the NIC experiments, while ruling out a variety of mix possibilities. Plasma transport parameters affect the 1D buoyancy-drag mix model primarily through its phenomenological drag coefficient as well as the 1D hydro model to which the buoyancy-drag equation is coupled.

  5. Rayleigh-Taylor instability in inertial confinement fusion

    International Nuclear Information System (INIS)

    Gupta, N.K.

    1987-01-01

    This report summarises the main results of theoretical analysis on the problem of Rayleigh-Tylor instability in inertial confinement fusion (ICF). Work presented in this report essentially covers four basic problems. Firstly, an analytical formulation to analyse the effects of plasma density inhomogeneities on the growth of the instability in plane geometry is presented. As a result of this analysis it is concluded that, for minimizing the growth rate of the instability, it may be advantageous to use the driver laser beams of higher irradiance and an optimum wave length in an ICF experiment. Secondly, a new formulation for the analysis of the instability in curved (cylindrical and spherical) geometries is presented. A general eigenvalue equation for the growth rate of the instability which is applicable for both plane and curved geometries is derived. A comparative study is made between the plane, cylindrical and spherical geometries. Also analytical expressions for the growth rates are obtained in the cases of spherical and cylindrical shell targets and their variations with respect to the aspect ratios of the shells are discussed. Thirdly, a semi-analytical analysis of the instability where the growth rate is obtained by solving numerically a (2N-1)x(2N-1) determinantal equation is presented. The semi-analytical analysis developed is applicable for the study of the growth of the instability in the present day multi-structured spherical shell targets. Finally, a dynamic analysis of the growth of the instability for a representative spherical solid target driven by laser beams symmetrically from all the sides is carried out numerically using a computer code developed for this purpose. This study confirms analytical predictions. Further, it is observed that an approximate analytical analysis with time independent density profile gives conservative estimates for the growth rate. In passing, the computer code is also used to estimate the pellet gain for spin

  6. Study of the shock ignition scheme in inertial confinement fusion

    International Nuclear Information System (INIS)

    Lafon, M.

    2011-01-01

    The Shock Ignition (SI) scheme is an alternative to classical ignition schemes in Inertial Confinement Fusion. Its singularity relies on the relaxation of constraints during the compression phase and fulfilment of ignition conditions by launching a short and intense laser pulse (∼500 ps, ∼300 TW) on the pre-assembled fuel at the end of the implosion.In this thesis, it has been established that the SI process leads to a non-isobaric fuel configuration at the ignition time thus modifying the ignition criteria of Deuterium-Tritium (DT) against the conventional schemes. A gain model has been developed and gain curves have been inferred and numerically validated. This hydrodynamical modeling has demonstrated that the SI process allows higher gain and lower ignition energy threshold than conventional ignition due to the high hot spot pressure at ignition time resulting from the ignitor shock propagation.The radiative hydrodynamic CHIC code developed at the CELIA laboratory has been used to determine parametric dependences describing the optimal conditions for target design leading to ignition. These numerical studies have enlightened the potential of SI with regards to saving up laser energy, obtain high gains but also to safety margins and ignition robustness.Finally, the results of the first SI experiments performed in spherical geometry on the OMEGA laser facility (NY, USA) are presented. An interpretation of the experimental data is proposed from mono and bidimensional hydrodynamic simulations. Then, different trails are explored to account for the differences observed between experimental and numerical data and alternative solutions to improve performances are suggested. (author) [fr

  7. X-ray sources by Z-pinch for inertial confinement fusion

    International Nuclear Information System (INIS)

    Akiyama, Hidenori; Katsuki, Sunao; Lisitsyn, Igor

    1999-01-01

    Inertial confinement nuclear fusion driven by X-ray from Z-pinch plasmas has been developed. Recently, extremely high X-ray power (290 TW) and energy (1.8 MJ) were produced in fast Z-pinch implosions on the Z accelerator (Sandia National Laboratories). Wire arrays are used to produce the initial plasma. The X-ray from Z-pinch plasmas produced by pulsed power has great potential as a driver of inertial confinement nuclear fusion. (author)

  8. Inertial Confinement Fusion quarterly report, January--March 1995. Volume 5, No. 2

    International Nuclear Information System (INIS)

    1995-01-01

    The ICF quarterly report is published by the Inertial Confinement Fusion Program at the Lawrence Livermore National Laboratory. Topics included this quarter include: the role of the National Ignition Facility in the development of Inertial Confinement Fusion, laser-plasma interactions in large gas-filled hohlraums, evolution of solid-state induction modulators for a heavy-ion recirculator, the National Ignition Facility project, and terminal-level relaxation in Nd-doped laser material

  9. Structural properties of hydrogen isotopes in solid phase in the context of inertial confinement fusion

    Directory of Open Access Journals (Sweden)

    Guerrero Carlo

    2013-11-01

    Full Text Available Quality of Deuterium-Tritium capsules is a critical aspect in Inertial Confinement Fusion. In this work, we present a Quantum Molecular Dynamics methodology able to model hydrogen isotopes and their structural molecular organisation at extreme pressures and cryogenic temperatures (< 15 K. Our study sets up the basis for a future analysis on the mechanical and structural properties of DT-ice in inertial confinement fusion (ICF target manufacturing conditions.

  10. 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.

  11. Direct-drive inertial confinement fusion: A review

    International Nuclear Information System (INIS)

    Craxton, R. S.; Anderson, K. S.; Boehly, T. R.; Goncharov, V. N.; Harding, D. R.; Knauer, J. P.; McKenty, P. W.; Myatt, J. F.; Short, R. W.; Skupsky, S.; Theobald, W.; Collins, T. J. B.; Delettrez, J. A.; Hu, S. X.; Marozas, J. A.; Maximov, A. V.; Michel, D. T.; Radha, P. B.; Regan, S. P.; Sangster, T. C.

    2015-01-01

    The direct-drive, laser-based approach to inertial confinement fusion (ICF) is reviewed from its inception following the demonstration of the first laser to its implementation on the present generation of high-power lasers. The review focuses on the evolution of scientific understanding gained from target-physics experiments in many areas, identifying problems that were demonstrated and the solutions implemented. The review starts with the basic understanding of laser–plasma interactions that was obtained before the declassification of laser-induced compression in the early 1970s and continues with the compression experiments using infrared lasers in the late 1970s that produced thermonuclear neutrons. The problem of suprathermal electrons and the target preheat that they caused, associated with the infrared laser wavelength, led to lasers being built after 1980 to operate at shorter wavelengths, especially 0.35 μm—the third harmonic of the Nd:glass laser—and 0.248 μm (the KrF gas laser). The main physics areas relevant to direct drive are reviewed. The primary absorption mechanism at short wavelengths is classical inverse bremsstrahlung. Nonuniformities imprinted on the target by laser irradiation have been addressed by the development of a number of beam-smoothing techniques and imprint-mitigation strategies. The effects of hydrodynamic instabilities are mitigated by a combination of imprint reduction and target designs that minimize the instability growth rates. Several coronal plasma physics processes are reviewed. The two-plasmon–decay instability, stimulated Brillouin scattering (together with cross-beam energy transfer), and (possibly) stimulated Raman scattering are identified as potential concerns, placing constraints on the laser intensities used in target designs, while other processes (self-focusing and filamentation, the parametric decay instability, and magnetic fields), once considered important, are now of lesser concern for mainline

  12. Direct-drive inertial confinement fusion: A review

    Energy Technology Data Exchange (ETDEWEB)

    Craxton, R. S.; Anderson, K. S.; Boehly, T. R.; Goncharov, V. N.; Harding, D. R.; Knauer, J. P.; McKenty, P. W.; Myatt, J. F.; Short, R. W.; Skupsky, S.; Theobald, W.; Collins, T. J. B.; Delettrez, J. A.; Hu, S. X.; Marozas, J. A.; Maximov, A. V.; Michel, D. T.; Radha, P. B.; Regan, S. P.; Sangster, T. C. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623-1299 (United States); and others

    2015-11-15

    The direct-drive, laser-based approach to inertial confinement fusion (ICF) is reviewed from its inception following the demonstration of the first laser to its implementation on the present generation of high-power lasers. The review focuses on the evolution of scientific understanding gained from target-physics experiments in many areas, identifying problems that were demonstrated and the solutions implemented. The review starts with the basic understanding of laser–plasma interactions that was obtained before the declassification of laser-induced compression in the early 1970s and continues with the compression experiments using infrared lasers in the late 1970s that produced thermonuclear neutrons. The problem of suprathermal electrons and the target preheat that they caused, associated with the infrared laser wavelength, led to lasers being built after 1980 to operate at shorter wavelengths, especially 0.35 μm—the third harmonic of the Nd:glass laser—and 0.248 μm (the KrF gas laser). The main physics areas relevant to direct drive are reviewed. The primary absorption mechanism at short wavelengths is classical inverse bremsstrahlung. Nonuniformities imprinted on the target by laser irradiation have been addressed by the development of a number of beam-smoothing techniques and imprint-mitigation strategies. The effects of hydrodynamic instabilities are mitigated by a combination of imprint reduction and target designs that minimize the instability growth rates. Several coronal plasma physics processes are reviewed. The two-plasmon–decay instability, stimulated Brillouin scattering (together with cross-beam energy transfer), and (possibly) stimulated Raman scattering are identified as potential concerns, placing constraints on the laser intensities used in target designs, while other processes (self-focusing and filamentation, the parametric decay instability, and magnetic fields), once considered important, are now of lesser concern for mainline

  13. Pulsed power ion accelerators for inertially confined fusion

    International Nuclear Information System (INIS)

    Olson, C.L.

    1976-01-01

    Current research is described on pulsed power ion accelerators for inertial fusion, i.e., ion diodes and collective accelerators. Particle beam energy and power requirements for fusion, and basic deposition characteristics of charged particle beams are discussed. Ion diodes and collective accelerators for fusion are compared with existing conventional accelerators

  14. The role of edge physics and confinement issues in the fusion reactor

    International Nuclear Information System (INIS)

    Lackner, K.; Schneider, R.

    1993-01-01

    The compatibility of the power exhaust to the walls and target plates with an adequate survival time of these structures and a sufficiently low impurity concentration in the bulk plasma, on the one hand, and the scaling of the energy confinement time with the tokamak parameters, on the other, are presently seen as two of the most critical issues determining the viability and minimum size of a fusion reactor. Statistical analysis of the world-wide data base of divertor tokamaks and experimental progress in superposing different forms of confinement improvements instil confidence that tokamak devices with the dimensions of ITER can meet the energy confinement requirements for ignition when operating in the H-regime. The critical problem consists, however, in simultaneously satisfying both the impurity and energy confinement related requirements. Regimes of improved confinement also tend to show a reduced rate of outward flow of impurities, thus increasing (for a given production rate at the walls or - in the case of helium - in the interior) their expected concentration. The confinement improvement in the H-regime also extends beyond the separatrix, leading to a thinner scrape-off layer with enhanced power concentration at the target plate and an enhanced penetration of impurities into the bulk plasma. These issues have therefore become the focal point of divertor and H-regime studies. Experiments on ASDEX, DIII-D and JET have successfully demonstrated that the impurity concentration in the H-regime can be effectively limited by the spontaneous cyclic destruction of the transport barrier at the edge (through the so-called ELMs) with only partial sacrifice of the energy confinement improvement. Expectations of handling the power load issues hinge on the capability of the divertor solution to decouple plasma properties in the edge of the bulk plasma from those near the target plates and the possibility of converting the energy flux arriving in the divertor into

  15. 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)

  16. Development of heavy-ion accelerators as drivers for inertially confined fusion

    International Nuclear Information System (INIS)

    Herrmannsfeldt, W.B.

    1979-06-01

    The commercialization of inertial confinement fusion is discussed in terms of power costs. A chapter on heavy ion accelerators covers the prinicpal components, beam loss mechanisms, and theoretical considerations. Other tyopics discussed include the following: (1) heavy ion fusion implementation plan, (2) driver with accumulator rings fed by an rf LINAC, (3) single pass driver with an induction LINAC, and (4) implementation scenarios

  17. Inertial Confinement Fusion quarterly report, April--June 1995. Volume 5, No. 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The ICF Quarterly Reports is published four times each fiscal year by the Inertial Confinement Fusion Program at the Lawrence Livermore National Laboratory. The journal reports selected current research within the ICF Program. Major areas of investigation presented here include fusion target theory and design, target fabrication, target experiments, and laser and optical science and technology.

  18. 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.

  19. High density plasmas formation in Inertial Confinement Fusion and Astrophysics

    International Nuclear Information System (INIS)

    Martinez-Val, J. M.; Minguez, E.; Velarde, P.; Perlado, J. M.; Velarde, G.; Bravo, E.; Eliezer, S.; Florido, R.; Garcia Rubiano, J.; Garcia-Senz, D.; Gil de la Fe, J. M.; Leon, P. T.; Martel, P.; Ogando, F.; Piera, M.; Relano, A.; Rodriguez, R.; Garcia, C.; Gonzalez, E.; Lachaise, M.; Oliva, E.

    2005-01-01

    In inertially confined fusion (ICF), high densities are required to obtain high gains. In Fast Ignition, a high density, low temperature plasma can be obtained during the compression. If the final temperature reached is low enough, the electrons of the plasma can be degenerate. In degenerate plasmas. Bremsstrahlung emission is strongly suppressed an ignition temperature becomes lower than in classical plasmas, which offers a new design window for ICF. The main difficulty of degenerate plasmas in the compression energy needed for high densities. Besides that, the low specific heat of degenerate electrons (as compared to classical values) is also a problem because of the rapid heating of the plasma. Fluid dynamic evolution of supernovae remnants is a very interesting problem in order to predict the thermodynamical conditions achieved in their collision regions. Those conditions have a strong influence in the emission of light and therefore the detection of such events. A laboratory scale system has been designed reproducing the fluid dynamic field in high energy experiments. The evolution of the laboratory system has been calculated with ARWEN code, 2D Radiation CFD that works with Adaptive Mesh Refinement. Results are compared with simulations on the original system obtained with a 3D SPH astrophysical code. New phenomena at the collision plane and scaling of the laboratory magnitudes will be described. Atomic physics for high density plasmas has been studied with participation in experiments to obtain laser produced high density plasmas under NLTE conditions, carried out at LULI. A code, ATOM3R, has been developed which solves rate equations for optically thin plasmas as well as for homogeneous optically thick plasmas making use of escape factors. New improvements in ATOM3R are been done to calculate level populations and opacities for non homogeneous thick plasmas in NLTE, with emphasis in He and H lines for high density plasma diagnosis. Analytical expression

  20. ICRF Traveling Wave launcher for fusion devices

    International Nuclear Information System (INIS)

    Ragona, R

    2017-01-01

    Ion Cyclotron Resonance Heating and Current Drive is a method that has the ability to heat directly the ions in the Deuterium-Tritrium fuel to the high temperature needed for the fusion reaction to works. The capability of efficiently couple the Radio Frequency power to the plasma plays a big role in the overall performance of a fusion device. A Traveling Wave Antenna in a resonant ring configuration is a good candidate for an Ion Cyclotron Resonance Heating and Current Drive system. It has the capability to increase the coupled power with respect to present designs and to have a highly selective power spectrum that can be peaked around the maximally absorbed wave. It is also insensitive to the loading variations due to fluctuation of the plasma edge increasing the reliability and the efficiency of the system. It works as a low power density launcher due to the possible large number of current carrying elements. (paper)

  1. Potential well formation in electrostatic confinement devices. Technical summary report

    International Nuclear Information System (INIS)

    Cherrington, B.E.; Verdeyen, J.T.

    1978-01-01

    The experimental and theoretical studies on Inertial Electrostatic Plasma Confinement that have been performed in the Gaseous Electronics Laboratory of the University of Illinois are reviewed. There has been experimental confirmation of the production of a multiple potential structure in both small and large spherical devices and the theoretical analysis has indicated the parameter range that is necessary in order to explain such results. Further experimental and theoretical approaches to testing the IEPC concept are suggested

  2. Simulation of transition dynamics to high confinement in fusion plasmas

    DEFF Research Database (Denmark)

    Nielsen, Anders Henry; Xu, G. S.; Madsen, Jens

    2015-01-01

    The transition dynamics from the low (L) to the high (H) confinement mode in magnetically confined plasmas is investigated using a first-principles four-field fluid model. Numerical results are in agreement with measurements from the Experimental Advanced Superconducting Tokamak - EAST...

  3. Fusion. An introduction of the physics and technology of magnetic confinement fusion. 2. compl. rev. and enl. ed.

    International Nuclear Information System (INIS)

    Stacey, Weston M.

    2010-01-01

    This second edition of a popular textbook is thoroughly revised with around 25% new and updated content. It provides an introduction to both plasma physics and fusion technology at a level that can be understood by advanced undergraduates and graduate students in the physical sciences and related engineering disciplines. As such, the contents cover various plasma confinement concepts, the support technologies needed to confine the plasma, and the designs of ITER as well as future fusion reactors. With end of chapter problems for use in courses. (orig.)

  4. Initiative taken by India in magnetically confined fusion reactor

    International Nuclear Information System (INIS)

    Bora, Dhiraj

    2017-01-01

    There is a growing gap between demand and supply of energy in the world. Any attempt to develop new and cleaner sources of energy to meet the future global requirement is welcome. Therefore, it is attractive to think of having fusion as an alternate clean source of energy to contribute in the energy mix towards the second half of the century, with a virtually inexhaustible fuel supply. The environmental impact of fusion would be acceptable and relatively safe. These advantages have driven the world fusion research programme since its inception. Indian progress in fusion science and technology and participation in ITER will be discussed during the talk

  5. EMP Fusion

    OpenAIRE

    KUNTAY, Isık

    2010-01-01

    This paper introduces a novel fusion scheme, called EMP Fusion, which has the promise of achieving breakeven and realizing commercial fusion power. The method is based on harnessing the power of an electromagnetic pulse generated by the now well-developed flux compression technology. The electromagnetic pulse acts as a means of both heating up the plasma and confining the plasma, eliminating intermediate steps. The EMP Fusion device is simpler compared to other fusion devices and this reduces...

  6. System and method of operating toroidal magnetic confinement devices

    Science.gov (United States)

    Chance, Morrell S.; Jardin, Stephen C.; Stix, Thomas H.; Grimm, deceased, Ray C.; Manickam, Janardhan; Okabayashi, Michio

    1987-01-01

    For toroidal magnetic confinement devices the second region of stability against ballooning modes can be accessed with controlled operation. Under certain modes of operation, the first and second stability regions may be joined together. Accessing the second region of stability is accomplished by forming a bean-shaped plasma and increasing the indentation until a critical value of indentation is reached. A pusher coil, located at the inner-major-radius side of the device, is engaged to form a bean-shaped poloidal cross-section in the plasma.

  7. Pressure measurements in magnetic-fusion devices

    International Nuclear Information System (INIS)

    Dylla, H.F.

    1981-11-01

    Accurate pressure measurements are important in magnetic fusion devices for: (1) plasma diagnostic measurements of particle balance and ion temperature; (2) discharge cleaning optimization; (3) vacuum system performance; and (4) tritium accountability. This paper reviews the application, required accuracy, and suitable instrumentation for these measurements. Demonstrated uses of ionization-type and capacitance-diaphragm gauges for various pressure and gas-flow measurements in tokamaks are presented, with specific reference to the effects of magnetic fields on gauge performance and the problems associated with gauge calibration

  8. Plasma Surface interaction in Controlled fusion devices

    International Nuclear Information System (INIS)

    1990-05-01

    The subjects presented in the 9th conference on plasma surface interaction in controlled fusion devices were: the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the effects observed in ergodic divertor experiments in Tore-Supra; the diffuse connexion induced by the ergodic divertor and the topology of the heat load patterns on the plasma facing components in Tore-Supra; the study of the influence of air exposure on graphite implanted by low energy high density deuterium plasma

  9. Protector in a nuclear fusion device

    International Nuclear Information System (INIS)

    Furukawa, Masayuki; Yamane, Katsumi; Niwa, Sadahiko; Ogata, Fumio; Masuda, Jun-ichi.

    1975-01-01

    Object: To block an abnormal voltage, which shifts from plasma to coil or power supply by means of action of mutual induction, by a circuit utilizing non-linear impedance elements. Structure: The nuclear fusion device includes a current transformer coil, a vertical field coil and a plasma circuit, with a non-linear impedance element disposed in parallel with at least the current transformer coil, said impedance element being disposed in parallel with a short-circuiting switch, relative to the abnormal voltage moving from the plasma by means of action of mutual induction. (Kamimura, M.)

  10. Pressure measurements in magnetic-fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Dylla, H.F.

    1981-11-01

    Accurate pressure measurements are important in magnetic fusion devices for: (1) plasma diagnostic measurements of particle balance and ion temperature; (2) discharge cleaning optimization; (3) vacuum system performance; and (4) tritium accountability. This paper reviews the application, required accuracy, and suitable instrumentation for these measurements. Demonstrated uses of ionization-type and capacitance-diaphragm gauges for various pressure and gas-flow measurements in tokamaks are presented, with specific reference to the effects of magnetic fields on gauge performance and the problems associated with gauge calibration.

  11. Historic overview of inertial confinement fusion: What have we learned

    International Nuclear Information System (INIS)

    Glass, A.J.

    1986-01-01

    Although laser fusion has been the subject of research since the early 1960s, it has only been intensively studied for about 14 years. During that time, substantive advances have been made in our understanding of the complex physics of laser-heated plasmas, in the development of sophisticated diagnostic instrumentation, and in the technology of fusion targets and inertial fusion drivers. These advances will be reviewed. Of equal importance are the lessons learned in the economic and political arenas. These lessons may be of greater significance for scientific endeavors in other fields of research. The economic and political issues surrounding inertial fusion research will be discussed. Possible future directions for inertial fusion development will be presented

  12. Classical impurity ion confinement in a toroidal magnetized fusion plasma.

    Science.gov (United States)

    Kumar, S T A; Den Hartog, D J; Caspary, K J; Magee, R M; Mirnov, V V; Chapman, B E; Craig, D; Fiksel, G; Sarff, J S

    2012-03-23

    High-resolution measurements of impurity ion dynamics provide first-time evidence of classical ion confinement in a toroidal, magnetically confined plasma. The density profile evolution of fully stripped carbon is measured in MST reversed-field pinch plasmas with reduced magnetic turbulence to assess Coulomb-collisional transport without the neoclassical enhancement from particle drift effects. The impurity density profile evolves to a hollow shape, consistent with the temperature screening mechanism of classical transport. Corroborating methane pellet injection experiments expose the sensitivity of the impurity particle confinement time to the residual magnetic fluctuation amplitude.

  13. Eddy current analysis in fusion devices

    International Nuclear Information System (INIS)

    Turner, L.R.

    1988-06-01

    In magnetic fusion devices, particularly tokamaks and reversed field pinch (RFP) experiments, time-varying magnetic fields are in intimate contact with electrically conducting components of the device. Induced currents, fields, forces, and torques result. This note reviews the analysis of eddy current effects in the following systems: Interaction of a tokamak plasma with the eddy currents in the first wall, blanket, and shield (FWBS) systems; Eddy currents in a complex but two-dimensional vacuum vessel, as in TFTR, JET, and JT-60; Eddy currents in the FWBS system of a tokamak reactor, such as NET, FER, or ITER; and Eddy currents in a RFP shell. The cited studies are chosen to be illustrative, rather than exhaustive. 42 refs

  14. Computational modeling of the axial-cylindrical inertial electrostatic confinement fusion neutron generator

    Science.gov (United States)

    Bromley, Blair Patrick

    2001-12-01

    The axial-cylindrical Inertial Electrostatic Confinement fusion neutron generator (IEC C-Device) is a high- voltage, low-pressure glow discharge device that produces neutrons from the deuterium-deuterium fusion reaction. Such a neutron source has potential applications for neutron activation analysis and capture therapies for cancer treatment. The IEC C-Device operating with deuterium fuel is modeled with the CHIMP computer code developed and written completely by the author to predict the fusion neutron generation rate and the plasma physics behavior using fundamental first principles. The CHIMP code is a time-dependent, spatially two-dimensional (r,z), particle-in-cell, Monte-Carlo-Collision (PIC-MCC) direct simulation model. The effects of secondary electron emission due to ion and electron impact on the metal electrodes and the glass walls and charge build-up on the glass wall are included. Either monatomic or molecular ions and electrons are modeled in a monatomic or molecular background neutral deuterium gas. CHIMP code predictions are compared against experimental results for the C-Device operating between 10 and 30 kV of anode voltage, between 10 and 40 mA of electrode current, and between 0.29 and 1.1 milliTorr of deuterium gas pressure. A calibration factor for the pressure accounts for the calibration of the ionization pressure gauge in the experiment, and an estimated pressure drop between the main chamber of the C-Device and the pressure gauge that is downstream of the exhaust port. Upgraded versions of the CHIMP code which have modifications to the algorithms for the boundary conditions, and which include charge exchange processes, and the contribution of fast neutrals to the neutron generation rate are also tested against several experimental data points. Although the CHIMP code gives predictions for the neutron generation rate that exhibit the same near-linear trends with current found in the experiment, it is apparent that at least five types of

  15. The technology benefits of inertial confinement fusion research

    International Nuclear Information System (INIS)

    Powell, H.T.

    1999-01-01

    The development and demonstration of inertial fusion is incredibly challenging because it requires simultaneously controlling and precisely measuring parameters at extreme values in energy, space, and time. The challenges range from building megajoule (10 6 J) drivers that perform with percent-level precision to fabricating targets with submicron specifications to measuring target performance at micron scale (10 -6 m) with picosecond (10 -12 s) time resolution. Over the past 30 years in attempting to meet this challenge, the inertial fusion community around the world has invented new technologies in lasers, particle beams, pulse power drivers, diagnostics, target fabrication, and other areas. These technologies have found applications in diverse fields of industry and science. Moreover, simply assembling the teams with the background, experience, and personal drive to meet the challenging requirements of inertial fusion has led to spin-offs in unexpected directions, for example, in laser isotope separation, extreme ultraviolet lithography for microelectronics, compact and inexpensive radars, advanced laser materials processing, and medical technology. The experience of inertial fusion research and development of spinning off technologies has not been unique to any one laboratory or country but has been similar in main research centers in the US, Europe, and Japan. Strengthening and broadening the inertial fusion effort to focus on creating a new source of electrical power (inertial fusion energy [IFE]) that is economically competitive and environmentally benign will yield rich rewards in technology spin-offs. The additional challenges presented by IFE are to make drivers affordable, efficient, and long-lived while operating at a repetition rate of a few Hertz; to make fusion targets that perform consistently at high-fusion yield; and to create target chambers that can repetitively handle greater than 100-MJ yields while producing minimal radioactive by

  16. Controlled Nuclear Fusion by Magnetic Confinement and ITER

    CERN Document Server

    CERN. Geneva. Audiovisual Unit; Alvarez-Gaumé, Luís

    2005-01-01

    For may years harnessing fusion energy was considered the final solution to the world's energy crisis. ITER is the last step in the elusive quest. This presentation will provide in its various acientific, technological and political aspects.

  17. Energy confinement in the tokamak devices pulsator and ASDEX

    International Nuclear Information System (INIS)

    Klueber, O.; Murmann, H.

    1982-04-01

    The energy confinement of ohmically heated hydrogen plasmas obtained in the ASDEX and Pulsator tokamaks is investigated. In both devices, the confinement time does not follow a simple scaling law of the type tausub(E) approx. equal to nsub(e)a 2 . In the case of Pulsator, a regime is identified in which the transport is governed by electron heat conduction. The experimental data are compared with an analytic solution of the energy balance equation from which a heat diffusivity chisub(e) approx. equal to Zsub(eff)sup(1/3)/nsub(e)(r)Tsub(e)sup(1/2)(r)q(r) is inferred. chisub(i) is supposed to be neoclassical (plateau regime). Heat conduction following these laws is shown to lead to a consistent description of the full data set. (orig.)

  18. 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.)

  19. Advanced real-time control systems for magnetically confined fusion plasmas

    International Nuclear Information System (INIS)

    Goncalves, B.; Sousa, J.; Fernandes, H.; Rodrigues, A.P.; Carvalho, B.B.; Neto, A.; Varandas, C.A.F.

    2008-01-01

    Real-time control of magnetically confined plasmas is a critical issue for the safety, operation and high performance scientific exploitation of the experimental devices on regimes beyond the current operation frontiers. The number of parameters and the data volumes used for the plasma properties identification scale normally not only with the machine size but also with the technology improvements, leading to a great complexity of the plant system. A strong computational power and fast communication infrastructure are needed to handle in real-time this information, allowing just-in-time decisions to achieve the fusion critical plasma conditions. These advanced control systems require a tiered infrastructure including the hardware layer, the signal-processing middleware, real-time timing and data transport, the real-time operating system tools and drivers, the framework for code development, simulation, deployment and experiment parameterization and the human real-time plasma condition monitoring and management. This approach is being implemented at CFN by offering a vertical solution for the forthcoming challenges, including ITER, the first experimental fusion reactor. A given set of tools and systems are described on this paper, namely: (i) an ATCA based hardware multiple-input-multiple-output (MIMO) platform, PCI and PCIe acquisition and control modules; (ii) FPGA and DSP parallelized signal processing algorithms; (iii) a signal data and event distribution system over a 2.5/10Gb optical network with sub-microsecond latencies; (iv) RTAI and Linux drivers; and (v) the FireSignal, FusionTalk, SDAS FireCalc application tools. (author)

  20. Consideration on nuclear fusion in plasma by the magnetic confinement as a heat engine

    International Nuclear Information System (INIS)

    Tsuji, Yoshio

    1990-01-01

    In comparing nuclear fusion in plasma by the magnetic confinement with nuclear fission and chemical reactions, the power density and the function of a heat engine are discussed using a new parameter G introduced as an eigenvalue of a reaction and the value of q introduced to estimate the thermal efficiency of a heat engine. It is shown that the fusion reactor by the magnetic confinement is very difficult to be a modern heat engine because of the lack of some indispensable functions as a modern heat engine. The value of G and q have the important role in the consideration. (author)

  1. SAFIRE: A systems analysis code for ICF [inertial confinement fusion] reactor economics

    International Nuclear Information System (INIS)

    McCarville, T.J.; Meier, W.R.; Carson, C.F.; Glasgow, B.B.

    1987-01-01

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code incorporates analytical models for scaling the cost and performance of several inertial confinement fusion reactor concepts for electric power. The code allows us to vary design parameters (e.g., driver energy, chamber pulse rate, net electric power) and evaluate the resulting change in capital cost of power plant and the busbar cost of electricity. The SAFIRE code can be used to identify the most attractive operating space and to identify those design parameters with the greatest leverage for improving the economics of inertial confinement fusion electric power plants

  2. A Novel Spectrometer for Measuring Laser-Produced Plasma X-Ray in Inertial Confinement Fusion

    Directory of Open Access Journals (Sweden)

    Zhu Gang

    2012-01-01

    Full Text Available In the experimental investigations of inertial confinement fusion, the laser-produced high-temperature plasma contains very abundant information, such as the electron temperature and density, ionization. In order to diagnose laser-plasma distribution in space and evolution in time, an elliptical curved crystal spectrometer has been developed and applied to diagnose X-ray of laser-produced plasma in 0.2~2.46 nm region. According to the theory of Bragg diffraction, four kinds of crystal including LiF, PET, MiCa, and KAP were chosen as dispersive elements. The distance of crystal lattice varies from 0.4 to 2.6 nm. Bragg angle is in the range of 30°~67.5°, and the spectral detection angle is in 55.4°~134°. The curved crystal spectrometer mainly consists of elliptical curved crystal analyzer, vacuum configuration, aligning device, spectral detectors and three-dimensional microadjustment devices. The spectrographic experiment was carried out on the XG-2 laser facility. Emission spectrum of Al plasmas, Ti plasma, and Au plasmas have been successfully recorded by using X-ray CCD camera. It is demonstrated experimentally that the measured wavelength is accorded with the theoretical value.

  3. Dispersion interferometer for controlled fusion devices

    International Nuclear Information System (INIS)

    Drachev, V.P.; Krasnikov, Yu.I.; Bagryansky, P.A.

    1992-01-01

    A common feature in interferometry is the presence of two independent optical channels. Since wave phase in a medium depends on the geometrical path, polarization and radiation frequency, respectively, one can distinguish three types of interferometric schemes when the channels are geometrically separated, or separation occurs in polarizations or radiation frequencies. We have developed a measurement scheme based on a dispersion interferometer (DI) for plasma diagnostics in the experiments on controlled fusion. DI optical channels have the same geometrical path and are separated in radiation frequency. Use of a common optical path causes the main advantage of the DI technique - low sensitivity to vibrations of optical elements. The use of the DI technique for diagnostics of a laser spark in air and of arc discharges has shown its essential advantages as compared to classical interferometers. Interest in the DI technique from the viewpoint of its application in controlled fusion devices is determined also generated by the possibility of developing a compact multichannel interferometer not requiring a vibration isolation structure. (author) 14 refs., 3 figs

  4. Studies in the evolution of hydrodynamic instabilities and their role in inertial confinement fusion

    International Nuclear Information System (INIS)

    Shvarts, D.; Oron, D.; Sadot, O.

    2001-01-01

    Hydrodynamic instabilities, such as the Rayleigh-Taylor and Richtmyer-Meshkov instabilities, have a central role when trying to achieve net thermonuclear fusion energy via the method of Inertial Confinement Fusion. We shall review recent theoretical, numerical and experimental work that describes the evolution of two- and three-dimensional perturbations. Finally, the effects of these perturbation on the ignition conditions, using new self-similar solutions for perturbed burn wave propagation will be discussed. (author)

  5. 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

  6. Vacuum vessel for a nuclear fusion device

    International Nuclear Information System (INIS)

    Watanabe, Takashi; Sato, Hiroshi; Owada, Koro.

    1976-01-01

    Object: To provide a reinforcing member on a bellows portion to reduce a stress at the bellows portion thereby increasing the strength of a vessel. Structure: A vacuum vessel for a nuclear fusion device has a bellows portion and a wall thick portion. A support extended toward the bellows portion is secured inside of a toroidal section in order to reduce the stress at the bellows portion. An insulator is interposed between the support and the bellows portion and is retained on the support by a bolt. Since the stress may be reduced by the support, the wall thick of the bellows portion may be decreased to sufficiently secure the low electric resistance value. (Yoshihara, H.)

  7. Present status of inertial confinement fusion in Japan

    International Nuclear Information System (INIS)

    Yamanaka, Chiyoe

    1984-01-01

    The Japanese inertial fusion program has made important progress towards implosion fusion process and the technical development required for realizing the breakeven of inertial fusion energy. The key issues for the ICF research are the development of a high power driver, the pertinent pellet design for implosion by a super computer code, and the diagnostics of implosion process with high space and time resolution. The Institute of Laser Engineering (ILE), Osaka University, is the central laboratory for ICF research in Japan. The ILE Osaka has advanced the Kongo Project aiming at the breakeven of inertial fusion since 1980, and as the first phase, the Gekko 12 Nd glass laser of 20 kJ having 12 beams was constructed. The ILE has also the Lekko 8 CO 2 laser and the Reiden 4 light ion beam machine. In the second phase, a 100 kJ class driver will be provided. At the ILE, rare gas halide lasers such as KrF and ArF have been investigated. Laser plasma coupling, the scaling law for implosion pressure, the invention of a new type target ''Cannonball'', and the development of computer codes are described. Also the activities in universities, government laboratories and industrial companies are reported. (Kako, I.)

  8. Characterization of local turbulence in magnetic confinement devices

    International Nuclear Information System (INIS)

    Rajkovic, Milan; Skoric, Milos; Solna, Knut; Antar, Ghassan

    2007-07-01

    A multifractal analysis based on evaluation and interpretation of Large Deviation spectra is applied to plasma edge turbulence data from different devices (MAST and Tore Supra). It is demonstrated that in spite of some universal features there are unique characteristics for each device as well as for different confinement regimes. In the second part of the exposition the issue of estimating the variable power law behavior of spectral densities is addressed. The analysis of this issue is performed using fractional Brownian motion (fBm) as the underlying stochastic model whose parameters are estimated locally in time by wavelet scale spectra. In such a manner information about the inertial range as well as variability of the fBm parameters is obtained giving more information important for understanding edge turbulence and intermittency. (author)

  9. 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.

  10. ITER and research works on magnetic confinement fusion

    International Nuclear Information System (INIS)

    Jacquinot, J.

    2013-01-01

    This article presents the main features of the ITER tokamak and its implications in terms of research concerning plasma instabilities, confinement and materials. It also shows how the financial and technological responsibilities have been shared between the 7 partners (European Union, China, South-Korea, Russia, Japan, India and United-States) and also gives a progress update of the fabrication of the components at the date of September 2012

  11. Spherical fusion plasma-confinement field of Surmac type

    Energy Technology Data Exchange (ETDEWEB)

    Wipf, S.L.

    1981-01-01

    The concept of a Surmac confinement field that can be completely closed is presented. The internal conductor is magnetically suspended inside large corrugations of a superconducting spherical shell structure that carries the return current. Presently available superconductor technology using superfluid helium cooling allows fields above 1.5T throughout the wall region. Such a Surmac has potential for the study of advanced fuel cycles.

  12. New schemes for confinement of fusion products in stellarators

    International Nuclear Information System (INIS)

    Cooper, W.A.; Isaev, M.Yu.; Heyn, M.F.

    2003-01-01

    Improved energetic-particle confinement is found in new stellarator and toroidal mirror field configurations. The possibility of fulfilling the condition of poloidal closure of the contours of the second adiabatic invariant for all reflected particles is studied for stellarators with poloidally closed contours of the magnetic field B on the magnetic surfaces through computational stellarator optimization. It is shown that by adjusting the geometry this is possible in a major fraction of the plasma volume. The most salient characteristic (as compared to previous quasi-iso dynamic configurations) is a magnetic axis whose curvature vanishes in all cross-sections with an extremum of B on the magnetic axis and renders possible a 3D structure of B with unprecedently high collisionless α-particle confinement. Sectionally isometric vacuum magnetic field toroidal mirror traps are analytically constructed with the help of the paraxial (or 'thin tube') approximation. Application of standard computational stellarator tools to this type of ι = 0 stellarator shows excellent alignment of second adiabatic invariant contours and equilibrium surfaces as well as directly calculated collisionless confinement of energetic particles. (author)

  13. Assessment of the critical engineering data needs for the commercialization of magnetic confinement fusion

    International Nuclear Information System (INIS)

    Waganer, L.M.; Zuckerman, D.S.

    1983-01-01

    A survey of twenty-two recent conceptual fusion reactor designs was conducted to ascertain both generic and specific engineering data needs critical for the commercialization of magnetic confinement fusion (MCF). Design experts or advocates for each concept were queried as to the more critical engineering issues and data needs affecting the achievement of commercialization. For each concept, the technical issues were identified and the data needs quantified. Issues and data needs were then ranked based upon the experts' perceptions of the relative importance of each to the concept. The issues encompassed all aspects of the fusion reactor plant design including materials, performance, maintainability, operability, cost, safety and resources

  14. Validating Inertial Confinement Fusion (ICF) predictive capability using perturbed capsules

    Science.gov (United States)

    Schmitt, Mark; Magelssen, Glenn; Tregillis, Ian; Hsu, Scott; Bradley, Paul; Dodd, Evan; Cobble, James; Flippo, Kirk; Offerman, Dustin; Obrey, Kimberly; Wang, Yi-Ming; Watt, Robert; Wilke, Mark; Wysocki, Frederick; Batha, Steven

    2009-11-01

    Achieving ignition on NIF is a monumental step on the path toward utilizing fusion as a controlled energy source. Obtaining robust ignition requires accurate ICF models to predict the degradation of ignition caused by heterogeneities in capsule construction and irradiation. LANL has embarked on a project to induce controlled defects in capsules to validate our ability to predict their effects on fusion burn. These efforts include the validation of feature-driven hydrodynamics and mix in a convergent geometry. This capability is needed to determine the performance of capsules imploded under less-than-optimum conditions on future IFE facilities. LANL's recently initiated Defect Implosion Experiments (DIME) conducted at Rochester's Omega facility are providing input for these efforts. Recent simulation and experimental results will be shown.

  15. Alpha Heating and Burning Plasmas in Inertial Confinement Fusion.

    Science.gov (United States)

    Betti, R; Christopherson, A R; Spears, B K; Nora, R; Bose, A; Howard, J; Woo, K M; Edwards, M J; Sanz, J

    2015-06-26

    Estimating the level of alpha heating and determining the onset of the burning plasma regime is essential to finding the path towards thermonuclear ignition. In a burning plasma, the alpha heating exceeds the external input energy to the plasma. Using a simple model of the implosion, it is shown that a general relation can be derived, connecting the burning plasma regime to the yield enhancement due to alpha heating and to experimentally measurable parameters such as the Lawson ignition parameter. A general alpha-heating curve is found, independent of the target and suitable to assess the performance of all laser fusion experiments whether direct or indirect drive. The onset of the burning plasma regime inside the hot spot of current implosions on the National Ignition Facility requires a fusion yield of about 50 kJ.

  16. 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

  17. Performance of the Cascade inertial-confinement-fusion conceptual reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1984-01-01

    A 4.5-m-radius rotating fusion reactor made of silicon carbide and containing a moving 1-m-thick lithium-ceramic granular blanket can produce 3000 MW/sub t/. The blanket operates at high temperature (>1200 K) leading to gross plant efficiencies of up to 60% using a combined helium-gas turbine (Brayton cycle) with a vapor bottoming cycle

  18. Upgrade of the LLNL Nova laser for inertial confinement fusion

    International Nuclear Information System (INIS)

    Murray, J.R.; Trenholme, J.B.; Hunt, J.T.; Frank, D.N.; Lowdermilk, W.H.; Storm, E.

    1991-01-01

    The Lawrence Livermore National Laboratory has proposed to construct an upgrade to the Nova glass laser facility to give an output energy of 1.5-2 megajoules at 350 nanometers wavelength in a nominally 3--5 nanosecond shaped pulse. The Nova Upgrade will be suitable for driving inertial fusion targets to ignition. This paper reviews the design proposed for the laser. 14 refs., 10 figs., 1 tab

  19. Pre-Amplifier Module for Laser Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Heebner, J E; Bowers, M W

    2008-02-06

    The Pre-Amplifier Modules (PAMs) are the heart of the National Ignition Facility (NIF), providing most of the energy gain for the most energetic laser in the world. Upon completion, NIF will be the only laboratory in which scientists can examine the fusion processes that occur inside stars, supernovae, and exploding nuclear weapons and that may someday serve as a virtually inexhaustible energy source for electricity. Consider that in a fusion power plant 50 cups of water could provide the energy comparable to 2 tons of coal. Of paramount importance for achieving laser-driven fusion ignition with the least energy input is the synchronous and symmetric compression of the target fuel--a condition known as laser power balance. NIF's 48 PAMs thus must provide energy gain in an exquisitely stable and consistent manner. While building one module that meets performance requirements is challenging enough, our design has already enabled the construction and fielding of 48 PAMs that are stable, uniform, and interchangeable. PAM systems are being tested at the University of Rochester's Laboratory for Laser Energetics, and the Atomic Weapons Enterprise of Great Britain has purchased the PAM power system.

  20. Civilian applications of particle-beam-initiated inertial confinement fusion technology

    International Nuclear Information System (INIS)

    Varnado, S.G.; Mitchiner, J.L.

    1977-05-01

    Electrical power generation by controlled fusion may provide a partial solution to the world's long-term energy supply problem. Achievement of a fusion reaction requires the confinement of an extremely hot plasma for a time long enough to allow fuel burnup. Inertial confinement of the plasma may be possible through the use of tightly focused, relativistic electron or ion beams to compress a fuel pellet. The Sandia Particle Beam Fusion program is developing the particle-beam accelerators necessary to achieve fuel ignition. In this report we review the status of the particle-beam fusion technology development program and identify several potential civilian applications for this technology. We describe program objectives, discuss the specific accelerators presently under development, and briefly review the results of beam-focusing and target-irradiation experiments. Then we identify and discuss applications for the beam technology and for the fusion neutrons. The applications are grouped into near-term, intermediate-term, and long-term categories. Near-term applications for the beam technology include electron-beam (e-beam) pumping of gas lasers and several commercial applications. Intermediate-term applications (pellet gain less than 50) include hybrid reactors for electrical power production and fissile fuel breeding, pure fusion reactors for electrical power production, and medical therapy using ion accelerators. In the long term, complex, high-gain pellets may be used in pure fusion reactors

  1. Heavy ion beam propagation through a gas-filled chamber for inertial confinement fusion

    International Nuclear Information System (INIS)

    Barboza, N.O.

    1996-10-01

    The work presented here evaluates the dynamics of a beam of heavy ions propagating through a chamber filled with gas. The motivation for this research stems from the possibility of using heavy ion beams as a driver in inertial confinement fusion reactors for the purpose of generating electricity. Such a study is important in determining the constraints on the beam which limit its focus to the small radius necessary for the ignition of thermonuclear microexplosions which are the source of fusion energy. Nuclear fusion is the process of combining light nuclei to form heavier ones. One possible fusion reaction combines two isotopes of hydrogen, deuterium and tritium, to form an alpha particle and a neutron, with an accompanying release of ∼17.6 MeV of energy. Generating electricity from fusion requires that we create such reactions in an efficient and controlled fashion, and harness the resulting energy. In the inertial confinement fusion (ICF) approach to energy production, a small spherical target, a few millimeters in radius, of deuterium and tritium fuel is compressed so that the density and temperature of the fuel are high enough, ∼200 g/cm 3 and ∼20 keV, that a substantial number of fusion reactions occur; the pellet microexplosion typically releases ∼350 MJ of energy in optimized power plant scenarios

  2. Coil supporting device in a nuclear fusion device

    International Nuclear Information System (INIS)

    Takano, Hirohisa; Sasaki, Katsutoki.

    1976-01-01

    Object: To slide a vacuum vessel in the nuclear fusion device and a coil within the vacuum vessel and to mount the coil within the vacuum vessel in a manner that it may not be moved by an electromagnetic force, thereby preventing stress from being produced in the coil. Structure: A coil supporting plate mounted at upper and lower parts prevents damage to an insulation of the coil, said coil being held in a U-shaped groove, and can be moved integral with the coil by the action of a roller bearing with a plurality of needle-like rollers arranged in parallel. The coil supporting plate has a plurality of projections disposed on the lower surface thereof, and flat springs are placed in the projections one over another so that the spring action exerted in the lower plate causes the coil to be resiliently bias in a direction of an electromagnetic force applied thereto and to support the coil. (Yoshino, Y.)

  3. 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

  4. NRL inertial confinement fusion theory program. 1979 annual report, October 1978 - December 1979

    International Nuclear Information System (INIS)

    1980-01-01

    This is the 1979 annual report of the NRL Inertial Confinement Fusion Theory Program. It covers research performed from October 1978 through December 1979. Research in each of the four current program areas is reported: laser light absorption;fluid dynamics of ablative acceleration; development of computational techniques, and Rayleigh-Taylor stabilization techniques

  5. 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

  6. Modeling hydrodynamic instabilities of double ablation fronts in inertial confinement fusion

    International Nuclear Information System (INIS)

    Yanez, C.; Sanz, J.; Olazabal-Loume, M.; Ibanez, L. F.

    2013-01-01

    A linear Rayleigh-Taylor instability theory of double ablation (DA) fronts is developed for direct-drive inertial confinement fusion. Two approaches are discussed: an analytical discontinuity model for the radiation dominated regime of very steep DA front structure, and a numerical self-consistent model that covers more general hydrodynamic profiles behaviours. Dispersion relation results are compared to 2D simulations. (authors)

  7. Ion distribution in the hot spot of an inertial confinement fusion plasma

    Science.gov (United States)

    Tang, Xianzhu; Guo, Zehua; Berk, Herb

    2012-10-01

    Maximizing the fusion gain of inertial confinement fusion (ICF) for inertial fusion energy (IFE) applications leads to the standard scenario of central hot spot ignition followed by propagating burn wave through the cold/dense assembled fuel. The fact that the hot spot is surrounded by cold but dense fuel layer introduces subtle plasma physics which requires a kinetic description. Here we perform Fokker-Planck calculations and kinetic PIC simulations for an ICF plasma initially in pressure balance but having large temperature gradient over a narrow transition layer. The loss of the fast ion tail from the hot spot, which is important for fusion reactivity, is quantified by Fokker-Planck models. The role of electron energy transport and the ambipolar electric field is investigated via kinetic simulations and the fluid moment models. The net effect on both hot spot ion temperature and the ion tail distribution, and hence the fusion reactivity, is elucidated.

  8. Physical and engineering aspects of a fusion engineering test facility based on mirror confinement

    International Nuclear Information System (INIS)

    Kawabe, T.; Hirayama, S.; Hojo, H.; Kozaki, Y.; Yoshikawa, K.

    1986-01-01

    Controlled fusion research has accomplished great progress in the field of confinement of high-density and high-temperature plasmas and breakeven experiments are expected before the end of the 1980s. Many experiments have been proposed as the next step for fusion research. Among them is the study of ignited plasmas and another is the study of fusion engineering. Some of the important studies in fusion engineering are the integrated test in a fusion reactor environment as well as tests of first-wall materials and of the reactor structures, and test for tritium breeding and blanket modules or submodules. An ideal neutron source for the study of fusion engineering is the deuterium-tritium (D-T) fusion plasma itself. A neutron facility based on a D-T-burning plasma consists of all of the components that a real fusion power reactor would have, so eventually the integrated test for fusion reactor engineering can be done as well as the tests for each engineering component

  9. Indirectly driven, high convergence inertial confinement fusion implosions

    International Nuclear Information System (INIS)

    Cable, M.D.; Hatchett, S.P.; Caird, J.A.; Kilkenny, J.D.; Kornblum, H.N.; Lane, S.M.; Laumann, C.; Lerche, R.A.; Murphy, T.J.; Murray, J.; Nelson, M.B.; Phillion, D.W.; Powell, H.; Ress, D.B.

    1994-01-01

    A series of high convergence indirectly driven implosions has been done with the Nova Laser Fusion facility. These implosions were well characterized by a variety of measurements; computer models are in good agreement. The imploded fuel areal density was measured using a technique based on secondary neutron spectroscopy. At capsule convergences of 24:1, comparable to what is required for the hot spot of ignition scale capsules, these capsules achieved fuel densities of 19 g/cm 3 . Independent measurements of density, burn duration, and ion temperature gave nτθ=1.7±0.9x10 14 keV s/cm 3

  10. Neutronic analysis of fusion tokamak devices by PHITS

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  11. 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)

  12. 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

  13. An innovative accelerator-driven inertial electrostatic confinement device using converging ion beams

    International Nuclear Information System (INIS)

    Bauer, T. H.; Wigeland, R. A.

    1999-01-01

    Fundamental physics issues facing development of fusion power on a small-scale are assessed with emphasis on the idea of Inertial Electrostatic Confinement (IEC). The authors propose a new concept of accelerator-driven IEC fusion, termed Converging Beam Inertial Electrostatic Confinement (CB-IEC). CB-IEC offers a number of innovative features that make it an attractive pathway toward resolving fundamental physics issues and assessing the ultimate viability of the IEC concept for power generation

  14. 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

  15. Potential role of advanced fuels in inertial confinement fusion

    International Nuclear Information System (INIS)

    Miley, G.

    1981-01-01

    The potential importance of developing advanced (non D-T) fuel pellets for inertial confinement is discussed. Reduced radioactivity due to low tritium involvement and less neutron activation, improved blanket flexibility with the removal of tritium breeding requirements, and improved mating of the output energy spectrum with non-electrical applications such as synthetic fuel production could lead to technical advantages and earlier public acceptance. As a possible first step to advanced-fuel pellets, the A-FLINT concept of a D-T core ignited, deuterium pellet is proposed which would offer tritium self-sufficiency. A design is described that uses 0.1-MJ internal energy in a rhoR1--7 gm/cm2'' compressed pellet, giving a tritium breeding ratio of 1--1.0 and an internal pellet gain of 1--700

  16. Magnetic field coil in nuclear fusion device

    International Nuclear Information System (INIS)

    Yamaguchi, Mitsugi; Takano, Hirohisa.

    1975-01-01

    Object: To provide an electrical-insulatively stabilized magnetic field coil in nuclear fusion device, restraining an increase in voltage when plasma current is rapidly changed. Structure: A magnetic field coil comprises coils arranged coaxial with respective vacuum vessels, said coils being wound in positive and reverse polarities so as to form a vertical magnetic field within the plasma. The coils of the positive polarity are arranged along the vacuum vessel inside of an axis vertical in section of the annular plasma and are arranged symmetrically up and down of a horizontal axis. On the other hand, the coils of the reverse polarity are arranged along the vacuum vessel outside of a vertical axis and arranged symmetrically up and down of the horizontal axis. These positive and reverse polarity coils are alternately connected in series, and lead portions of the coils are connected to a power source by means of connecting wires. In this case, lead positions of the coils are arranged in one direction, and the connecting wires are disposed in closely contact relation to offset magnetic fields formed by the connecting wires each other. (Kawakami, Y.)

  17. Inertial-confinement-fusion applications of ion-stopping theory

    International Nuclear Information System (INIS)

    More, R.M.; Lee, Y.T.; Bailey, D.S.

    1982-01-01

    Methods were developed to calculate: (1) the stopping power of a hot plasma target, (2) the charge-state of a fast ion projectile, and (3) the final disposition of the deposited energy. The first issue refers to the stopping power for protons. The proton stopping power is altered in high-density or high-temperature targets, especially at velocities below the stopping peak. The second issue concerns the application of a proton stopping curve to the arbitrary projectile. The third topic is more specialized to inertial fusion and concerns the partition of deposited energy between ion (nuclear motion) degrees of freedom and those corresponding to bound and free electrons. The question here is whether a thermal equilibrium plasma is produced

  18. Liquid jet experiments: relevance to inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1981-01-01

    In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

  19. High convergence, indirect drive inertial confinement fusion experiments at Nova

    International Nuclear Information System (INIS)

    Lerche, R.A.; Cable, M.D.; Hatchett, S.P.; Caird, J.A.; Kilkenny, J.D.; Kornblum, H.N.; Lane, S.M.; Laumann, C.; Murphy, T.J.; Murray, J.; Nelson, M.B.; Phillion, D.W.; Powell, H.; Ress, D.

    1996-01-01

    High convergence, indirect drive implosion experiments have been done at the Nova Laser Facility. The targets were deuterium and deuterium/tritium filled, glass microballoons driven symmetrically by x rays produced in a surrounding uranium hohlraum. Implosions achieved convergence ratios of 24:1 with fuel densities of 19 g/cm 3 ; this is equivalent to the range required for the hot spot of ignition scale capsules. The implosions used a shaped drive and were well characterized by a variety of laser and target measurements. The primary measurement was the fuel density using the secondary neutron technique (neutrons from the reaction 2 H( 3 H,n) 4 He in initially pure deuterium fuel). Laser measurements include power, energy and pointing. Simultaneous measurement of neutron yield, fusion reaction rate, and x-ray images provide additional information about the implosion process. Computer models are in good agreement with measurement results. copyright 1996 American Institute of Physics

  20. Fuel pellets and optical systems for inertially confined fusion

    International Nuclear Information System (INIS)

    Hendricks, C.D.

    1979-01-01

    Current laser-driven ICF targets are complex sets of concentric spherical shells made from a variety of materials including the fuel (e.g., deuterium-tritium), glass, beryllium, gold, polymeric materials, organo-metallics, and several additional organic and inorganic materials depending on the particular experiments to be done. While it is not yet known what the reactor targets will be exactly, there is little reason to believe they will be just simple, low quality glass shells containing DT gas or simple spheres of deuterated polyethylene or other fuel. Consequently, many of the current targets, materials, and fabrication techniques are considered to be applicable to the long range problems of ICF reactor target fabrication. Many current material problems and fabrication techniques are discussed and various quality factors are presented in an attempt to bring an awareness of the possible fusion reactor target materials problems to the scientific and technical community

  1. Role of hydrodynamic instability growth in hot-spot mass gain and fusion performance of inertial confinement fusion implosions

    International Nuclear Information System (INIS)

    Srinivasan, Bhuvana; Tang, Xian-Zhu

    2014-01-01

    In an inertial confinement fusion target, energy loss due to thermal conduction from the hot-spot will inevitably ablate fuel ice into the hot-spot, resulting in a more massive but cooler hot-spot, which negatively impacts fusion yield. Hydrodynamic mix due to Rayleigh-Taylor instability at the gas-ice interface can aggravate the problem via an increased gas-ice interfacial area across which energy transfer from the hot-spot and ice can be enhanced. Here, this mix-enhanced transport effect on hot-spot fusion-performance degradation is quantified using contrasting 1D and 2D hydrodynamic simulations, and its dependence on effective acceleration, Atwood number, and ablation speed is identified

  2. MAGNETIC END CLOSURES FOR PLASMA CONFINING AND HEATING DEVICES

    Science.gov (United States)

    Post, R.F.

    1963-08-20

    More effective magnetic closure field regions for various open-ended containment magnetic fields used in fusion reactor devices are provided by several spaced, coaxially-aligned solenoids utilized to produce a series of nodal field regions of uniform or, preferably, of incrementally increasing intensity separated by lower intensity regions outwardly from the ends of said containment zone. Plasma sources may also be provided to inject plasma into said lower intensity areas to increase plasma density therein. Plasma may then be transported, by plasma diffusion mechanisms provided by the nodal fields, into the containment field. With correlated plasma densities and nodal field spacings approximating the mean free partl cle collision path length in the zones between the nodal fields, optimum closure effectiveness is obtained. (AEC)

  3. Advanced neutral gas diagnostics for magnetic confinement devices

    International Nuclear Information System (INIS)

    Wenzel, U.; Schlisio, G.; Marquardt, M.; Pedersen, T.S.; Kremeyer, T.; Schmitz, O.; Mackie, B.; Maisano-Brown, J.

    2017-01-01

    For the study of particle exhaust in nuclear fusion devices the neutral pressure must be measured in strong magnetic fields. We describe as an example the neutral pressure gauges in the Wendelstein 7-X stellarator. Two types are used: hot cathode ionization gauges (or ASDEX pressure gauges) and Penning gauges. We show some results from the first experimental campaign. The main problems were runtime effects and the failure of some ASDEX pressure gauges. To improve the reliability we integrated a new LaB 6 electron emitter into the ASDEX pressure gauges. In addition, a special Penning gauge without permanent magnets was developed in order to operate Penning gauges near the plasma edge. These new pressure gauges will be used in the upcoming campaign of Wendelstein 7-X.

  4. Hybrid indirect-drive/direct-drive target for inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, Lindsay John

    2018-02-27

    A hybrid indirect-drive/direct drive for inertial confinement fusion utilizing laser beams from a first direction and laser beams from a second direction including a central fusion fuel component; a first portion of a shell surrounding said central fusion fuel component, said first portion of a shell having a first thickness; a second portion of a shell surrounding said fusion fuel component, said second portion of a shell having a second thickness that is greater than said thickness of said first portion of a shell; and a hohlraum containing at least a portion of said fusion fuel component and at least a portion of said first portion of a shell; wherein said hohlraum is in a position relative to said first laser beam and to receive said first laser beam and produce X-rays that are directed to said first portion of a shell and said fusion fuel component; and wherein said fusion fuel component and said second portion of a shell are in a position relative to said second laser beam such that said second portion of a shell and said fusion fuel component receive said second laser beam.

  5. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave rf energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected rf energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected rf energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range delta . The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width delta in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma

  6. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    International Nuclear Information System (INIS)

    Bers, A.

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave rf energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected rf energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected rf energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range delta . The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width delta in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma

  7. Summary of the status of lasers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Holzrichter, J.F.

    1979-01-01

    Laser systems designed for plasma research are operating in many laboratories throughout the world. The laser performance itself has become reasonably consistant from laboratory to laboratory and the focusing properties of the laser beams are understood. The plasma physics data, obtained with these systems, also appears to be reasonably self-consistant and is of great interest for inertial fusion applications. These lasers are commonly providing output powers of 0.5 > 2 TW, and power densities on target of 10 13 -10- 16 W/cm 2 , pulse durations on the order of 100 psec to 3 nsec, wavelengths between 0.5 μ and 10 μ, and focal spot sizes of 100 μ or larger where focal spot edge effects are becoming less dominant. In addition, spurious target responses due to such behavior as pre-pulses, self-focusing, or imprecise focal spot measurement are being observed less often. The technical problems of large multi-beam systems, performing at the 10 TW level, have been overcome and these systems (e.g. the Shiva and Helios lasers) are providing high density compression data with ablative targets. The next step in laser design, the 100 to 300 kJ systems, are under construction and 1 MJ lasers are being contemplated

  8. Computational challenges in magnetic-confinement fusion physics

    Science.gov (United States)

    Fasoli, A.; Brunner, S.; Cooper, W. A.; Graves, J. P.; Ricci, P.; Sauter, O.; Villard, L.

    2016-05-01

    Magnetic-fusion plasmas are complex self-organized systems with an extremely wide range of spatial and temporal scales, from the electron-orbit scales (~10-11 s, ~ 10-5 m) to the diffusion time of electrical current through the plasma (~102 s) and the distance along the magnetic field between two solid surfaces in the region that determines the plasma-wall interactions (~100 m). The description of the individual phenomena and of the nonlinear coupling between them involves a hierarchy of models, which, when applied to realistic configurations, require the most advanced numerical techniques and algorithms and the use of state-of-the-art high-performance computers. The common thread of such models resides in the fact that the plasma components are at the same time sources of electromagnetic fields, via the charge and current densities that they generate, and subject to the action of electromagnetic fields. This leads to a wide variety of plasma modes of oscillations that resonate with the particle or fluid motion and makes the plasma dynamics much richer than that of conventional, neutral fluids.

  9. Use of high current density superconducting coils in fusion devices

    International Nuclear Information System (INIS)

    Green, M.A.

    1979-11-01

    Superconducting magnets will play an important role in fusion research in years to come. The magnets which are currently proposed for fusion research use the concept of cryostability to insure stable operation of the superconducting coils. This paper proposes the use of adiabatically stable high current density superconducting coils in some types of fusion devices. The advantages of this approach are much lower system cold mass, enhanced cryogenic safety, increased access to the plasma and lower cost

  10. Maintenance and waste treatment of tritium existing in and out of the fusion reactor systems. 6. Study of tritium confinement in the facility of a fusion reactor

    International Nuclear Information System (INIS)

    Kobayashi, Kazuhiro

    2000-01-01

    In a future fusion reactor, tritium confinement is one of the key issues for safety. Large amount of tritium (a few grams to a hundred grams level) has been handled safely at the several facilities in the world for fusion research under the multiple confinement concept. In this chapter, the study of tritium behavior in large space such as the building is described using the Caisson Assembly for Tritium Safety (CATS) study such as the final confinement and the present R and D status concerning the tritium confinement is reviewed. (author)

  11. Fusion Engineering Device. Volume 1. Mission and program summary

    International Nuclear Information System (INIS)

    1981-10-01

    This volume presents, in summary form, a recommended approach to implementing the Magnetic Fusion Energy Engineering Act of 1980. These recommendations constitute the findings of the FED Technical Management Board (TMB). The TMB and the affiliated technical managers gave particular scrutiny to elucidating the role of FED in fusion development and to defining the device mission

  12. 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

  13. 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

  14. Fusion plasma theory grant: Task 1, Magnetic confinement fusion plasma theory

    International Nuclear Information System (INIS)

    Callen, J.D.

    1989-07-01

    The research performed under this grant during the current year has concentrated on key tokamak plasma confinement and heating theory issues: further development of neoclassical MHD; development of a new fluid/kinetic hybrid model; energy confinement degradation due to macroscopic phenomena in tokamaks; and some other topics (magnetics analysis, coherent structures, presheath structure). Progress and publications in these areas are briefly summarized in this report. 20 refs

  15. Learning on probabilistic manifolds in massive fusion databases: Application to confinement regime identification

    International Nuclear Information System (INIS)

    Verdoolaege, Geert; Van Oost, Guido

    2012-01-01

    Highlights: ► We present an integrated framework for pattern recognition in fusion data. ► We model measurement uncertainty through an appropriate probability distribution. ► We use the geodesic distance on probabilistic manifolds as a similarity measure. ► We apply the framework to confinement mode classification. ► The classification accuracy benefits from uncertainty information and its geometry. - Abstract: We present an integrated framework for (real-time) pattern recognition in fusion data. The main premise is the inherent probabilistic nature of measurements of plasma quantities. We propose the geodesic distance on probabilistic manifolds as a similarity measure between data points. Substructure induced by data dependencies may further reduce the dimensionality and redundancy of the data set. We present an application to confinement mode classification, showing the distinct advantage obtained by considering the measurement uncertainty and its geometry.

  16. Effects of large-angle Coulomb collisions on inertial confinement fusion plasmas.

    Science.gov (United States)

    Turrell, A E; Sherlock, M; Rose, S J

    2014-06-20

    Large-angle Coulomb collisions affect the rates of energy and momentum exchange in a plasma, and it is expected that their effects will be important in many plasmas of current research interest, including in inertial confinement fusion. Their inclusion is a long-standing problem, and the first fully self-consistent method for calculating their effects is presented. This method is applied to "burn" in the hot fuel in inertial confinement fusion capsules and finds that the yield increases due to an increase in the rate of temperature equilibration between electrons and ions which is not predicted by small-angle collision theories. The equilibration rate increases are 50%-100% for number densities of 10(30)  m(-3) and temperatures around 1 keV.

  17. Index of light ion inertial confinement fusion publications and presentations January 1989 through December 1993

    International Nuclear Information System (INIS)

    Sweeney, M.A.

    1995-11-01

    This report lists publications and presentations that are related to inertial confinement fusion and were authored or coauthored by Sandians in the Pulsed Power Sciences Center from 1989 through 1993. The 661 publications and presentations are categorized into the following general topics: (1) reviews, (2) ion sources, (3) ion diodes, (4) plasma opening switches, (5) ion beam transport, (6) targets and deposition physics, (7) advanced driver and pulsed power technology development, (8) diagnostics, and (9) code development. Research in these areas is arranged by topic in chronological order, with the early efforts under each topic presented first. The work is also categorized alphabetically by first author. A list of acronyms, abbreviations, and definitions of use in understanding light ion inertial confinement fusion research is also included

  18. Recent diagnostic development for inertial confinement fusion research at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, T.J.; Oertel, J.A.; Archuleta, T.N. [and others

    1997-09-01

    Inertial Confinement Fusion (ICF) experiments require sophisticated diagnostics with temporal resolution measured in tens of picoseconds and spatial resolutions measured in microns. The Los Alamos ICF Program is currently supporting a number of diagnostics on the Nova and Triden laser facilities, and is developing new diagnostics for use on the Omega laser facility. New systems and technologies are being developed for use on the National Ignition Facility, which is expected to be operational early in the next decade.

  19. Inertial Confinement Fusion quarterly report, October--December 1994. Volume 5, No. 1

    International Nuclear Information System (INIS)

    1995-01-01

    The ICF quarterly report is published by the Inertial Confinement Fusion Program at the Lawrence Livermore National Laboratory. Topics included in this issue include: system description and initial performance results for beamlet, design and performance of the beamlet amplifiers and optical switch, beamlet pulse-generation and wavefront-control system, large-aperture, high- damage-threshold optics for beamlet, beamlet pulsed power system, beamlet laser diagnostics, and beam propagation and frequency conversion modeling for the beamlet laser

  20. Early history of experimental inertial confinement fusion and diagnostics in China

    International Nuclear Information System (INIS)

    Wang Chuanke; Jiang Shao'en; Ding Yongkun

    2014-01-01

    The early history of China's research on experimental laser inertial confinement fusion (ICF) and diagnostics technology is reviewed. The long and difficult path started from scratch, from learning the basics, looking up the literature and copying experiments, to independent research and development of comprehensive experimental facilities. This article fills a gap in the history of China's ICF experimental and diagnostics research. (authors)

  1. Raman-Brillouin interplay for inertial confinement fusion relevant laser–plasma interaction

    Czech Academy of Sciences Publication Activity Database

    Riconda, C.; Weber, Stefan A.

    2016-01-01

    Roč. 4, Jul (2016), 1-16, č. článku e23. ISSN 2095-4719 R&D Projects: GA MŠk EF15_008/0000162 Grant - others:ELI Beamlines(XE) CZ.02.1.01/0.0/0.0/15_008/0000162 Institutional support: RVO:68378271 Keywords : inertial confinement fusion * kinetic effects * laser- plasma interaction Subject RIV: BL - Plasma and Gas Discharge Physics

  2. Recent diagnostic development for inertial confinement fusion research at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Murphy, T.J.; Oertel, J.A.; Archuleta, T.N.

    1997-01-01

    Inertial Confinement Fusion (ICF) experiments require sophisticated diagnostics with temporal resolution measured in tens of picoseconds and spatial resolutions measured in microns. The Los Alamos ICF Program is currently supporting a number of diagnostics on the Nova and Triden laser facilities, and is developing new diagnostics for use on the Omega laser facility. New systems and technologies are being developed for use on the National Ignition Facility, which is expected to be operational early in the next decade

  3. Linear stability analysis of double ablation fronts in direct-drive inertial confinement fusion

    International Nuclear Information System (INIS)

    Yanez, C.; Sanz, J.; Ibanez, L. F.; Olazabal-Loume, M.

    2011-01-01

    A linear stability theory of double ablation fronts is developed for direct-drive inertial confinement fusion targets. The so-called electron radiative ablation front [S. Fujioka et al., Phys. Rev. Lett. 92, 195001 (2004)] is studied with a self-consistent model. Numerical results are presented as well as an analytical approach for the radiation dominated regime of very steep double ablation front structure. Dispersion relation formula is tackled by means of a sharp boundary model.

  4. Diagnostics considerations for the inertial confinement approach to controlled thermonuclear fusion power production

    International Nuclear Information System (INIS)

    Wood, L.

    1978-01-01

    It is concluded that although the challenges facing diagnosticians working on the inertial confinement approach to controlled fusion are large and varied, the means potentially available to meet them are more than adequate. No new instrumentation fields need be opened; rather, substantial extensions of those already being explored by workers in ICF will suffice. Also, large contributions may be expected from other technological applications thrusts, as well as from the general, currently rapid advance of the entire physical technology base

  5. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from a LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs, 90 Sr, 129 I, 99 Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R = 1.0 to 3.0) requirements. These studies also indicate that masses on the order of 1.0 g at densities of rho greater than or equal to 500.0 g/cm 3 are required for a practical fusion-based fission product transmutation system

  6. Inertial confinement fusion research and development studies. Final report, October 1979-August 1980

    International Nuclear Information System (INIS)

    Bullis, R.; Finkelman, M.; Leng, J.; Luzzi, T.; Ojalvo, I.; Powell, E.; Sedgley, D.

    1980-08-01

    These Inertial Confinement Fusion (ICF) research and development studies were selected for structural, thermal, and vacuum pumping analyses in support of the High Yield Lithium Injection Fusion Energy (HYLIFE) concept development. An additional task provided an outlined program plan for an ICF Engineering Test Facility, using the HYLIFE concept as a model, although the plan is generally applicable to other ICF concepts. The HYLIFE is one promising type of ICF concept which features a falling array of liquid lithium jets. These jets surround the fusion reaction to protect the first structural wall (FSW) of the vacuum chamber by absorbing the fusion energy, and to act as the tritium breeder. The fusion energy source is a deuterium-tritium pellet injected into the chamber every second and driven by laser or heavy ion beams. The studies performed by Grumman have considered the capabilities of specific HYLIFE features to meet life requirements and the requirement to recover to preshot conditions prior to each subsequent shot. The components under investigation were the FSW which restrains the outward motion of the liquid lithium, the nozzle plate which forms the falling jet array, the graphite shield which is in direct top view of the fusion pellet, and the vacuum pumping system. The FSW studies included structural analysis, and definition of an experimental program to validate computer codes describing lithium motion and the resulting impact on the wall

  7. Scaling of the Inertial Electrostatic Confinement (IEC) for near-term thrusters and future fusion propulsion

    International Nuclear Information System (INIS)

    Miley, G.; Bromley, B.; Jurczyk, B.; Stubbers, R.; DeMora, J.; Chacon, L.; Gu, Y.

    1998-01-01

    Inertial Electrostatic Confinement (IEC) is a unique approach to fusion and plasma energy systems that was conceptualized in the 1960s (Hirsch 1967) and has been the focus of recent development in the 1990s (Miley et al. 1995a). In the interests of space power and propulsion systems, conceptual rocket design studies (Bussard and Jameson 1994, Miley et al. 1995b) using the IEC have predicted excellent performance for a variety of space missions, since the power unit avoids the use of magnets and heavy drives resulting in a very high, specific impulse compared to other fusion systems. In their recent survey of prior conceptual design studies of fusion rockets, Williams and Borowski (1997) found that the Bussard IEC conceptual study (the open-quotes QEDclose quotes engine) offered a thrust-to-weight ratio of 10 milli-g close-quote s, a factor of five higher than conventional magnetic confinement concepts and even slightly above anti-proton micro fission/fusion designs. Thus there is considerable motivation to study IEC concepts for eventual space applications. However, the physics feasibility of the IEC still requires experimental demonstration, and an expanded data base is needed to insure that a power unit can in fact be built. copyright 1998 American Institute of Physics

  8. Scaling of the Inertial Electrostatic Confinement (IEC) for near-term thrusters and future fusion propulsion

    International Nuclear Information System (INIS)

    Miley, G.; Bromley, B.; Jurczyk, B.; Stubbers, R.; DeMora, J.; Chacon, L.; Gu, Y.

    1998-01-01

    Inertial Electrostatic Confinement (IEC) is a unique approach to fusion and plasma energy systems that was conceptualized in the 1960s (Hirsch 1967) and has been the focus of recent development in the 1990s (Miley et al. 1995a). In the interests of space power and propulsion systems, conceptual rocket design studies (Bussard and Jameson 1994, Miley et al. 1995b) using the IEC have predicted excellent performance for a variety of space missions, since the power unit avoids the use of magnets and heavy drives resulting in a very high, specific impulse compared to other fusion systems. In their recent survey of prior conceptual design studies of fusion rockets, Williams and Borowski (1997) found that the Bussard IEC conceptual study (the ''QED'' engine) offered a thrust-to-weight ratio of 10 milli-g's, a factor of five higher than conventional magnetic confinement concepts and even slightly above anti-proton micro fission/fusion designs. Thus there is considerable motivation to study IEC concepts for eventual space applications. However, the physics feasibility of the IEC still requires experimental demonstration, and an expanded data base is needed to insure that a power unit can in fact be built

  9. First downscattered neutron images from Inertial Confinement Fusion experiments at the National Ignition Facility

    Directory of Open Access Journals (Sweden)

    Guler Nevzat

    2013-11-01

    Full Text Available Inertial Confinement Fusion experiments at the National Ignition Facility (NIF are designed to understand and test the basic principles of self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT filled cryogenic plastic (CH capsules. The experimental campaign is ongoing to tune the implosions and characterize the burning plasma conditions. Nuclear diagnostics play an important role in measuring the characteristics of these burning plasmas, providing feedback to improve the implosion dynamics. The Neutron Imaging (NI diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by collecting images at two different energy bands for primary (13–15 MeV and downscattered (10–12 MeV neutrons. From these distributions, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. The first downscattered neutron images from imploding ICF capsules are shown in this paper.

  10. First downscattered neutron images from Inertial Confinement Fusion experiments at the National Ignition Facility

    Science.gov (United States)

    Guler, Nevzat; Aragonez, Robert J.; Archuleta, Thomas N.; Batha, Steven H.; Clark, David D.; Clark, Deborah J.; Danly, Chris R.; Day, Robert D.; Fatherley, Valerie E.; Finch, Joshua P.; Gallegos, Robert A.; Garcia, Felix P.; Grim, Gary; Hsu, Albert H.; Jaramillo, Steven A.; Loomis, Eric N.; Mares, Danielle; Martinson, Drew D.; Merrill, Frank E.; Morgan, George L.; Munson, Carter; Murphy, Thomas J.; Oertel, John A.; Polk, Paul J.; Schmidt, Derek W.; Tregillis, Ian L.; Valdez, Adelaida C.; Volegov, Petr L.; Wang, Tai-Sen F.; Wilde, Carl H.; Wilke, Mark D.; Wilson, Douglas C.; Atkinson, Dennis P.; Bower, Dan E.; Drury, Owen B.; Dzenitis, John M.; Felker, Brian; Fittinghoff, David N.; Frank, Matthias; Liddick, Sean N.; Moran, Michael J.; Roberson, George P.; Weiss, Paul; Buckles, Robert A.; Cradick, Jerry R.; Kaufman, Morris I.; Lutz, Steve S.; Malone, Robert M.; Traille, Albert

    2013-11-01

    Inertial Confinement Fusion experiments at the National Ignition Facility (NIF) are designed to understand and test the basic principles of self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT) filled cryogenic plastic (CH) capsules. The experimental campaign is ongoing to tune the implosions and characterize the burning plasma conditions. Nuclear diagnostics play an important role in measuring the characteristics of these burning plasmas, providing feedback to improve the implosion dynamics. The Neutron Imaging (NI) diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by collecting images at two different energy bands for primary (13-15 MeV) and downscattered (10-12 MeV) neutrons. From these distributions, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. The first downscattered neutron images from imploding ICF capsules are shown in this paper.

  11. Diagnostic technique for measuring fusion reaction rate for inertial confinement fusion experiments at Shen Guang-III prototype laser facility

    International Nuclear Information System (INIS)

    Wang Feng; Peng Xiao-Shi; Liu Shen-Ye; Xu Tao; Kang Dong-Guo

    2013-01-01

    A study is conducted using a two-dimensional simulation program (Lared-s) with the goal of developing a technique to evaluate the effect of Rayleigh-Taylor growth in a neutron fusion reaction region. Two peaks of fusion reaction rate are simulated by using a two-dimensional simulation program (Lared-s) and confirmed by the experimental results. A neutron temporal diagnostic (NTD) system is developed with a high temporal resolution of ∼ 30 ps at the Shen Guang-III (SG-III) prototype laser facility in China, to measure the fusion reaction rate history. With the shape of neutron reaction rate curve and the spherical harmonic function in this paper, the degree of Rayleigh-Taylor growth and the main source of the neutron yield in our experiment can be estimated qualitatively. This technique, including the diagnostic system and the simulation program, may provide important information for obtaining a higher neutron yield in implosion experiments of inertial confinement fusion

  12. Beam limiter for thermonuclear fusion devices

    International Nuclear Information System (INIS)

    Kaminsky, M.S.

    1977-01-01

    The invention pertains to a beam limiter to prevent collisions between a plasma and the inner surface of a hollow body in which the plasma is confined. The patent claims pertain to suitable geometrical shapes of the beam limiter. (GG) [de

  13. Particle and impurity control in toroidal fusion devices

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1986-01-01

    A review of working particle and impurity control techniques used in and proposed for magnetic fusion devices is presented. The requirements of both present-day machines and envisaged fusion reactors are considered. The various techniques which have been proposed are characterized by whether they affect sources, sinks, or fluxes; in many cases a particular method or device can appear in more than one category. Examples are drawn from published results. The solutions proposed for the large devices which will be operating during the next 5 years are discussed

  14. HYPERFUSE: a hypervelocity inertial confinement system for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs, 90 Sr, 129 I, 99 Tc, etc. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n,2n), (n,α), (n,γ), etc.) that convert the long-lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product. The transmutation parametric studies conclude that the design of the hypervelocity projectiles should emphasize the achievement of high densities in the transmutation regions (greater than the DT fusion fuel density), as well as the DT ignition and burn criterion (rho R=1.0 to 3.0) requirements

  15. Numerical simulations of inertial confinement fusion hohlraum with LARED-integration code

    International Nuclear Information System (INIS)

    Li Jinghong; Li Shuanggui; Zhai Chuanlei

    2011-01-01

    In the target design of the Inertial Confinement Fusion (ICF) program, it is common practice to apply radiation hydrodynamics code to study the key physical processes happened in ICF process, such as hohlraum physics, radiation drive symmetry, capsule implosion physics in the radiation-drive approach of ICF. Recently, many efforts have been done to develop our 2D integrated simulation capability of laser fusion with a variety of optional physical models and numerical methods. In order to effectively integrate the existing codes and to facilitate the development of new codes, we are developing an object-oriented structured-mesh parallel code-supporting infrastructure, called JASMIN. Based on two-dimensional three-temperature hohlraum physics code LARED-H and two-dimensional multi-group radiative transfer code LARED-R, we develop a new generation two-dimensional laser fusion code under the JASMIN infrastructure, which enable us to simulate the whole process of laser fusion from the laser beams' entrance into the hohlraum to the end of implosion. In this paper, we will give a brief description of our new-generation two-dimensional laser fusion code, named LARED-Integration, especially in its physical models, and present some simulation results of holhraum. (author)

  16. Three equipment concepts for the Fusion Engineering Device

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Masson, L.S.; Watts, K.D.; Grant, N.R.; Kuban, D.P.

    1982-01-01

    Maintenance equipment which is needed to remotely handle fusion device components is being conceptually developed for the Fusion Engineering Design Center. This will test the assumption that these equipment needs can be satisfied by present technology. In addition, the development of equipment conceptual designs will allow for cost estimates which have a much higher degree of certainty. Accurate equipment costs will be useful for assessments which trade off gains in availability as a function of increased investments in maintenance equipment

  17. Database for fusion devices and associated fuel systems

    International Nuclear Information System (INIS)

    Woolgar, P.W.

    1983-03-01

    A computerized database storage and retrieval system has been set up for fusion devices and the associated fusion fuel systems which should be a useful tool for the CFFTP program and other users. The features of the Wang 'Alliance' system are discussed for this application, as well as some of the limitations of the system. Recommendations are made on the operation, upkeep and further development that should take place to implement and maintain the system

  18. DD fusion neutron production at UW-Madison using IEC devices

    Science.gov (United States)

    Fancher, Aaron; Michalak, Matt; Kulcinski, Gerald; Santarius, John; Bonomo, Richard

    2017-10-01

    An inertial electrostatic confinement (IEC) device using spherical, gridded electrodes at high voltage accelerates deuterium ions, allowing for neutrons to be produced within the device from DD fusion reactions. The effects of the device cathode voltage (30-170 kV), current (30-100 mA), and pressure (0.15-1.25 mTorr) on the neutron production rate have been measured. New high voltage capabilities have resulted in the achievement of a steady state neutron production rate of 3.3x108 n/s at 175 kV, 100 mA, and 1.0 mTorr of deuterium. Applications of IEC devices include the production of DD neutrons to detect chemical explosives and special nuclear materials using active interrogation methods. Research supported by US Dept. of Homeland Security Grant 2015-DN-077-AR1095 and the Grainger Foundation.

  19. RF-heating and plasma confinement studies in HANBIT mirror device

    International Nuclear Information System (INIS)

    Kwon, M.; Bak, J.G.; Choh, K.K.

    2003-01-01

    HANBIT is a magnetic mirror confinement device. Recently, with almost finishing the first campaign for the basic system development, it started the second campaign for the high-temperature plasma confinement physics study in mirror configuration. Here, we introduce briefly the HANBIT device and report initial physics experiments results on RF-plasma heating and confinement in the simple mirror configuration. It appears that the discharge characteristics of HANBIT are quite different from those in other mirror devices, and an explanation is presented to clarify the difference. (author)

  20. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

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

  1. Mode 1 drive asymmetry in inertial confinement fusion implosions on the National Ignition Facility

    International Nuclear Information System (INIS)

    Spears, Brian K.; Edwards, M. J.; Hatchett, S.; Kritcher, A.; Lindl, J.; Munro, D.; Patel, P.; Robey, H. F.; Town, R. P. J.; Kilkenny, J.; Knauer, J.

    2014-01-01

    Mode 1 radiation drive asymmetry (pole-to-pole imbalance) at significant levels can have a large impact on inertial confinement fusion implosions at the National Ignition Facility (NIF). This asymmetry distorts the cold confining shell and drives a high-speed jet through the hot spot. The perturbed hot spot shows increased residual kinetic energy and reduced internal energy, and it achieves reduced pressure and neutron yield. The altered implosion physics manifests itself in observable diagnostic signatures, especially the neutron spectrum which can be used to measure the neutron-weighted flow velocity, apparent ion temperature, and neutron downscattering. Numerical simulations of implosions with mode 1 asymmetry show that the resultant simulated diagnostic signatures are moved toward the values observed in many NIF experiments. The diagnostic output can also be used to build a set of integrated implosion performance metrics. The metrics indicate that P 1 has a significant impact on implosion performance and must be carefully controlled in NIF implosions

  2. Radiochemical determination of Inertial Confinement Fusion capsule compression at the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Shaughnessy, D. A., E-mail: shaughnessy2@llnl.gov; Moody, K. J.; Gharibyan, N.; Grant, P. M.; Gostic, J. M.; Torretto, P. C.; Wooddy, P. T.; Bandong, B. B.; Cerjan, C. J.; Hagmann, C. A.; Caggiano, J. A.; Yeamans, C. B.; Bernstein, L. A.; Schneider, D. H. G.; Henry, E. A.; Fortner, R. J. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94551 (United States); Despotopulos, J. D. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94551 (United States); Radiochemistry Program, University of Nevada Las Vegas, Las Vegas, Nevada 89154 (United States)

    2014-06-15

    We describe a radiochemical measurement of the ratio of isotope concentrations produced in a gold hohlraum surrounding an Inertial Confinement Fusion capsule at the National Ignition Facility (NIF). We relate the ratio of the concentrations of (n,γ) and (n,2n) products in the gold hohlraum matrix to the down-scatter of neutrons in the compressed fuel and, consequently, to the fuel's areal density. The observed ratio of the concentrations of {sup 198m+g}Au and {sup 196g}Au is a performance signature of ablator areal density and the fuel assembly confinement time. We identify the measurement of nuclear cross sections of astrophysical importance as a potential application of the neutrons generated at the NIF.

  3. Assessment of ion kinetic effects in shock-driven inertial confinement fusion implosions using fusion burn imaging

    International Nuclear Information System (INIS)

    Rosenberg, M. J.; Séguin, F. H.; Rinderknecht, H. G.; Zylstra, A. B.; Li, C. K.; Sio, H.; Johnson, M. Gatu; Frenje, J. A.; Petrasso, R. D.; Amendt, P. A.; Wilks, S. C.; Pino, J.; Atzeni, S.; Hoffman, N. M.; Kagan, G.; Molvig, K.; Glebov, V. Yu.; Stoeckl, C.; Seka, W.; Marshall, F. J.

    2015-01-01

    The significance and nature of ion kinetic effects in D 3 He-filled, shock-driven inertial confinement fusion implosions are assessed through measurements of fusion burn profiles. Over this series of experiments, the ratio of ion-ion mean free path to minimum shell radius (the Knudsen number, N K ) was varied from 0.3 to 9 in order to probe hydrodynamic-like to strongly kinetic plasma conditions; as the Knudsen number increased, hydrodynamic models increasingly failed to match measured yields, while an empirically-tuned, first-step model of ion kinetic effects better captured the observed yield trends [Rosenberg et al., Phys. Rev. Lett. 112, 185001 (2014)]. Here, spatially resolved measurements of the fusion burn are used to examine kinetic ion transport effects in greater detail, adding an additional dimension of understanding that goes beyond zero-dimensional integrated quantities to one-dimensional profiles. In agreement with the previous findings, a comparison of measured and simulated burn profiles shows that models including ion transport effects are able to better match the experimental results. In implosions characterized by large Knudsen numbers (N K  ∼ 3), the fusion burn profiles predicted by hydrodynamics simulations that exclude ion mean free path effects are peaked far from the origin, in stark disagreement with the experimentally observed profiles, which are centrally peaked. In contrast, a hydrodynamics simulation that includes a model of ion diffusion is able to qualitatively match the measured profile shapes. Therefore, ion diffusion or diffusion-like processes are identified as a plausible explanation of the observed trends, though further refinement of the models is needed for a more complete and quantitative understanding of ion kinetic effects

  4. Assessment of ion kinetic effects in shock-driven inertial confinement fusion implosions using fusion burn imaging

    Energy Technology Data Exchange (ETDEWEB)

    Rosenberg, M. J., E-mail: mros@lle.rochester.edu; Séguin, F. H.; Rinderknecht, H. G.; Zylstra, A. B.; Li, C. K.; Sio, H.; Johnson, M. Gatu; Frenje, J. A.; Petrasso, R. D. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Amendt, P. A.; Wilks, S. C.; Pino, J. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Atzeni, S. [Dipartimento SBAI, Università di Roma “La Sapienza” and CNISM, Via A. Scarpa 14-16, I-00161 Roma (Italy); Hoffman, N. M.; Kagan, G.; Molvig, K. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Glebov, V. Yu.; Stoeckl, C.; Seka, W.; Marshall, F. J. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States); and others

    2015-06-15

    The significance and nature of ion kinetic effects in D{sup 3}He-filled, shock-driven inertial confinement fusion implosions are assessed through measurements of fusion burn profiles. Over this series of experiments, the ratio of ion-ion mean free path to minimum shell radius (the Knudsen number, N{sub K}) was varied from 0.3 to 9 in order to probe hydrodynamic-like to strongly kinetic plasma conditions; as the Knudsen number increased, hydrodynamic models increasingly failed to match measured yields, while an empirically-tuned, first-step model of ion kinetic effects better captured the observed yield trends [Rosenberg et al., Phys. Rev. Lett. 112, 185001 (2014)]. Here, spatially resolved measurements of the fusion burn are used to examine kinetic ion transport effects in greater detail, adding an additional dimension of understanding that goes beyond zero-dimensional integrated quantities to one-dimensional profiles. In agreement with the previous findings, a comparison of measured and simulated burn profiles shows that models including ion transport effects are able to better match the experimental results. In implosions characterized by large Knudsen numbers (N{sub K} ∼ 3), the fusion burn profiles predicted by hydrodynamics simulations that exclude ion mean free path effects are peaked far from the origin, in stark disagreement with the experimentally observed profiles, which are centrally peaked. In contrast, a hydrodynamics simulation that includes a model of ion diffusion is able to qualitatively match the measured profile shapes. Therefore, ion diffusion or diffusion-like processes are identified as a plausible explanation of the observed trends, though further refinement of the models is needed for a more complete and quantitative understanding of ion kinetic effects.

  5. A scheme for reducing deceleration-phase Rayleigh-Taylor growth in inertial confinement fusion implosions

    Science.gov (United States)

    Wang, L. F.; Ye, W. H.; Wu, J. F.; Liu, Jie; Zhang, W. Y.; He, X. T.

    2016-05-01

    It is demonstrated that the growth of acceleration-phase instabilities in inertial confinement fusion implosions can be controlled, especially in the high-foot implosions [O. A. Hurricane et al., Phys. Plasmas 21, 056314 (2014)] on the National Ignition Facility. However, the excessive growth of the deceleration-phase instabilities can still destroy the hot spot ignition. A scheme is proposed to retard the deceleration-phase Rayleigh-Taylor instability growth by shock collision near the waist of the inner shell surface. Two-dimensional radiation hydrodynamic simulations confirm the improved deceleration-phase hot spot stability properties without sacrificing the fuel compression.

  6. Improving hot-spot pressure for ignition in high-adiabat Inertial Confinement Fusion implosion

    OpenAIRE

    Kang, Dongguo; Zhu, Shaoping; Pei, Wenbing; Zou, Shiyang; Zheng, Wudi; Gu, Jianfa; Dai, Zhensheng

    2017-01-01

    A novel capsule target design to improve the hot-spot pressure in the high-adiabat implosion for inertial confinement fusion is proposed, where a layer of comparatively high-density material is used as a pusher between the fuel and the ablator. This design is based on our theoretical finding of the stagnation scaling laws, which indicates that the hot spot pressure can be improved by increasing the kinetic energy density $\\rho_d V_{imp}^2/2$ ($\\rho_d$ is the shell density when the maximum she...

  7. Response of a lithium fall to an inertially confined fusion microexplosion

    International Nuclear Information System (INIS)

    Hovingh, J.; Blink, J.; Glenn, L.

    1978-01-01

    One of the most difficult technology problems in an inertially confined fusion reactor is the survival of the structure from the repeated stresses caused by the microexplosion products. To mitigate the damage from the microexplosion products, a thick lithium fall can be circulated in front of the structure. This fall will absorb the short-ranged products and moderate and attenuate the neutrons. This paper discusses the response of the fall to the microexplosion products, and estimates the resulting loading and stresses in the first structural wall

  8. Advances in HYDRA and its applications to simulations of inertial confinement fusion targets

    Directory of Open Access Journals (Sweden)

    Marinak M.M.

    2013-11-01

    Full Text Available A new set of capabilities has been implemented in the HYDRA 2D/3D multiphysics inertial confinement fusion simulation code. These include a Monte Carlo particle transport library. It models transport of neutrons, gamma rays and light ions, as well as products they generate from nuclear and coulomb collisions. It allows accurate simulations of nuclear diagnostic signatures from capsule implosions. We apply it to here in a 3D simulation of a National Ignition Facility (NIF ignition capsule which models the full capsule solid angle. This simulation contains a severely rough ablator perturbation and provides diagnostics signatures of capsule failure due to excessive instability growth.

  9. Multiple-beam laser–plasma interactions in inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Myatt, J. F., E-mail: jmya@lle.rochester.edu; Zhang, J.; Maximov, A. V. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Department of Mechanical Engineering, University of Rochester, Rochester, New York 14627 (United States); Short, R. W.; Seka, W.; Edgell, D. H.; Michel, D. T.; Igumenshchev, I. V. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Froula, D. H. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Department of Physics and Astronomy, University of Rochester, Rochester, New York 14627-0171 (United States); Hinkel, D. E.; Michel, P.; Moody, J. D. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551-0808 (United States)

    2014-05-15

    The experimental evidence for multiple-beam laser-plasma instabilities of relevance to laser driven inertial confinement fusion at the ignition scale is reviewed, in both the indirect and direct-drive approaches. The instabilities described are cross-beam energy transfer (in both indirectly driven targets on the NIF and in direct-drive targets), multiple-beam stimulated Raman scattering (for indirect-drive), and multiple-beam two-plasmon decay instability (in direct drive). Advances in theoretical understanding and in the numerical modeling of these multiple beam instabilities are presented.

  10. A scheme for reducing deceleration-phase Rayleigh–Taylor growth in inertial confinement fusion implosions

    Energy Technology Data Exchange (ETDEWEB)

    Wang, L. F., E-mail: wang-lifeng@iapcm.ac.cn; Ye, W. H.; Liu, Jie [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China); HEDPS, Center for Applied Physics and Technology, Peking University, Beijing 100871 (China); Center for Fusion Energy Science and Technology, Chinese Academy of Engineering Physics, Beijing 100088 (China); Wu, J. F. [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China); Center for Fusion Energy Science and Technology, Chinese Academy of Engineering Physics, Beijing 100088 (China); Zhang, W. Y. [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China); He, X. T. [Institute of Applied Physics and Computational Mathematics, Beijing 100094 (China); HEDPS, Center for Applied Physics and Technology, Peking University, Beijing 100871 (China)

    2016-05-15

    It is demonstrated that the growth of acceleration-phase instabilities in inertial confinement fusion implosions can be controlled, especially in the high-foot implosions [O. A. Hurricane et al., Phys. Plasmas 21, 056314 (2014)] on the National Ignition Facility. However, the excessive growth of the deceleration-phase instabilities can still destroy the hot spot ignition. A scheme is proposed to retard the deceleration-phase Rayleigh–Taylor instability growth by shock collision near the waist of the inner shell surface. Two-dimensional radiation hydrodynamic simulations confirm the improved deceleration-phase hot spot stability properties without sacrificing the fuel compression.

  11. Prolate-Spheroid ('Rugby-Shaped') Hohlraum for Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Vandenboomgaerde, M.; Bastian, J.; Casner, A.; Galmiche, D.; Jadaud, J.-P.; Laffite, S.; Liberatore, S.; Malinie, G.; Philippe, F.

    2007-01-01

    A novel rugby-ball shaped hohlraum is designed in the context of the indirect-drive scheme of inertial-confinement fusion (ICF). Experiments were performed on the OMEGA laser and are the first use of rugby hohlraums for ICF studies. Analysis of experimental data shows that the hohlraum energetics is well understood. We show that the rugby-ball shape exhibits advantages over cylinder, in terms of temperature and of symmetry control of the capsule implosion. Simulations indicate that rugby hohlraum driven targets may be candidates for ignition in a context of early Laser MegaJoule experiments with reduced laser energy

  12. Simulations of mixing in Inertial Confinement Fusion with front tracking and sub-grid scale models

    Science.gov (United States)

    Rana, Verinder; Lim, Hyunkyung; Melvin, Jeremy; Cheng, Baolian; Glimm, James; Sharp, David

    2015-11-01

    We present two related results. The first discusses the Richtmyer-Meshkov (RMI) and Rayleigh-Taylor instabilities (RTI) and their evolution in Inertial Confinement Fusion simulations. We show the evolution of the RMI to the late time RTI under transport effects and tracking. The role of the sub-grid scales helps capture the interaction of turbulence with diffusive processes. The second assesses the effects of concentration on the physics model and examines the mixing properties in the low Reynolds number hot spot. We discuss the effect of concentration on the Schmidt number. The simulation results are produced using the University of Chicago code FLASH and Stony Brook University's front tracking algorithm.

  13. 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)

  14. Fiber scintillator/streak camera detector for burn history measurement in inertial confinement fusion experiment

    International Nuclear Information System (INIS)

    Miyanaga, N.; Ohba, N.; Fujimoto, K.

    1997-01-01

    To measure the burn history in an inertial confinement fusion experiment, we have developed a new neutron detector based on plastic scintillation fibers. Twenty-five fiber scintillators were arranged in a geometry compensation configuration by which the time-of-flight difference of the neutrons is compensated by the transit time difference of light passing through the fibers. Each fiber scintillator is spliced individually to an ultraviolet optical fiber that is coupled to a streak camera. We have demonstrated a significant improvement of sensitivity compared with the usual bulk scintillator coupled to a bundle of the same ultraviolet fibers. copyright 1997 American Institute of Physics

  15. Standard method for economic analyses of inertial confinement fusion power plants

    International Nuclear Information System (INIS)

    Meier, W.R.

    1986-01-01

    A standard method for calculating the total capital cost and the cost of electricity for a typical inertial confinement fusion electric power plant has been developed. A standard code of accounts at the two-digit level is given for the factors making up the total capital cost of the power plant. Equations are given for calculating the indirect capital costs, the project contingency, and the time-related costs. Expressions for calculating the fixed charge rate, which is necessary to determine the cost of electricity, are also described. Default parameters are given to define a reference case for comparative economic analyses

  16. Coil for a nuclear fusion device

    International Nuclear Information System (INIS)

    Kadotani, Kenzo.

    1975-01-01

    Object: To provide a thin nuclear fusion coil having good thermal insulation and insulating properties in which mica and glass materials are wound round conductors subjected to varnish treatment and hardened, which is then sealed into a metallic case along with negative gases of more than two atmospheric pressures. Structure: A plurality of conductors impregnated with varnish are hardened by a rare insulating layer, after which it is coated with a layer of mica not impregnated with varnish and a layer of glass substance and is then received into a metallic case and filled under pressure with negative gases at a pressure more than two atmospheric pressures. (Kamimura, M.)

  17. Overview of the Los Alamos National Laboratory Inertial Confinement Fusion Program

    International Nuclear Information System (INIS)

    Harris, D.B.

    1991-01-01

    The Los Alamos Inertial Confinement Fusion (ICF) Program is focused on preparing for a National Ignition Facility. Target physics research is addressing specific issues identified for the Ignition Facility target, and materials experts are developing target fabrication techniques necessary for the advanced targets. We are also working with Lawrence Livermore National Laboratory on the design of the National Ignition Facility target chamber. Los Alamos is also continuing to develop the KrF laser-fusion driver for ICF. We are modifying the Aurora laser to higher intensity and shorter pulses and are working with the Naval Research Laboratory on the development of the Nike KrF laser. 9 refs., 1 fig., 2 tabs

  18. Inertial confinement fusion quarterly report, April--June 1994. Volume 4, Number 3

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, M.J. [ed.

    1994-06-01

    This issue of the ICF Quarterly contains six articles covering a wide range of activities within the Inertial Confinement Fusion (ICF) Program. It concentrates on target design; theoretical spectral analysis of ICF capsule surfaces; laser fusion experimental methods; and an alternative ICF design, based on ultrafast, ultrapowerful lasers. A key issue for the success of the ICF process is the hydrodynamic stability of the imploding capsule. There are two primary sources of instability growth in the ICF process: (1) asymmetries in the x-ray flux that drive the compression lead to asymmetric in the imploding surface; (2) imperfections on the capsule surface can grow into large perturbations, degrading the capsule performance. In recent years, a great deal of effort, both experimentally and theoretically, has been spent to enhance the Program`s ability to measure, model, and minimize instability growth during an implosion. Four the articles in this issue discuss this subject.

  19. Manufactured solutions for the three-dimensional Euler equations with relevance to Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Waltz, J.; Canfield, T.R.; Morgan, N.R.; Risinger, L.D.; Wohlbier, J.G.

    2014-01-01

    We present a set of manufactured solutions for the three-dimensional (3D) Euler equations. The purpose of these solutions is to allow for code verification against true 3D flows with physical relevance, as opposed to 3D simulations of lower-dimensional problems or manufactured solutions that lack physical relevance. Of particular interest are solutions with relevance to Inertial Confinement Fusion (ICF) capsules. While ICF capsules are designed for spherical symmetry, they are hypothesized to become highly 3D at late time due to phenomena such as Rayleigh–Taylor instability, drive asymmetry, and vortex decay. ICF capsules also involve highly nonlinear coupling between the fluid dynamics and other physics, such as radiation transport and thermonuclear fusion. The manufactured solutions we present are specifically designed to test the terms and couplings in the Euler equations that are relevant to these phenomena. Example numerical results generated with a 3D Finite Element hydrodynamics code are presented, including mesh convergence studies

  20. Acceleration systems for heavy-ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    Faltens, A.; Judd, D.L.; Keefe, D.

    1977-01-01

    Heavy-ion beam pulse parameters needed to achieve useful electric power generation through inertial confinement fusion have been set forth. For successful ignition of a high-gain D-T target a few magajoules of energy per pulse, delivered at a peak power of several hundred terawatts, are needed; it must be deposited with an energy density of 20 to 30 magajoules per gram of the target material on which it impinges. Additional requirements must be met if this form of fusion is to be used for practical power generation; for example, the igniter system for a 1 GWe power plant should have a repetition rate in the neighborhood of 1 to 10 Hz, an overall electrical conversion efficiency from mains to beam of greater than 10%, and high availability. At present under discussion are the needs for a Heavy-Ion Demonstration Experiment (HIDE); an example set of parameters is given for comparison with those for a power plant

  1. Progress in direct-drive inertial confinement fusion research at the Laboratory for Laser Energetics

    International Nuclear Information System (INIS)

    McCrory, R.L.

    2002-01-01

    Significant theoretical and experimental progress towards the validation of direct-drive inertial confinement fusion (ICF) has been recently made at the Laboratory for Laser Energetics (LLE). Direct-drive ICF offers the potential for high-gain implosions and is a leading candidate for an inertial fusion energy power plant. LLE's base-line direct-drive ignition design for NIF is an 'all-DT' design that has a 1-D gain of ∼45. Recent calculations show that targets composed of foam shells, wicked with DT, can potentially achieve 1-D gains of ∼100. LLE experiments are conducted on the OMEGA 60-beam, 30-kJ, UV laser system. Beam smoothing of OMEGA includes 1-THz, 2-D SSD and polarization smoothing. Cryogenic D2 and plastic shell (warm) spherical targets and a comprehensive suite of x-ray, nuclear, charged particle and optical diagnostics are used in these experiments. Future experiments will use cryogenic DT targets. (author)

  2. Recent progress in inertial confinement fusion at the Lawrence Livermore Laboratory

    International Nuclear Information System (INIS)

    Ahlstrom, H.G.; Manes, K.R.

    1979-01-01

    The Shiva and Argus laser systems at Livermore have been developed to study the physics of inertial confinement fusion. Both laser system designs are predicated on the use of large aperture Nd-glass disk amplifiers and high power spatial filters. During the past year we have irradiated DT filled microshell targets with and without polymer coatings. Recently new instruments have been developed to investigate implosion dynamics and to determine the maximum fuel density achieved by these imploded fusion pellets. A series of target irradiations with thin wall microshells at 15 to 20 TW, exploding pusher designs, resulted in a maximum neutron yield of 3 x 10 10 . Polymer coated microshells designed for high compression were subjected to 4 kJ for 0.2 ns and reached fuel densities of 2.0 to 3.0 gm/cm 3 . Results of these and other recent experiments will be reviewed

  3. The US inertial confinement fusion (ICF) ignition programme and the inertial fusion energy (IFE) programme

    Science.gov (United States)

    Lindl, J. D.; Hammel, B. A.; Logan, B. Grant; Meyerhofer, David D.; Payne, S. A.; Sethian, John D.

    2003-12-01

    There has been rapid progress in inertial fusion in the past few years. This progress spans the construction of ignition facilities, a wide range of target concepts and the pursuit of integrated programmes to develop fusion energy using lasers, ion beams and z-pinches. Two ignition facilities are under construction, the national ignition facility (NIF) in the United States and the laser megajoule (LMJ) in France, and both projects are progressing towards an initial experimental capability. The laser integration line prototype beamline for LMJ and the first four beams of NIF will be available for experiments in 2003. The full 192 beam capability of NIF will be available in 2009 and ignition experiments are expected to begin shortly after that time. There is steady progress in target science and target fabrication in preparation for indirect-drive ignition experiments on NIF. Advanced target designs may lead to 5 10 times more yield than initial target designs. There has also been excellent progress on the science of ion beam and z-pinch-driven indirect-drive targets. Excellent progress on direct-drive targets has been obtained on the Omega laser at the University of Rochester. This includes improved performance of targets with a pulse shape predicted to result in reduced hydrodynamic instability. Rochester has also obtained encouraging results from initial cryogenic implosions. There is widespread interest in the science of fast ignition because of its potential for achieving higher target gain with lower driver energy and relaxed target fabrication requirements. Researchers from Osaka have achieved outstanding implosion and heating results from the Gekko XII Petawatt facility and implosions suitable for fast ignition have been tested on the Omega laser. A broad-based programme to develop lasers and ion beams for inertial fusion energy (IFE) is under way with excellent progress in drivers, chambers, target fabrication and target injection. KrF and diode pumped solid

  4. The US inertial confinement fusion (ICF) ignition programme and the inertial fusion energy (IFE) programme

    International Nuclear Information System (INIS)

    Lindl, J D; Hammel, B A; Logan, B Grant; Meyerhofer, David D; Payne, S A; Sethian, John D

    2003-01-01

    There has been rapid progress in inertial fusion in the past few years. This progress spans the construction of ignition facilities, a wide range of target concepts and the pursuit of integrated programmes to develop fusion energy using lasers, ion beams and z-pinches. Two ignition facilities are under construction, the national ignition facility (NIF) in the United States and the laser megajoule (LMJ) in France, and both projects are progressing towards an initial experimental capability. The laser integration line prototype beamline for LMJ and the first four beams of NIF will be available for experiments in 2003. The full 192 beam capability of NIF will be available in 2009 and ignition experiments are expected to begin shortly after that time. There is steady progress in target science and target fabrication in preparation for indirect-drive ignition experiments on NIF. Advanced target designs may lead to 5-10 times more yield than initial target designs. There has also been excellent progress on the science of ion beam and z-pinch-driven indirect-drive targets. Excellent progress on direct-drive targets has been obtained on the Omega laser at the University of Rochester. This includes improved performance of targets with a pulse shape predicted to result in reduced hydrodynamic instability. Rochester has also obtained encouraging results from initial cryogenic implosions. There is widespread interest in the science of fast ignition because of its potential for achieving higher target gain with lower driver energy and relaxed target fabrication requirements. Researchers from Osaka have achieved outstanding implosion and heating results from the Gekko XII Petawatt facility and implosions suitable for fast ignition have been tested on the Omega laser. A broad-based programme to develop lasers and ion beams for inertial fusion energy (IFE) is under way with excellent progress in drivers, chambers, target fabrication and target injection. KrF and diode pumped solid

  5. Inertial Confinement Fusion. Annual report 10/1/98 through 9/30/99

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, Jane

    1999-12-01

    General Atomics (GA) has served as the Inertial Confinement Fusion (ICF) Target Component Fabrication and Technology Development Support contractor for the U.S. Department of Energy since December 30, 1990. This report documents the technical activities of the period October 1, 1998 through September 30, 1999. During this period, GA and our partner Schafer Corporation were assigned 17 formal tasks in support of the ICF program and its five laboratories. A portion of the effort on these tasks included providing direct ''Onsite Support'' at Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), and Sandia National Laboratory (SNL). We fabricated and delivered over 1790 hohlraum mandrels and numerous other micromachined components to LLNL, LANL, and SNL. We produced more than 1380 glass and plastic target capsules over a wide range of sizes and designs (plus over 300 near target-quality capsules) for LLNL, LANL, SNL, and University of Rochester/Laboratory for Laser Energetic (UR/LLE). We also delivered various target foils and films for Naval Research Lab (NRL) and UWLLE in FY99. We fabricated a device to polish NIF-sized beryllium shells and prepared a laboratory for the safe operation of beryllium polishing activities. This report describes these target fabrication activities and the target fabrication and characterization development activities that made the deliveries possible. During FY99, the GA/Schafer portion of the GA/Schafer-UR/LLE-LANL team effort for design, procurement, installation, and testing of the OMEGA Cryogenic Target System (OCTS) that will field cryogenic targets on OMEGA was completed. All components of the OCTS were procured, fabricated, assembled, tested, and shipped to UR/LLE. Only minor documentation tasks remain to be done in FY00. The ICF program is anticipating experiments at the OMEGA laser and the National Ignition Facility (NIF) which will require targets containing cryogenic layered D

  6. 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

  7. Local wall power loading variations in thermonuclear fusion devices

    International Nuclear Information System (INIS)

    Carroll, M.C.; Miley, G.H.

    1989-01-01

    A 2 1/2-dimensional geometric model is presented that allows calculation of power loadings at various points on the first wall of a thermonuclear fusion device. Given average wall power loadings for brems-strahlung, cyclotron radiation charged particles, and neutrons, which are determined from various plasma-physics computation models, local wall heat loads are calculated by partitioning the plasma volume and surface into cells and superimposing the heating effects of the individual cells on selected first-wall differential areas. Heat loads from the entire plasma are thus determined as a function of position on the first-wall surface. Significant differences in local power loadings were found for most fusion designs, and it was therefore concluded that the effect of local power loading variations must be taken into account when calculating temperatures and heat transfer rates in fusion device first walls

  8. The first operation of the superconducting optimized stellarator fusion device Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Klinger, Thomas [Max-Planck-Institut fuer Plasmaphysik, Greifswald (Germany); Ernst-Moritz-Arndt Universitaet, Greifswald (Germany)

    2016-07-01

    The confinement of a high-temperature plasma by a suitable magnetic field is the most promising path to master nuclear fusion of Deuterium and Tritium on the scale of a reasonable power station. The two leading confinement concepts are the tokamak and the stellarator. Different from a tokamak, the stellarator does not require a strong current in the plasma but generates the magnetic field by external coils only. This has significant advantages, e.g. better stability properties and inherent steady-state capability. But stellarators need optimization, since ad hoc chosen magnetic field geometries lead to insufficient confinement properties, unfavourable plasma equilibria, and loss of fast particles. Wendelstein 7-X is a large (plasma volume 30 m{sup 3}) stellarator device with shaped superconducting coils that were determined via pure physics optimization criteria. After 19 years of construction, Wendelstein 7-X has now started operation. This talk introduces into the stellarator concept as a candidate for a future fusion power plant, summarizes the optimization principles, and presents the first experimental results with Helium and Hydrogen high temperature plasmas. An outlook on the physics program and the main goals of the project is given, too.

  9. 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.

  10. Application of spatially resolved high resolution crystal spectrometry to inertial confinement fusion plasmas.

    Science.gov (United States)

    Hill, K W; Bitter, M; Delgado-Aparacio, L; Pablant, N A; Beiersdorfer, P; Schneider, M; Widmann, K; Sanchez del Rio, M; Zhang, L

    2012-10-01

    High resolution (λ∕Δλ ∼ 10 000) 1D imaging x-ray spectroscopy using a spherically bent crystal and a 2D hybrid pixel array detector is used world wide for Doppler measurements of ion-temperature and plasma flow-velocity profiles in magnetic confinement fusion plasmas. Meter sized plasmas are diagnosed with cm spatial resolution and 10 ms time resolution. This concept can also be used as a diagnostic of small sources, such as inertial confinement fusion plasmas and targets on x-ray light source beam lines, with spatial resolution of micrometers, as demonstrated by laboratory experiments using a 250-μm (55)Fe source, and by ray-tracing calculations. Throughput calculations agree with measurements, and predict detector counts in the range 10(-8)-10(-6) times source x-rays, depending on crystal reflectivity and spectrometer geometry. Results of the lab demonstrations, application of the technique to the National Ignition Facility (NIF), and predictions of performance on NIF will be presented.

  11. Summary of the 16th IAEA Technical Meeting on 'Research using Small Fusion Devices'

    International Nuclear Information System (INIS)

    Gribkov, V.; Oost, G. van; Malaquias, A.; Herrera, J.

    2006-01-01

    Common research topics that are being studied in small, medium and large devices such as H-mode like or improved confinement, turbulence and transport are reported. These included modelling and diagnostic developments for edge and core, to characterize plasma density, temperature, electric potential, plasma flows, turbulence scale, etc. Innovative diagnostic methods were designed and implemented which could be used to develop experiments in small devices (in some cases not possible in large devices due to higher power deposition) to allow a better understanding of plasma edge and core properties. Reports are given addressing research in linear devices that can be used to study particular plasma physics topics relevant for other magnetic confinement devices such as the radial transport and the modelling of self-organized plasma jets involved in spheromak-like plasma formation. Some aspects of the work presented are of interest to the astrophysics community since they are believed to shed light on the basis of the physics of stellar jets. On the dense magnetized plasmas (DMP) topic, the present status of research, operation of new devices, plasma dynamics modelling and diagnostic developments is reported. The main devices presented belong to the class of Z-pinches, mostly plasma foci, and several papers were presented under this topic. The physics of DMP is important both for the main-stream fusion investigations as well as for providing the basis for elaboration of new concepts. New high-current technology introduced in the DMP devices design and construction make these devices nowadays more reliably fitted to various applications and give the possibility to widen the energy range used by them in both directions-to the multi-MJ level facilities and down to miniature plasma focus devices with energy of just a few J. (conference report)

  12. Modifications of the laser beam coherence inertial confinement fusion plasmas; Modifications des proprietes de coherence des faisceaux laser dans les plasmas de fusion par confinement inertiel

    Energy Technology Data Exchange (ETDEWEB)

    Grech, M

    2007-06-15

    Inertial confinement fusion by laser requires smoothed laser beam with well-controlled coherence properties. Such beams are made of many randomly distributed intensity maxima: the so-called speckles. As the laser beam propagates through plasma its temporal and spatial coherence can be reduced. This phenomenon is called plasma induced smoothing. For high laser intensities, instabilities developing independently inside the speckles are responsible for the coherence loss. At lower intensities, only collective effects, involving many speckles, can lead to induced smoothing. This thesis is a theoretical, numerical and experimental study of these mechanisms. Accounting for the partially incoherent behavior of the laser beams requires the use of statistical description of the laser-plasma interaction. A model is developed for the multiple scattering of the laser light on the self-induced density perturbations that is responsible for a spreading of the temporal and spatial spectra of the transmitted light. It also serves as a strong seed for the instability of forward stimulated Brillouin scattering that induces both, angular spreading and red-shift of the transmitted light. A statistical model is developed for this instability. A criterion is obtained that gives a laser power (below the critical power for filamentation) above which the instability growth is important. Numerical simulations with the interaction code PARAX and an experiment performed on the ALISE laser facility confirm the importance of these forward scattering mechanisms in the modification of the laser coherence properties. (author)

  13. Atomic processes in Inertial Electrostatic Confinement (IEC) devices

    International Nuclear Information System (INIS)

    Nebel, R.A.; Turner, L.; Tiouririne, T.N.; Barnes, D.C.; Nystrom, W.D.; Bussard, R.W.; Miley, G.H.; Javedani, J.; Yamamoto, Y.

    1993-01-01

    Inertial Electrostatic Confinement (IEC) is one of the earliest plasma confinement concepts, having first been suggested by P. T. Farnsworth in the 1950s. The concept involves a simple apparatus of concentric spherical electrostatic grids or a combination of grids and magnetic fields. An electrostatic structure is formed from the confluence of electron or ion beams. Gridded IEC systems have demonstrated neutron yields as high as 2*10 10 neutrons/sec. These systems have considerable potential as small, inexpensive, portable neutron sources for assaying applications. Neutron tomography is also a potential application. Atomic physics effects strongly influence the performance of all of these systems. Important atomic effects include elastic scattering, ionization, excitation, and charge exchange. This paper discusses how an IEC system is influenced by these effects and how to design around them. Theoretical modeling and experimental results are presented

  14. Kinetic transport in a magnetically confined and flux-constrained fusion plasma

    International Nuclear Information System (INIS)

    Darmet, G.

    2007-11-01

    This work deals with the kinetic transport in a fusion plasma magnetically confined and flux-constrained. The author proposes a new interpretation of the dynamics of zonal flows. The model that has been studied is a gyrokinetic model reduced to the transport of trapped ions. The inter-change stability that is generated allows the study of the kinetic transport of trapped ions. This model has a threshold instability and can be simulated over a few tens confining time for either thermal bath constraint or flux constraint. For thermal baths constraint, the simulation shows a metastable state where zonal flows are prevailing while turbulence is non-existent. In the case of a flux-constraint, zonal flows appear and relax by exchanging energy with system's kinetic energy and turbulence energy. The competition between zonal flows and turbulence can be then simulated by a predator-prey model. 2 regimes can be featured out: an improved confining regime where zonal flows dominate transport and a turbulent regime where zonal flows and turbulent transport are of the same magnitude order. We show that flux as well as the Reynolds tensor play an important role in the dynamics of the zonal flows and that the gyrokinetic description is relevant for all plasma regions. (A.C.)

  15. Assessment of liquid hydrogen cooled MgB2 conductors for magnetically confined fusion

    International Nuclear Information System (INIS)

    Glowacki, B A; Nuttall, W J

    2008-01-01

    Importantly environmental factors are not the only policy-driver for the hydrogen economy. Over the timescale of the development of fusion energy systems, energy security issues are likely to motivate a shift towards both hydrogen production and fusion as an energy source. These technologies combine local control of the system with the collaborative research interests of the major energy users in the global economy. A concept Fusion Island Reactor that might be used to generate H 2 (rather than electricity) is presented. Exploitation of produced hydrogen as a coolant and as a fuel is proposed in conjunction with MgB 2 conductors for the tokomak magnets windings, and electrotechnical devices for Fusion Island's infrastructure. The benefits of using MgB 2 over the Nb-based conductors during construction, operation and decommissioning of the Fusion Island Reactor are presented. The comparison of Nb 3 Sn strands for ITER fusion magnet with newly developed high field composite MgB 2 PIT conductors has shown that at 14 Tesla MgB 2 possesses better properties than any of the Nb 3 Sn conductors produced. In this paper the potential of MgB 2 conductors is examined for tokamaks of both the conventional ITER type and a Spherical Tokamak geometry. In each case MgB 2 is considered as a conductor for a range of field coil applications and the potential for operation at both liquid helium and liquid hydrogen temperatures is considered. Further research plans concerning the application of MgB 2 conductors for Fusion Island are also considered

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

    International Nuclear Information System (INIS)

    Park, Hyeon, K.

    1996-05-01

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

  17. Inertial confinement fusion: steady progress towards ignition and high gain (summary talk)

    International Nuclear Information System (INIS)

    Basko, M.M.

    2005-01-01

    Based on the results presented at the 20th IAEA Fusion Energy Conference 2004, this paper highlights the most important recent advances in inertial confinement fusion (ICF). With the construction of the National Ignition Facility (NIF) and the Laser Megajoule facility and many improvements in the target design, the conventional indirect-drive approach is advancing steadily towards the demonstration of ignition and high gain. The development of the polar direct-drive concept also made the prospects for direct-drive ignition on the NIF very favourable. Substantial progress was reported on the exploration of the fast-ignition approach to ICF. Parallel to that, multi-wire Z-pinches have become a competitive driver option for achieving ignition at the lowest possible cost. In heavy-ion fusion, experiments have been devoted so far to studying the generation, transport, and final focusing of high-current ion beams. A new concept for a power plant with a heavy-ion driver, based on a cylindrical direct-drive target compressed and ignited (in the fast-ignition mode) by two separate beams of very energetic (E i ≥ 0.5 GeV u -1 ) heavy ions, has been proposed

  18. Inertial confinement fusion: steady progress towards ignition and high gain (summary talk)

    Science.gov (United States)

    Basko, M. M.

    2005-10-01

    Based on the results presented at the 20th IAEA Fusion Energy Conference 2004, this paper highlights the most important recent advances in inertial confinement fusion (ICF). With the construction of the National Ignition Facility (NIF) and the Laser Mégajoule facility and many improvements in the target design, the conventional indirect-drive approach is advancing steadily towards the demonstration of ignition and high gain. The development of the polar direct-drive concept also made the prospects for direct-drive ignition on the NIF very favourable. Substantial progress was reported on the exploration of the fast-ignition approach to ICF. Parallel to that, multi-wire Z-pinches have become a competitive driver option for achieving ignition at the lowest possible cost. In heavy-ion fusion, experiments have been devoted so far to studying the generation, transport, and final focusing of high-current ion beams. A new concept for a power plant with a heavy-ion driver, based on a cylindrical direct-drive target compressed and ignited (in the fast-ignition mode) by two separate beams of very energetic (Ei>~ 0.5 GeV u-1) heavy ions, has been proposed.

  19. Inertial confinement fusion reaction chamber and power conversion system study. Final report

    International Nuclear Information System (INIS)

    Maya, I.; Schultz, K.R.; Bourque, R.F.

    1985-10-01

    This report summarizes the results of the second year of a two-year study on the design and evaluation of the Cascade concept as a commercial inertial confinement fusion (ICF) reactor. We developed a reactor design based on the Cascade reaction chamber concept that would be competitive in terms of both capital and operating costs, safe and environmentally acceptable in terms of hazard to the public, occupational exposure and radioactive waste production, and highly efficient. The Cascade reaction chamber is a double-cone-shaped rotating drum. The granulated solid blanket materials inside the rotating chamber are held against the walls by centrifugal force. The fusion energy is captured in a blanket of solid carbon, BeO, and LiAlO 2 granules. These granules are circulated to the primary side of a ceramic heat exchanger. Primary-side granule temperatures range from 1285 K at the LiAlO 2 granule heat exchanger outlet to 1600 K at the carbon granule heat exchanger inlet. The secondary side consists of a closed-cycle gas turbine power conversion system with helium working fluid, operating at 1300 K peak outlet temperature and achieving a thermal power conversion efficiency of 55%. The net plant efficiency is 49%. The reference design is a plant producing 1500 MW of D-T fusion power and delivering 815 MW of electrical power for sale to the utility grid. 88 refs., 44 figs., 47 tabs

  20. Inertial confinement fusion. 1995 ICF annual report, October 1994--September 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-06-01

    Lawrence Livermore National Laboratory`s (LLNL`s) Inertial Confinement Fusion (ICF) Program is a Department of Energy (DOE) Defense Program research and advanced technology development program focused on the goal of demonstrating thermonuclear fusion ignition and energy gain in the laboratory. During FY 1995, the ICF Program continued to conduct ignition target physics optimization studies and weapons physics experiments in support of the Defense Program`s stockpile stewardship goals. It also continued to develop technologies in support of the performance, cost, and schedule goals of the National Ignition Facility (NIF) Project. The NIF is a key element of the DOE`s Stockpile Stewardship and Management Program. In addition to its primary Defense Program goals, the ICF Program provides research and development opportunities in fundamental high-energy-density physics and supports the necessary research base for the possible long-term application to inertial fusion energy (IFE). Also, ICF technologies have had spin-off applications for industrial and governmental use. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  1. 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.

  2. Tungsten as First Wall Material in Fusion Devices

    International Nuclear Information System (INIS)

    Kaufmann, M.

    2006-01-01

    In the PLT tokamak with a tungsten limiter strong cooling of the central plasma was observed. Since then mostly graphite has been used as limiter or target plate material. Only a few tokamaks (limiter: FTU, TEXTOR; divertor: Alcator C-Mod, ASDEX Upgrade) gained experience with high-Z-materials. With the observed strong co- deposition of tritium together with carbon in JET and as a result of design studies of fusion reactors, it became clear that in the long run tungsten is the favourite for the first-wall material. Tungsten as a plasma facing material requires intensive research in all areas, i.e. in plasma physics, plasma wall-interaction and material development. Tungsten as an impurity in the confined plasma reveals considerable differences to carbon. Strong radiation at high temperatures, in connection with mostly a pronounced inward drift forms a particular challenge. Turbulent transport plays a beneficial role in this regard. The inward drift is an additional problem in the pedestal region of H-mode plasmas in ITER-like configurations. The erosion by low energy hydrogen atoms is in contrast to carbon small. However, erosion by fast particles from heating measures and impurity ions, accelerated in the sheath potential, play an important role in the case of tungsten. Radiation by carbon in the plasma boundary reduces the load to the target plates. Neon or Argon as substitutes will increase the erosion of tungsten. So far experiments have demonstrated that in most scenarios the tungsten content in the central plasma can be kept sufficiently small. The material development is directed to the specific needs of existing or future devices. In ASDEX Upgrade, which will soon be a divertor experiment with a complete tungsten first-wall, graphite tiles are coated with tungsten layers. In ITER, the solid tungsten armour of the target plates has to be castellated because of its difference in thermal expansion compared to the cooling structure. In a reactor the technical

  3. A new class of magnetic confinement device in the shape of a knot

    Energy Technology Data Exchange (ETDEWEB)

    Hudson, S. R., E-mail: shudson@pppl.gov; Startsev, E.; Feibush, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, New Jersey 08543 (United States)

    2014-01-15

    We describe a new class of magnetic confinement device, with the magnetic axis in the shape of a knot. We call such devices “knotatrons.” An example is given that has a large volume filled with magnetic surfaces, with significant rotational-transform, and with the magnetic field produced entirely by external circular coils.

  4. (Experimental development, testing and research work in support of the inertial confinement fusion program)

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.; Luckhardt, R.; Terry, N.; Drake, D.; Gaines, J. (eds.)

    1990-04-27

    This KMS Fusion Semi-Annual Technical Report covers the period October 1989 through March 1990. It contains a review of work performed by KMS Fusion, Inc. (KMSF), in support of the national program to achieve inertially confined fusion (ICF). A major section of the report is devoted to target technology, a field which is expected to play an increasingly important role in the overall KMSF fusion effort. Among the highlights of our efforts in this area covered in this report are: improvements and new developments in target fabrication techniques, including a discussion of techniques for introducing gaussian bumps and bands on target surfaces. Development of a single automated system for the interferometric characterization of transparent shells. Residual gas analysis of the blowing gases contained in glass shells made from xerogels. These usually include CO{sub 2}, O{sub 2} and N{sub 2}, and are objectionable because they dilute the fuel. Efforts to observe the ice layers formed in the {beta}-layering process in cryogenic targets, and to simulate the formation of these layers. In addition to our work on target technology, we conducted experiments with the Chroma laser and supported the ICF effort at other labs with theoretical and computational support as well as diagnostic development. Included in the work covered in this report are: experiments on Chroma to study interpenetration of and ionization balance in laser generated plasmas. Diagnostic development, including an optical probe for the Aurora laser at Los Alamos National Laboratory, and a high energy x-ray continuum spectrograph for Aurora. Investigation of the radiation cooling instability as a possible mechanism for the generation of relatively cold, dense jets observed in ICF experiments.

  5. Interfacing between concrete and steel construction and fusion research devices

    International Nuclear Information System (INIS)

    Willoughby, E.

    1981-01-01

    In 1976 Giffels Associates, Inc. an architect/engineer organization, was retained by the United States Department of Energy to provide Title I and Title II design services and Title III construction inspection services for the Tokamak Fusion Test Reactor now being installed at the Princeton Plasma Physics Laboratory in Princeton, New Jersey. Construction of the complex required to house and serve the reactor itself, designed by others, now commencing. During building construction several problems occurred with respect to the interface between the building design, construction and the fusion device (reactor). A brief description of some of these problems and related factors is presented, which may be of benefit to those persons active in continuing fusion research and experimental work

  6. Effects of Non-Maxwellian Plasma Species on ICRF Propagation and Absorption in Toroidal Magnetic Confinement Devices

    International Nuclear Information System (INIS)

    Dumont, R.J.; Phillips, C.K.; Smithe, D.N.

    2003-01-01

    Auxiliary heating supplied by externally launched electromagnetic waves is commonly used in toroidal magnetically confined fusion experiments for profile control via localized heating, current drive and perhaps flow shear. In these experiments, the confined plasma is often characterized by the presence of a significant population of non-thermal species arising from neutral beam injection, from acceleration of the particles by the applied waves, or from copious fusion reactions in future devices. Such non-thermal species may alter the wave propagation as well as the wave absorption dynamics in the plasma. Previous studies have treated the corresponding velocity distributions as either equivalent Maxwellians, or else have included realistic distributions only in the finite Larmor radius limit. In this work, the hot plasma dielectric response of the plasma has been generalized to treat arbitrary distribution functions in the non-relativistic limit. The generalized dielectric tensor has been incorporated into a one-dimensional full wave all-orders kinetic field code. Initial comparative studies of ion cyclotron range of frequency wave propagation and heating in plasmas with nonthermal species, represented by realistic distribution functions or by appropriately defined equivalent Maxwellians, have been completed for some specific experiments and are presented

  7. Progress in direct-drive inertial confinement fusion research at the laboratory for laser energetics

    International Nuclear Information System (INIS)

    McCrory, R.L.; Meyerhofer, D.D.; Loucks, S.J.

    2003-01-01

    Significant theoretical and experimental progress toward the validation of direct-drive inertial confinement fusion (ICF) has been made at the Laboratory for Laser Energetics (LLE). Direct-drive ICF offers the potential for high-gain implosions and is a leading candidate for an inertial fusion energy power plant. LLE's base-line direct-drive ignition design for the National Ignition Facility (NIF) is an 'all-DT' design that has a 1-D gain of ∼45 (∼30 when two-dimensional calculations are performed). The 'all-DT target' consists of a thin (∼3 μm) plastic shell enclosing a thick (∼330 μm) DT-ice layer. Recent calculations show that targets composed of foam shells, wicked with DT, can potentially achieve 1-D gains ∼100 at NIF energy levels (∼1.5 MJ). The addition of a 'picket' pulse to the beginning of the all-DT pulse shape reduces the target sensitivity to laser nonuniformities, increasing the potentially achievable gains. LLE experiments are conducted on the OMEGA 60-beam, 30-kJ, UV laser system. Beam smoothing includes 1-THz, 2-D SSD and polarization smoothing. Ignition-scaled cryogenic D 2 and plastic-shell spherical targets and a comprehensive suite of x-ray, nuclear, charged-particle, and optical diagnostics are used to understand the characteristics of the implosions. Recent cryogenic D 2 implosions with high adiabat (α ∼ 25) perform as predicted by one-dimensional (perfectly symmetric) simulations. Moderateconvergence- ratio (CR ∼ 15), high-adiabat (α ∼ 25), warm-capsule (surrogates for cryogenic capsules) implosions produce >30% of the 1-D predicted neutron yield and nearly 100% of the predicted fuel and shell areal densities. From a combination of x-ray, nuclear, and particle spectroscopy, a 'Lawson' fusion parameter (n i T i τi) of ∼7 x 10 20 m -3 keV was measured, the highest directly measured in inertial confinement fusion experiments to date. Estimates from cryogenic target performance give similar Lawson conditions. Future

  8. Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  9. Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  10. Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  11. Studies on the feasibility of heavy ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    1985-08-01

    This Annual Report summarizes experimental and theoretical investigations carried out in the framework of a feasibility study of inertial confinement fusion by heavy ion beams, funded by the Federal Ministry for Research and Technology. After the completion of the conceptual design study HIBALL with an upgraded version, the investigations concentrated in 1984 mainly on problems of accelerator and target physics. In the area of accelerator physics the main interest was in the production and acceleration of high intensity heavy ion beams of high phase space density and in beam dynamics theory, in the area of target physics on beam-target interaction, radiation hydrodynamics, instabilities and the equation of state of highly compressed hot matter. (orig./AH)

  12. Aurora multikilojoule KrF laser system prototype for inertial confinement fusion

    International Nuclear Information System (INIS)

    Rosocha, L.A.; Hanlon, J.A.; Mc Leod, J.; Kang, M.; Kortegaard, B.L.; Burrows, M.D.; Bowling, P.S.

    1987-01-01

    Aurora is the Los Alamos National Laboratory short-pulse, high-power, KrF laser system. It serves as an end-to-end technology demonstration for large-scale ultraviolet laser systems of interest for short wavelength, inertial confinement fusion (ICF) investigations. The systems is a prototype for using optical angular multiplexing and serial amplification by large electron-beam-driven KrF laser amplifiers to deliver stacked, 248-nm, 5-ns duration multikilojoule laser pulses to ICF targets using an --1-km-long optical beam path. The entire Aurora KrF laser system is described and the design features of the following major system components are summarized: front-end lasers, amplifier train, multiplexer, optical relay train, demultiplexer, target irradiation apparatus, and alignment and controls systems

  13. Investigation of methods for fabricating, characterizing, and transporting cryogenic inertial-confinement-fusion tartets

    International Nuclear Information System (INIS)

    Fanning, J.J.; Kim, K.

    1981-01-01

    The objective of this work is to investigate methods for fabricating, characterizing and transporting cryogenic inertial confinement fusion targets on a continuous basis. A microprocessor-based data acquisition system has been built that converts a complete target image to digital data, which are then analyzed by automated software procedures. The low temperatures required to freeze the hydrogen isotopes contained in a target is provided by a cryogenic cold chamber capable of attaining 15 K. A new method for target manipulation and positioning is studied that employs molecular gas beams to levitate a target and an electrostatic quadrupole structure to provide for its lateral containment. Since the electrostatic target-positioning scheme requires that the targets be charged, preliminary investigation has been carried out for a target-charging mechanism based on ion-bombardment

  14. Laser ablation under different electron heat conduction models in inertial confinement fusion

    Science.gov (United States)

    Li, Shuanggui; Ren, Guoli; Huo, Wen Yi

    2018-06-01

    In this paper, we study the influence of three different electron heat conduction models on the laser ablation of gold plane target. Different from previous studies, we concentrate on the plasma conditions, the conversion efficiency from laser into soft x rays and the scaling relation of mass ablation, which are relevant to hohlraum physics study in indirect drive inertial confinement fusion. We find that the simulated electron temperature in corona region is sensitive to the electron heat conduction models. For different electron heat conduction models, there are obvious differences in magnitude and spatial profile of electron temperature. For the flux limit model, the calculated conversion efficiency is sensitive to flux limiters. In the laser ablation of gold, most of the laser energies are converted into x rays. So the scaling relation of mass ablation rate is quite different from that of low Z materials.

  15. Thin shell, high velocity inertial confinement fusion implosions on the national ignition facility.

    Science.gov (United States)

    Ma, T; Hurricane, O A; Callahan, D A; Barrios, M A; Casey, D T; Dewald, E L; Dittrich, T R; Döppner, T; Haan, S W; Hinkel, D E; Berzak Hopkins, L F; Le Pape, S; MacPhee, A G; Pak, A; Park, H-S; Patel, P K; Remington, B A; Robey, H F; Salmonson, J D; Springer, P T; Tommasini, R; Benedetti, L R; Bionta, R; Bond, E; Bradley, D K; Caggiano, J; Celliers, P; Cerjan, C J; Church, J A; Dixit, S; Dylla-Spears, R; Edgell, D; Edwards, M J; Field, J; Fittinghoff, D N; Frenje, J A; Gatu Johnson, M; Grim, G; Guler, N; Hatarik, R; Herrmann, H W; Hsing, W W; Izumi, N; Jones, O S; Khan, S F; Kilkenny, J D; Knauer, J; Kohut, T; Kozioziemski, B; Kritcher, A; Kyrala, G; Landen, O L; MacGowan, B J; Mackinnon, A J; Meezan, N B; Merrill, F E; Moody, J D; Nagel, S R; Nikroo, A; Parham, T; Ralph, J E; Rosen, M D; Rygg, J R; Sater, J; Sayre, D; Schneider, M B; Shaughnessy, D; Spears, B K; Town, R P J; Volegov, P L; Wan, A; Widmann, K; Wilde, C H; Yeamans, C

    2015-04-10

    Experiments have recently been conducted at the National Ignition Facility utilizing inertial confinement fusion capsule ablators that are 175 and 165  μm in thickness, 10% and 15% thinner, respectively, than the nominal thickness capsule used throughout the high foot and most of the National Ignition Campaign. These three-shock, high-adiabat, high-foot implosions have demonstrated good performance, with higher velocity and better symmetry control at lower laser powers and energies than their nominal thickness ablator counterparts. Little to no hydrodynamic mix into the DT hot spot has been observed despite the higher velocities and reduced depth for possible instability feedthrough. Early results have shown good repeatability, with up to 1/2 the neutron yield coming from α-particle self-heating.

  16. Inhibition of turbulence in inertial-confinement-fusion hot spots by viscous dissipation.

    Science.gov (United States)

    Weber, C R; Clark, D S; Cook, A W; Busby, L E; Robey, H F

    2014-05-01

    Achieving ignition in inertial confinement fusion (ICF) requires the formation of a high-temperature (>10 keV) central hot spot. Turbulence has been suggested as a mechanism for degrading the hot-spot conditions by altering transport properties, introducing colder, mixed material, or reducing the conversion of radially directed kinetic energy to hot-spot heating. We show, however, that the hot spot is very viscous, and the assumption of turbulent conditions in the hot spot is incorrect. This work presents the first high-resolution, three-dimensional simulations of National Ignition Facility (NIF) implosion experiments using detailed knowledge of implosion dynamics and instability seeds and including an accurate model of physical viscosity. We find that when viscous effects are neglected, the hot spot can exhibit a turbulent kinetic energy cascade. Viscous effects, however, are significant and strongly damp small-scale velocity structures, with a hot-spot Reynolds number in the range of only 10-100.

  17. 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

  18. First Liquid Layer Inertial Confinement Fusion Implosions at the National Ignition Facility

    Science.gov (United States)

    Olson, R. E.; Leeper, R. J.; Kline, J. L.; Zylstra, A. B.; Yi, S. A.; Biener, J.; Braun, T.; Kozioziemski, B. J.; Sater, J. D.; Bradley, P. A.; Peterson, R. R.; Haines, B. M.; Yin, L.; Berzak Hopkins, L. F.; Meezan, N. B.; Walters, C.; Biener, M. M.; Kong, C.; Crippen, J. W.; Kyrala, G. A.; Shah, R. C.; Herrmann, H. W.; Wilson, D. C.; Hamza, A. V.; Nikroo, A.; Batha, S. H.

    2016-12-01

    The first cryogenic deuterium and deuterium-tritium liquid layer implosions at the National Ignition Facility (NIF) demonstrate D2 and DT layer inertial confinement fusion (ICF) implosions that can access a low-to-moderate hot-spot convergence ratio (12 30 ) DT ice layer implosions. Although high CR is desirable in an idealized 1D sense, it amplifies the deleterious effects of asymmetries. To date, these asymmetries prevented the achievement of ignition at the NIF and are the major cause of simulation-experiment disagreement. In the initial liquid layer experiments, high neutron yields were achieved with CRs of 12-17, and the hot-spot formation is well understood, demonstrated by a good agreement between the experimental data and the radiation hydrodynamic simulations. These initial experiments open a new NIF experimental capability that provides an opportunity to explore the relationship between hot-spot convergence ratio and the robustness of hot-spot formation during ICF implosions.

  19. Onset of hydrodynamic mix in high-velocity, highly compressed inertial confinement fusion implosions.

    Science.gov (United States)

    Ma, T; Patel, P K; Izumi, N; Springer, P T; Key, M H; Atherton, L J; Benedetti, L R; Bradley, D K; Callahan, D A; Celliers, P M; Cerjan, C J; Clark, D S; Dewald, E L; Dixit, S N; Döppner, T; Edgell, D H; Epstein, R; Glenn, S; Grim, G; Haan, S W; Hammel, B A; Hicks, D; Hsing, W W; Jones, O S; Khan, S F; Kilkenny, J D; Kline, J L; Kyrala, G A; Landen, O L; Le Pape, S; MacGowan, B J; Mackinnon, A J; MacPhee, A G; Meezan, N B; Moody, J D; Pak, A; Parham, T; Park, H-S; Ralph, J E; Regan, S P; Remington, B A; Robey, H F; Ross, J S; Spears, B K; Smalyuk, V; Suter, L J; Tommasini, R; Town, R P; Weber, S V; Lindl, J D; Edwards, M J; Glenzer, S H; Moses, E I

    2013-08-23

    Deuterium-tritium inertial confinement fusion implosion experiments on the National Ignition Facility have demonstrated yields ranging from 0.8 to 7×10(14), and record fuel areal densities of 0.7 to 1.3 g/cm2. These implosions use hohlraums irradiated with shaped laser pulses of 1.5-1.9 MJ energy. The laser peak power and duration at peak power were varied, as were the capsule ablator dopant concentrations and shell thicknesses. We quantify the level of hydrodynamic instability mix of the ablator into the hot spot from the measured elevated absolute x-ray emission of the hot spot. We observe that DT neutron yield and ion temperature decrease abruptly as the hot spot mix mass increases above several hundred ng. The comparison with radiation-hydrodynamic modeling indicates that low mode asymmetries and increased ablator surface perturbations may be responsible for the current performance.

  20. Studies on the feasibility of heavy-ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    1982-04-01

    This annual report summarizes the scientific results of work carried out in 1981 in the framework of a feasibility study for inertial confinement fusion (ICF) with heavy ion beams. This program, established in autumn 1979 and funded by the German Ministry for Science and Technology, is devoted in a first period until 1984 to the study of fundamental aspects of the field. Its principal aims are the investigation of key problems and the identification of critical issues of the heavy ion ICF concept in the fields of accelerator research, pellet physics, atomic physics, and reactor design. The research is carried out by about ten working groups at various German research centers and universities. In addition, together with a group of the University of Wisconsin a conceptual design study for a reactor plant (HIBALL) has been started in 1980 and was continued 1981. (orig.) [de

  1. Anomalous deceleration of light ion beam in plasm of inertial confinement fusion

    International Nuclear Information System (INIS)

    Abe, Takashi; Niu, Keishiro

    1981-01-01

    The ion beam propagation in inertial confinement fusion by light ion beam is analysed. The anomalous deceleration of the beam ion occurs, when the beam including the electron interacts with the background plasma with a comparable number density. This deceleration is caused by the two stream instability between the beam and the background plasma electrons and then becomes maximum when each density is equivalent. The anomalous deceleration rate of the beam ion is computed by using the quasilinear theory. It is shown that the anomalous deceleration which the beam ion (10 17 cm - 3 ) accepts from the background plasma (10 18 cm - 3 ) is equivalent to the classical one from the background plasma with solid density (10 21 cm - 3 ). (author)

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

    International Nuclear Information System (INIS)

    Mancini, R. C.

    2007-01-01

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

  3. Diagnostic for determining the mix in inertial confinement fusion capsule hotspot

    Energy Technology Data Exchange (ETDEWEB)

    He, Shibei; Ding, Yongkun, E-mail: ding-yk@vip.sina.com; Miao, Wenyong; Zhang, Xing; Tu, Shaoyong; Yuan, Yongteng; Pu, Yudong; Yan, Ji; Wei, Minxi; Yin, Chuansheng [Research Center of Laser Fusion, China Academy of Engineering Physics, Mianyang 621900 (China)

    2016-07-15

    A diagnostic is developed for determining the hotspot mix in inertial confinement fusion experiments. A multi-channel pinhole camera measures Bremsstrahlung emissions from implosion capsules ranging from 6 keV to 30 keV and records an image of the hotspot. Meanwhile, a planar crystal spectrometer measures Ar line emissions used to deduce the electron density of the hotspot. An X-ray streaked camera records the burn duration. With the Bremsstrahlung spectrum, electron density, hotspot volume, and burn duration, the mix quantity is determined by solving a pair of linear equations. This inferred mix amount has an uncertainty due to the uncertainty of the electron density, but with the help of the measured neutron product, the most likely mix quantity value can be determined. This technique is applied to experimental images to infer the quantity of CH ablator mix into the hotspot.

  4. Diagnostic for determining the mix in inertial confinement fusion capsule hotspot

    International Nuclear Information System (INIS)

    He, Shibei; Ding, Yongkun; Miao, Wenyong; Zhang, Xing; Tu, Shaoyong; Yuan, Yongteng; Pu, Yudong; Yan, Ji; Wei, Minxi; Yin, Chuansheng

    2016-01-01

    A diagnostic is developed for determining the hotspot mix in inertial confinement fusion experiments. A multi-channel pinhole camera measures Bremsstrahlung emissions from implosion capsules ranging from 6 keV to 30 keV and records an image of the hotspot. Meanwhile, a planar crystal spectrometer measures Ar line emissions used to deduce the electron density of the hotspot. An X-ray streaked camera records the burn duration. With the Bremsstrahlung spectrum, electron density, hotspot volume, and burn duration, the mix quantity is determined by solving a pair of linear equations. This inferred mix amount has an uncertainty due to the uncertainty of the electron density, but with the help of the measured neutron product, the most likely mix quantity value can be determined. This technique is applied to experimental images to infer the quantity of CH ablator mix into the hotspot.

  5. Investigation of methods for fabricating, characterizing, and transporting cryogenic inertial-confinement-fusion tartets

    Energy Technology Data Exchange (ETDEWEB)

    Fanning, J.J.; Kim, K.

    1981-01-01

    The objective of this work is to investigate methods for fabricating, characterizing and transporting cryogenic inertial confinement fusion targets on a continuous basis. A microprocessor-based data acquisition system has been built that converts a complete target image to digital data, which are then analyzed by automated software procedures. The low temperatures required to freeze the hydrogen isotopes contained in a target is provided by a cryogenic cold chamber capable of attaining 15 K. A new method for target manipulation and positioning is studied that employs molecular gas beams to levitate a target and an electrostatic quadrupole structure to provide for its lateral containment. Since the electrostatic target-positioning scheme requires that the targets be charged, preliminary investigation has been carried out for a target-charging mechanism based on ion-bombardment.

  6. ENEA-Frascati inertial confinement fusion: Multi-channel digitizer control, acquisition and analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, A.; Strangio, C.; Elman, G.; Hugh, J. Mc; Rogers, G.; Pastina, E.; Raimondi, F.; Umbro, A.

    1991-09-01

    The ICF (Inertial Confinement Fusion) data acquisition and analysis system described in this paper incorporates digitizer channels, signal switching, an instrument controller, a graphic hardcopy unit, laser printer and software. The digitizers, signal switching and instrument controller are standard components appropriate to acquire the single fast shot signals (rise-time: from about 100 picoseconds to nanoseconds). The input signals are switched to the digitizers through the TSI-8150 test system interface by TSS40 switch controller cards and TSS46 18GHz microwave switches. Graphic hardcopy is accomplished using either a 4693PX colour hardcopy unit or a laser printer connected through a printer spooler/multiplexer. The software is based on an existing package from Ressler called RAI/DAC reviewed by Fus-Inerz and Tektronix-Italy to define the modifications needed for data acquisition and handling. The adopted software solution is based on an IBM PC compatible instrument controller running software developed by Ressler Associates.

  7. Inertial Confinement Fusion Quarterly Report: April--June 1993. Volume 3, Number 3

    Energy Technology Data Exchange (ETDEWEB)

    MacGowan, B.J.; Kotowski, M.; Schleich, D. [eds.

    1993-11-01

    This issue of the ICF Quarterly contains six articles describing recent advances in Lawrence Livermore National Laboratory`s inertial confinement fusion (ICF) program. The current emphasis of the ICF program is in support of DOE`s National Ignition Facility (NIF) initiative for demonstrating ignition and gain with a 1-2 MJ glass laser. The articles describe recent Nova experiments and investigations tailored towards enhancing understanding of the key physics and technological issues for the NIF. Titles of the articles are: development of large-aperture KDP crystals; inner-shell photo-ionized X-ray lasers; X-ray radiographic measurements of radiation-driven shock and interface motion in solid density materials; the role of nodule defects in laser-induced damage of multilayer optical coatings; techniques for Mbar to near-Gbar equation-of-state measurements with the Nova laser; parametric instabilities and laser-beam smoothing.

  8. Advances in compact proton spectrometers for inertial-confinement fusion and plasma nuclear science.

    Science.gov (United States)

    Seguin, F H; Sinenian, N; Rosenberg, M; Zylstra, A; Manuel, M J-E; Sio, H; Waugh, C; Rinderknecht, H G; Johnson, M Gatu; Frenje, J; Li, C K; Petrasso, R; Sangster, T C; Roberts, S

    2012-10-01

    Compact wedge-range-filter proton spectrometers cover proton energies ∼3-20 MeV. They have been used at the OMEGA laser facility for more than a decade for measuring spectra of primary D(3)He protons in D(3)He implosions, secondary D(3)He protons in DD implosions, and ablator protons in DT implosions; they are now being used also at the National Ignition Facility. The spectra are used to determine proton yields, shell areal density at shock-bang time and compression-bang time, fuel areal density, and implosion symmetry. There have been changes in fabrication and in analysis algorithms, resulting in a wider energy range, better accuracy and precision, and better robustness for survivability with indirect-drive inertial-confinement-fusion experiments.

  9. A novel three-axis cylindrical hohlraum designed for inertial confinement fusion ignition

    Science.gov (United States)

    Kuang, Longyu; Li, Hang; Jing, Longfei; Lin, Zhiwei; Zhang, Lu; Li, Liling; Ding, Yongkun; Jiang, Shaoen; Liu, Jie; Zheng, Jian

    2016-10-01

    A novel ignition hohlraum for indirect-drive inertial confinement fusion is proposed, which is named three-axis cylindrical hohlraum (TACH). TACH is a kind of 6 laser entrance holes (LEHs) hohlraum, which is orthogonally jointed of three cylindrical hohlraums. Laser beams are injected through every entrance hole with the same incident angle of 55°. A view-factor simulation result shows that the time-varying drive asymmetry of TACH is less than 1.0% in the whole drive pulse period without any supplementary technology. Coupling efficiency of TACH is close to that of 6 LEHs spherical hohlraum with corresponding size. Its plasma-filling time is close to that of typical cylindrical ignition hohlraum. Its laser plasma interaction has as low backscattering as the outer cone of the cylindrical ignition hohlraum. Therefore, TACH combines most advantages of various hohlraums and has little predictable risk, providing an important competitive candidate for ignition hohlraum.

  10. Summary of research for the Inertial Confinement Fusion Program at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Cartwright, D.C.

    1985-03-01

    The information presented in this report is a summary of the status of the Inertial Confinement Fusion (ICF) program at the Los Alamos National Laboratory as of February 1985. This report contains material on the existing high-power CO 2 laser driver (Antares), the program to determine the potential of KrF as an ICF driver, heavy-ion accelerators as drivers for ICF, target fabrication for ICF, and a summary of our understanding of laser-plasma interactions. A classified companion report contains material on our current understanding of capsule physics and lists the contributions to the Laboratory's weapons programs made by the ICF program. The information collected in these two volumes is meant to serve as a report on the status of some of the technological components of the Los Alamos ICF program rather than a detailed review of specific technical issues

  11. Hot-spot mix in ignition-scale inertial confinement fusion targets.

    Science.gov (United States)

    Regan, S P; Epstein, R; Hammel, B A; Suter, L J; Scott, H A; Barrios, M A; Bradley, D K; Callahan, D A; Cerjan, C; Collins, G W; Dixit, S N; Döppner, T; Edwards, M J; Farley, D R; Fournier, K B; Glenn, S; Glenzer, S H; Golovkin, I E; Haan, S W; Hamza, A; Hicks, D G; Izumi, N; Jones, O S; Kilkenny, J D; Kline, J L; Kyrala, G A; Landen, O L; Ma, T; MacFarlane, J J; MacKinnon, A J; Mancini, R C; McCrory, R L; Meezan, N B; Meyerhofer, D D; Nikroo, A; Park, H-S; Ralph, J; Remington, B A; Sangster, T C; Smalyuk, V A; Springer, P T; Town, R P J

    2013-07-26

    Mixing of plastic ablator material, doped with Cu and Ge dopants, deep into the hot spot of ignition-scale inertial confinement fusion implosions by hydrodynamic instabilities is diagnosed with x-ray spectroscopy on the National Ignition Facility. The amount of hot-spot mix mass is determined from the absolute brightness of the emergent Cu and Ge K-shell emission. The Cu and Ge dopants placed at different radial locations in the plastic ablator show the ablation-front hydrodynamic instability is primarily responsible for hot-spot mix. Low neutron yields and hot-spot mix mass between 34(-13,+50)  ng and 4000(-2970,+17 160)  ng are observed.

  12. Transport vehicle for manned Mars missions powered by inertial confinement fusion

    International Nuclear Information System (INIS)

    Orth, C.D.; Klein, G.; Sercel, J.; Hoffman, N.; Murray, K.; Chang-Diaz, F.

    1987-01-01

    Inertial confinement fusion (ICF) is an ideal engine power source for manned spacecraft to Mars because of its inherently high power-to-mass ratios and high specific impulses. We have produced a concept for a vehicle powered by ICF and utilizing a magnetic thrust chamber to avoid plasma thermalization with wall structures and the resultant degradation of specific impulse that are unavoidable with the use of mechanical thrust chambers. This vehicle is capable of 100-day manned Mars missions with a 100-metric-ton payload and a total vehicle launch mass near 6000 metric tons, based on advanced technology assumed to be available by A.D. 2020. Such short-duration missions minimize radiation exposures and physiological deterioration of astronauts

  13. Development of aerogel-lined targets for inertial confinement fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Braun, Tom [Technical Univ. Munchen (Germany)

    2013-03-28

    This thesis explores the formation of ICF compatible foam layers inside of an ablator shell used for inertial confinement fusion experiments at the National Ignition Facility. In particular, the capability of p- DCPD polymer aerogels to serve as a scaffold for the deuterium-tritium mix was analyzed. Four different factors were evaluated: the dependency of different factors such as thickness or composition of a precursor solution on the uniformity of the aerogel layer, how to bring the optimal composition inside of the ablator shell, the mechanical stability of ultra-low density p-DCPD aerogel bulk pieces during wetting and freezing with hydrogen, and the wetting behavior of thin polymer foam layers in HDC carbon ablator shells with liquid deuterium. The research for thesis was done at Lawrence Livermore National Laboratory in cooperation with the Technical University Munich.

  14. Fabrication of thin-wall, freestanding inertial confinement fusion targets by chemical vapor deposition

    International Nuclear Information System (INIS)

    Carroll, D.W.; McCreary, W.J.

    1982-01-01

    To meet the requirements for plasma physics experiments in the inertial confinement fusion (ICF) program, chemical vapor deposition (CVD) in fluid beds was used to fabricate freestanding tungsten spheres and cylinders with wall thicknesses less than 5.0 μm. Molybdenum and molybdenum alloy (TZM) mandrels of the desired geometry were suspended in a carrier bed of dense microspheres contained in an induction-heated fluid-bed reactor. The mandrels were free to float randomly through the bed, and using the reaction WF 6 +3H 2 →/sub /KW +6HF, very fine-grained tungsten was deposited onto the surface at a rate and in a grain size determined by temperature, gas flow rate, system pressure, and duration of the reaction. After coating, a portion of each mandrel was exposed by hole drilling or grinding. The mandrel was then removed by acid leaching, leaving a freestanding tungsten shape. Experimental procedures, mandrel preparation, and results obtained are discussed

  15. Models and analyses for inertial-confinement fusion-reactor studies

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.

    1981-05-01

    This report describes models and analyses devised at Los Alamos National Laboratory to determine the technical characteristics of different inertial confinement fusion (ICF) reactor elements required for component integration into a functional unit. We emphasize the generic properties of the different elements rather than specific designs. The topics discussed are general ICF reactor design considerations; reactor cavity phenomena, including the restoration of interpulse ambient conditions; first-wall temperature increases and material losses; reactor neutronics and hydrodynamic blanket response to neutron energy deposition; and analyses of loads and stresses in the reactor vessel walls, including remarks about the generation and propagation of very short wavelength stress waves. A discussion of analytic approaches useful in integrations and optimizations of ICF reactor systems concludes the report

  16. Numerical Experiments Providing New Insights into Plasma Focus Fusion Devices

    Directory of Open Access Journals (Sweden)

    Sing Lee

    2010-04-01

    Full Text Available Recent extensive and systematic numerical experiments have uncovered new insights into plasma focus fusion devices including the following: (1 a plasma current limitation effect, as device static inductance is reduced towards very small values; (2 scaling laws of neutron yield and soft x-ray yield as functions of storage energies and currents; (3 a global scaling law for neutron yield as a function of storage energy combining experimental and numerical data showing that scaling deterioration has probably been interpreted as neutron ‘saturation’; and (4 a fundamental cause of neutron ‘saturation’. The ground-breaking insights thus gained may completely change the directions of plasma focus fusion research.

  17. Research progress in intense ion beam production for inertial confinement fusion at Cornell University

    International Nuclear Information System (INIS)

    Bluhm, H.; Greenly, J.B.; Hammer, D.A.

    1983-01-01

    Recent results obtained in the generation of intense pulsed light ion beams and their application to inertial confinement fusion are described. Studies of time-integrated and time-dependent beam divergence using a magnetically insulated ion diode with a ''flashboard'' anode at 11 W diode power show a directionality which is apparently due to electron dynamics in the diode. Nevertheless, ion beams having divergence angle as small as 0.5 0 have been produced at >10 8 W.cm - 2 . In another experiment with a similar diode, the anode plasma formation time varied with the detailed anode configuration, the diode voltage and the insulating magnetic field, with the longer times obtained at lower voltage and higher insulating magnetic field strength. The anode plasma density was determined to be in the 10 15 cm - 3 density range and to move away from the anode at approx.2 cm.μs - 1 in another similar experiment. Preliminary experiments performed on a 10 12 W accelerator show reasonable power coupling to a magnetically insulated ion diode, with >10 9 W.cm - 2 beams at approx.1.5 MV being generated. Computer simulations suggest that if such a beam can be focused into a plasma channel, most of its energy can be delivered to a pellet one to two metres away. In experiments on the applied Bsub(theta) diode, microwave radiation, ion production efficiency, and ion beam fluctuations all reach a maximum when the insulating magnetic field is about 1.4 times the critical field for magnetic insulation. Finally, relatively pure beams of heavy ions have been produced by making the anode with hydrocarbon-free dielectric material which contains the desired species together with other ions having substantially higher ionization potential. The sum of these results suggests that flashboard anodes operated at the few-MV level can be used to produce beams with properties suitable for inertial confinement fusion experiments on sufficiently powerful pulsed power generators. (author)

  18. On the possibility of D-3He fusion based on fast - ignition inertial confinement scheme

    International Nuclear Information System (INIS)

    Nakao, Y.; Hegi, K.; Ohmura, T.; Katsube, M.; Kudo, K.; Johzaki, T.; Ohta, M.

    2007-01-01

    Although nuclear fusion reactors adopting D 3 He fuel could provide many advantages, such as low neutron generation and efficient conversion of output fusion energy, the achievement of ignition is a difficult problem. It is therefore of particular importance to find some methods or schemes that relax the ignition requirements. In inertial confinement scheme, the use of pure D 3 He fuel is impractical because of the excessive requirement on driver energy. A small amount of DT fuel as 'igniter' is hence indispensable [1]. Our previous burn simulation [1] for DT/D 3 He fuels compressed to 5000 times the liquid density showed that substantial fuel gains (∼500) are obtained from fuels having parameters ρ R D T = 3 g/cm 2 , ρ R t otal 14 g/cm 2 and a central spark temperature of 5 keV. The driver energy needed to achieve these gains is estimated to be ∼30 MJ when the coupling efficiency is 10%; in this case the target gain is ∼50. Subsequent implosion simulation [2], however, showed that after void closure the central DT fuel is ignited while the bulk of the main D 3 He fuel is still imploding with high velocities. This pre-ignition of DT fuel leads to a low compression of the main fuel and prevents the DT/D 3 He fuel from obtaining required gain. These difficulties associated with the pre-ignition of DT fuel could be resolved or mitigated if other ignition schemes such as fast-ignition [3] and/or impact-ignition [4] are adopted, because in these schemes compression and ignition phases are separated. Furthermore, the reduction of driver energy can be expected. In the present study, we examine the possibility of D 3 He fusion in the fast-ignition scheme. Simulations until now have been made for a DT/D 3 He fuel compressed to 5000 times the liquid density by using FIBMET (2D fusion ignition and burning code) [5] and a newly developed neutron diffusion code. DT igniter was assumed to be placed at a corner of the compressed fuel. The ρ R values and temperature of

  19. Mode 1 drive asymmetry in inertial confinement fusion implosions on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Spears, Brian K., E-mail: spears9@llnl.gov; Edwards, M. J.; Hatchett, S.; Kritcher, A.; Lindl, J.; Munro, D.; Patel, P.; Robey, H. F.; Town, R. P. J. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551-0808 (United States); Kilkenny, J. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Knauer, J. [Laboratory for Laser Energetics, 250 E. River Road Rochester, New York 14623-1212 (United States)

    2014-04-15

    Mode 1 radiation drive asymmetry (pole-to-pole imbalance) at significant levels can have a large impact on inertial confinement fusion implosions at the National Ignition Facility (NIF). This asymmetry distorts the cold confining shell and drives a high-speed jet through the hot spot. The perturbed hot spot shows increased residual kinetic energy and reduced internal energy, and it achieves reduced pressure and neutron yield. The altered implosion physics manifests itself in observable diagnostic signatures, especially the neutron spectrum which can be used to measure the neutron-weighted flow velocity, apparent ion temperature, and neutron downscattering. Numerical simulations of implosions with mode 1 asymmetry show that the resultant simulated diagnostic signatures are moved toward the values observed in many NIF experiments. The diagnostic output can also be used to build a set of integrated implosion performance metrics. The metrics indicate that P{sub 1} has a significant impact on implosion performance and must be carefully controlled in NIF implosions.

  20. Impurity studies in fusion devices using laser-fluorescence-spectroscopy

    International Nuclear Information System (INIS)

    Husinsky, W.R.

    1980-08-01

    Resonance fluorescence excitation of neutral atoms using tunable radiation from dye lasers offers a number of unique advantages for impurity studies in fusion devices. Using this technique, it is possible to perform local, time-resolved measurements of the densities and velocity distributions of metallic impurities in fusion devices without disturbing the plasma. Velocities are measured by monitoring the fluorescence intensity while tuning narrow bandwidth laser radiation through the Doppler - broadened absorbtion spectrum of the transition. The knowledge of the velocity distribution of neutral impurities is particularly useful for the determination of impurity introduction mechanisms. The laser fluorescence technique will be described in terms of its application to metallic impurities in fusion devices and related laboratory experiments. Particular attention will be given to recent results from the ISX-B tokamak using pulsed dye lasers where detection sensitivities for neutral Fe of 10 6 atoms/cm 3 with a velocity resolution of 600 m/sec (0.1 eV) have been achieved. Techniques for exciting plasma particles (H,D) will also be discussed

  1. A generalized scaling law for the ignition energy of inertial confinement fusion capsules

    International Nuclear Information System (INIS)

    Herrmann, M.C.

    2001-01-01

    The minimum energy needed to ignite an inertial confinement fusion capsule is of considerable interest in the optimization of an inertial fusion driver. Recent computational work investigating this minimum energy has found that it depends on the capsule implosion history, in particular, on the capsule drive pressure. This dependence is examined using a series of LASNEX simulations to find ignited capsules which have different values of the implosion velocity, fuel adiabat and drive pressure. It is found that the main effect of varying the drive pressure is to alter the stagnation of the capsule, changing its stagnation adiabat, which, in turn, affects the energy required for ignition. To account for this effect a generalized scaling law has been devised for the ignition energy, E ign ∝α if 1.88±0.05 υ -5.89±0.12 P -0.77±0.03 . This generalized scaling law agrees with the results of previous work in the appropriate limits. (author)

  2. The influence of asymmetry on mix in direct-drive inertial confinement fusion experiments

    International Nuclear Information System (INIS)

    Christensen, C.R.; Wilson, D.C.; Barnes, Cris W.; Grim, G.P.; Morgan, G.L.; Wilke, M.D.; Marshall, F.J.; Glebov, V.Yu.; Stoeckl, C.

    2004-01-01

    The mix of shell material into the fuel of inertial confinement fusion (ICF) implosions is thought to be a major cause of the failure of most ICF experiments to achieve the fusion yield predicted by computer codes. Implosion asymmetry is a simple measurable quantity that is expected to affect the mix. In order to measure the coupling of asymmetry to mix in ICF implosions, we have performed experiments on the OMEGA laser [T. R. Boehly et al., Rev. Sci. Instrum. 66, 508 (1995)] that vary the energy of each of the sixty beams individually to achieve a given fraction of L2, the second-order Legendre polynomial. Prolate, symmetric, and oblate implosions resulted. Three different fill pressures were used. Simultaneous x-ray and neutron images were obtained. The experiments were modeled with a radiation/hydrodynamics code using the multi-fluid interpenetration mix model of Scannapieco and Cheng. It fits the data well with a single value of its one adjustable parameter (0.07±0.01). This agreement is demonstrated by neutron yield, x-ray images, neutron images, and ion temperatures. The degree of decline of the neutron yield with asymmetry at different fill pressures provides a hard constraint on ICF mix modeling

  3. Mode-selective symmetry control for indirect-drive inertial confinement fusion hohlraums

    International Nuclear Information System (INIS)

    Vesey, R. A.; Slutz, S. A.; Herrmann, M. C.; Mehlhorn, T. A.; Campbell, R. B.

    2008-01-01

    Achieving a high degree of radiation symmetry is a critical feature of target designs for indirect-drive inertial confinement fusion. Typically, the radiation flux incident on the capsule is required to be uniform to 1% or better. It is generally possible to design a hohlraum that provides low values of higher-order asymmetry (Legendre mode P 10 and above) due to geometric averaging effects. Because low-order intrinsic asymmetry (e.g., Legendre modes P 2 and P 4 ) are less strongly reduced by geometric averaging alone, the development of innovative control techniques has been an active area of research in the inertial fusion community over the years. Shields placed inside the hohlraum are one example of a technique that has often been proposed and incorporated into hohlraum target designs. Simple mathematical considerations are presented indicating that radiation shields may be designed to specifically tune lower-order modes (e.g., P 4 ) without deleterious effects on the higher order modes. Two-dimensional view factor and radiation-hydrodynamics simulations confirm these results and support such a path to achieving a highly symmetric x-ray flux. The term ''mode-selective'' is used because these shields, essentially ring structures offset from the capsule, are designed to affect only a specific Legendre mode (or multiple modes) of interest

  4. Inertial Confinement Fusion: steady progress towards ignition and high gain (summary talk)

    International Nuclear Information System (INIS)

    Basko, M.M.

    2005-01-01

    Most important recent advances in inertial confinement fusion (ICF) are highlighted. With the construction of the NIF and LMJ facilities, and a number of improvements in the target design, the conventional indirect-drive approach is making a steady progress towards demonstration of ignition and high gain. The development of the polar direct-drive concept made also the prospects for direct-drive ignition on the NIF extremely favorable. A substantial progress has been reported from the Institute of Laser Engineering in Osaka on exploration of the fast-ignition approach to ICF. Parallel to that, multi-wire Z-pinches have become a competitive driver option for achieving ignition at a lowest possible cost. In heavy ion fusion, experiments have been devoted so far to studying the generation, transport, and final focusing of high-current ion beams. A new concept for a power plant with a heavy-ion driver, based on a cylindrical direct-drive target compressed and ignited (in the fast-ignition mode) by two separate beams of very energetic (E i > or ∼ 0.5 GeV/u) heavy ions, has been proposed. (author)

  5. Electron-detachment cross sections of halogen negative-ion projectiles for inertial confinement fusion

    Science.gov (United States)

    Sant'Anna, M. M.; Zappa, F.; Santos, A. C. F.; de Barros, A. L. F.; Wolff, W.; Coelho, L. F. S.; de Castro Faria, N. V.

    2004-07-01

    Negative-ion beams have recently been suggested as sources of high-energy heavy atoms to be used as drivers for inertial confinement fusion (ICF). Owing to their electron affinities limited to a few eV, anions can be efficiently photo-detached in the vicinity of the fusion chamber, with the resulting high-velocity neutral projectiles following ballistic trajectories towards the hydrogen pellet target. Electron-detachment cross sections are needed as parameters to estimate the beam attenuation in the path from the ion source to the hydrogen pellet. Halogen anions are possible projectile choices. In this paper we present experimental data for total electron-detachment cross sections for F-, Cl-, Br- and I- ions incident on N2, in the 0.94-74 keV u-1 energy range. Our measurements can benchmark theory on anion electron detachment at intermediate to high velocities. Comparison between different projectiles shows very similar collision velocity dependencies. A simple geometrical scaling is presented, providing an estimate for electron-detachment cross sections at the MeV u-1 energy range. The presented scaling indicates that the vacuum requirements due to the use of halogen anions for ICF are less critical than previously suggested.

  6. Achieving competitive excellence in nuclear energy: The threat of proliferation; the challenge of inertial confinement fusion

    International Nuclear Information System (INIS)

    Nuckolls, J.H.

    1994-06-01

    Nuclear energy will have an expanding role in meeting the twenty-first-century challenges of population and economic growth, energy demand, and global warming. These great challenges are non-linearly coupled and incompletely understood. In the complex global system, achieving competitive excellence for nuclear energy is a multi-dimensional challenge. The growth of nuclear energy will be driven by its margin of economic advantage, as well as by threats to energy security and by growing evidence of global warming. At the same time, the deployment of nuclear energy will be inhibited by concerns about nuclear weapons proliferation, nuclear waste and nuclear reactor safety. These drivers and inhibitors are coupled: for example, in the foreseeable future, proliferation in the Middle East may undermine energy security and increase demand for nuclear energy. The Department of Energy's nuclear weapons laboratories are addressing many of these challenges, including nuclear weapons builddown and nonproliferation, nuclear waste storage and burnup, reactor safety and fuel enrichment, global warming, and the long-range development of fusion energy. Today I will focus on two major program areas at the Lawrence Livermore National Laboratory (LLNL): the proliferation of nuclear weapons and the development of inertial confinement fusion (ICF) energy

  7. Progress in laboratory high gain ICF [inertial confinement fusion]: Prospects for the future

    International Nuclear Information System (INIS)

    Storm, E.; Lindl, J.D.; Campbell, E.M.

    1988-01-01

    Inertial confinement fusion (ICF), a thermonuclear reaction in a small (/approximately/5 mm diameter) fuel capsule filled with a few milligrams of deuterium and tritium, has been the subject of very fruitful experimentation since the early 1970's. High gain ICF is now on the threshold of practical applications. With a Laboratory Microfusion Facility (LMF), these applications will have major implications for national defense, basic and applied science, and power production. With a driver capable of delivering about 10 MJ in a 10-ns pulse at an intensity of /approximately/3 /times/ 10 14 W/cm 2 , an appropriately configured cryogenic capsule could be compressed to a density of about 200 g/cm 3 and a temperature of 3--5 keV. Under these conditions, up to 10 mg of DT could be ignited, and with a burn efficiency of about 30%, release up to 1000 MJ of fusion energy, an energy gain of about 100. A thousand megajoules is equivalent to about one quarter ton of TNT, or about 7 gallons of oil--an amount of energy tractable under laboratory conditions and potentially very useful for a variety of applications. 61 refs., 33 figs

  8. Los Alamos National Laboratory progress and path to inertial confinement fusion commercialization

    International Nuclear Information System (INIS)

    Harris, D.B.; Dudziak, D.J.

    1989-01-01

    KrF lasers appear to be an attractive driver for inertial confinement fusion commercial applications such as electric power production. Los Alamos National Laboratory is working to develop the technology required to demonstrate that KrF lasers can satisfy all of the driver requirements. The latest experimental and theoretical results indicate that cost currently appears to be the main issue for KrF lasers. The Los Alamos program is working to reduce the cost of KrF laser systems by developing damage-resistant optical coatings, low-cost optical blanks, high-intrinsic-efficiency gas mixtures, low-cost and high-efficiency pulsed power, and optimized system architectures. Other potential issues may cause problems after the 5 kJ Aurora KrF laser system becomes operational, such as amplified spontaneous emission, cross talk or temporal pulse distortion. Design solutions to issues such as these have been identified and will be experimentally demonstrated on Aurora. Issues specific to commercial-application drivers, such as cost, gas flow, repetively pulsed power, and high reliability cannot be experimentally addressed at this time. Projections will be made on the ability of KrF lasers to satisfy these requirements. The path to commercialization of inertial fusion for KrF lasers is also described. (orig.)

  9. Electron-detachment cross sections of halogen negative-ion projectiles for inertial confinement fusion

    International Nuclear Information System (INIS)

    Sant'Anna, M M; Zappa, F; Santos, A C F; Barros, A L F de; Wolff, W; Coelho, L F S; Faria, N V de Castro

    2004-01-01

    Negative-ion beams have recently been suggested as sources of high-energy heavy atoms to be used as drivers for inertial confinement fusion (ICF). Owing to their electron affinities limited to a few eV, anions can be efficiently photo-detached in the vicinity of the fusion chamber, with the resulting high-velocity neutral projectiles following ballistic trajectories towards the hydrogen pellet target. Electron-detachment cross sections are needed as parameters to estimate the beam attenuation in the path from the ion source to the hydrogen pellet. Halogen anions are possible projectile choices. In this paper we present experimental data for total electron-detachment cross sections for F - , Cl - , Br - and I - ions incident on N 2 , in the 0.94-74 keV u -1 energy range. Our measurements can benchmark theory on anion electron detachment at intermediate to high velocities. Comparison between different projectiles shows very similar collision velocity dependencies. A simple geometrical scaling is presented, providing an estimate for electron-detachment cross sections at the MeV u -1 energy range. The presented scaling indicates that the vacuum requirements due to the use of halogen anions for ICF are less critical than previously suggested

  10. BOOK REVIEW: Inertial confinement fusion: The quest for ignition and energy gain using indirect drive

    Science.gov (United States)

    Yamanaka, C.

    1999-06-01

    Inertial confinement fusion (ICF) is an alternative way to control fusion which is based on scaling down a thermonuclear explosion to a small size, applicable for power production, a kind of thermonuclear internal combustion engine. This book extends many interesting topics concerning the research and development on ICF of the last 25 years. It provides a systematic development of the physics basis and also various experimental data on radiation driven implosion. This is a landmark treatise presented at the right time. It is based on the article ``Development of the indirect-drive approach to inertial confinement fusion and the target physics basis for ignition and gain'' by J.D. Lindl, published in Physics of Plasmas, Vol. 2, November 1995, pp. 3933-4024. As is well known, in the United States of America research on the target physics basis for indirect drive remained largely classified until 1994. The indirect drive approaches were closely related to nuclear weapons research at Lawrence Livermore and Los Alamos National Laboratories. In Japan and other countries, inertial confinement fusion research for civil energy has been successfully performed to achieve DT fuel pellet compression up to 1000 times normal density, and indirect drive concepts, such as the `Cannon Ball' scheme, also prevailed at several international conferences. In these circumstances the international fusion community proposed the Madrid Manifesto in 1988, which urged openness of ICF information to promote international collaboration on civil energy research for the future resources of the human race. This proposal was also supported by some of the US scientists. The United States Department of Energy revised its classification guidelines for ICF six years after the Madrid Manifesto. This first book from the USA treating target physics issues, covering topics from implosion dynamics to hydrodynamic stability, ignition physics, high-gain target design and the scope for energy applications is

  11. Correlation between potential well structure and neutron production in inertial electrostatic confinement fusion

    International Nuclear Information System (INIS)

    Ohnishi, M.; Yamamoto, Y.; Yoshikawa, K.; Sato, K.H.

    1997-01-01

    The electrostatic potential well in inertial electrostatic confinement (IEC) is studied using two approaches. First, the equilibrium potential profile is obtained by solving the charge neutrality condition, i.e. n i n e , assuming the appropriate distribution functions for the ions and the electrons. The formation of a double well structure is demonstrated, with a depth depending upon the ratio between the focus radii of the electrons and the ions. The correlations between the well depth and the volume integrated neutron production due to deuterium-deuterium (DD) reactions are obtained. Second, in order to study the stability of the well, the dynamic behaviours of the potential well are calculated by performing time advancing numerical simulations on the basis of the particle in cell method. Single, double and triple wells, depending on the amount of injected ion current, are observed to be formed for ions with a monoenergetic distribution. The well in the centre of the multiwell structure is unstable and oscillates with a periods much longer than the inverse ion plasma frequency. A double well structure can be formed even for ions with a spread out energy distribution when the ion current is larger than the threshold value. The time averaged neutron production by DD fusion events is proportional to a power of the ion current involved in forming the double well structure. The results strongly suggest that the high neutron production rate should be attributed to not only the well depth but also the unstable behaviour of the potential, i.e. the intermittent peaking of the density in the centre region. A numerical simulation reveals that IEC possesses a favourable dependence of fusion reactions on the injected ion current for the application to a neutron source or a fusion reactor. (author). 9 refs, 9 figs

  12. Advances in inertial confinement fusion at the National Ignition Facility (NIF)

    International Nuclear Information System (INIS)

    Moses, Edward I.

    2010-01-01

    The 192-beam National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in Livermore, CA, is now operational and conducting experiments. NIF, the flagship facility of the U.S. Inertial Confinement Fusion (ICF) Program, will achieve high-energy-density conditions never previously obtained in the laboratory-temperatures over 100 million K, densities of 1000 g/cm 3 , and pressures exceeding 100 billion atmospheres. Such conditions exist naturally only in the interiors of the stars and during thermonuclear burn. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. To date, the NIF laser has demonstrated all pulse shape, beam quality, energy, and other specifications required to meet the ignition challenge. On March 10, 2009, the NIF laser delivered 1.1 MJ of ultraviolet laser energy to target chamber center, approximately 30 times more energy than any previous facility. The ignition program at NIF is the National Ignition Campaign (NIC), a national collaboration for ignition experimentation with participation from General Atomics, LLNL, Los Alamos National Laboratory (LANL), Sandia National Laboratories (SNL), and the University of Rochester Laboratory for Laser Energetics (LLE). The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on fusion as a viable energy option. A particular energy concept under investigation is the LIFE (Laser Inertial Fusion Energy) scheme. The LIFE engine is inherently safe, minimizes proliferation concerns associated with the nuclear fuel cycle, and can provide a sustainable carbon-free energy generation solution in the 21st century. This talk will describe NIF and its potential as a user facility and an experimental platform for high-energy-density science, NIC, and the LIFE approach for clean, sustainable energy.

  13. Advances in Inertial Confinement Fusion at the National Ignition Facility (NIF)

    International Nuclear Information System (INIS)

    Moses, E.

    2009-01-01

    The 192-beam National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in Livermore, CA, is now operational and conducting experiments. NIF, the flagship facility of the U.S. Inertial Confinement Fusion (ICF) Program, will achieve high-energy-density conditions never previously obtained in the laboratory - temperatures over 100 million K, densities of 1,000 g/cm3, and pressures exceeding 100 billion atmospheres. Such conditions exist naturally only in the interiors of the stars and during thermonuclear burn. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. To date, the NIF laser has demonstrated all pulse shape, beam quality, energy, and other specifications required to meet the ignition challenge. On March 10, 2009, the NIF laser delivered 1.1 MJ of ultraviolet laser energy to target chamber center, approximately 30 times more energy than any previous facility. The ignition program at NIF is the National Ignition Campaign (NIC), a national collaboration for ignition experimentation with participation from General Atomics, LLNL, Los Alamos National Laboratory (LANL), Sandia National Laboratories (SNL), and the University of Rochester Laboratory for Laser Energetics (LLE). The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on fusion as a viable energy option. A particular energy concept under investigation is the LIFE (Laser Inertial Fusion Energy) scheme. The LIFE engine is inherently safe, minimizes proliferation concerns associated with the nuclear fuel cycle, and can provide a sustainable carbon-free energy generation solution in the 21st century. This talk will describe NIF and its potential as a user facility and an experimental platform for high-energy-density science, NIC, and the LIFE approach for clean, sustainable energy.

  14. Quantum confined laser devices optical gain and recombination in semiconductors

    CERN Document Server

    Blood, Peter

    2015-01-01

    The semiconductor laser, invented over 50 years ago, has had an enormous impact on the digital technologies that now dominate so many applications in business, commerce and the home. The laser is used in all types of optical fibre communication networks that enable the operation of the internet, e-mail, voice and skype transmission. Approximately one billion are produced each year for a market valued at around $5 billion. Nearly all semiconductor lasers now use extremely thin layers of light emitting materials (quantum well lasers). Increasingly smaller nanostructures are used in the form of quantum dots. The impact of the semiconductor laser is surprising in the light of the complexity of the physical processes that determine the operation of every device. This text takes the reader from the fundamental optical gain and carrier recombination processes in quantum wells and quantum dots, through descriptions of common device structures to an understanding of their operating characteristics. It has a consistent...

  15. Effects of ExB velocity shear and magnetic shear on turbulence and transport in magnetic confinement devices

    International Nuclear Information System (INIS)

    Burrell, K.H.

    1996-11-01

    One of the scientific success stories of fusion research over the past decade is the development of the ExB shear stabilization model to explain the formation of transport barriers in magnetic confinement devices. This model was originally developed to explain the transport barrier formed at the plasma edge in tokamaks after the L (low) to H (high) transition. This concept has the universality needed to explain the edge transport barriers seen in limiter and divertor tokamaks, stellarators, and mirror machines. More recently, this model has been applied to explain the further confinement improvement from H (high)-mode to VH (very high)-mode seen in some tokamaks, where the edge transport barrier becomes wider. Most recently, this paradigm has been applied to the core transport barriers formed in plasmas with negative or low magnetic shear in the plasma core. These examples of confinement improvement are of considerable physical interest; it is not often that a system self-organizes to a higher energy state with reduced turbulence and transport when an additional source of free energy is applied to it. The transport decrease that is associated with ExB velocity shear effects also has significant practical consequences for fusion research. The fundamental physics involved in transport reduction is the effect of ExB shear on the growth, radial extent and phase correlation of turbulent eddies in the plasma. The same fundamental transport reduction process can be operational in various portions of the plasma because there are a number ways to change the radial electric field Er. An important theme in this area is the synergistic effect of ExB velocity shear and magnetic shear. Although the ExB velocity shear appears to have an effect on broader classes of microturbulence, magnetic shear can mitigate some potentially harmful effects of ExB velocity shear and facilitate turbulence stabilization

  16. Advanced laser-backlit grazing-incidence x-ray imaging systems for inertial confinement fusion research. II. Tolerance analysis

    International Nuclear Information System (INIS)

    Bennett, Guy R.; Folta, James A.

    2001-01-01

    Two example ultrahigh-spatial resolution laser-backlit grazing-incidence x-ray microscope designs for inertial confinement fusion (ICF) research have been described [Appl. Opt. 40, 4570 (2001)]. Here details of fabrication, assembly, and optical surface errors that are characteristic of present state-of-the-art superpolished multilayer-coated spherical mirrors are given. They indicate that good image qualities can be expected; in particular, <0.5-μm spatial resolution at very high x-ray energies (up to 25 keV) appears to be feasible. Existing ICF imaging diagnostics approach ∼2 μm spatial at low (<2 keV) energy. The improvement in resolution compared with that of other grazing-incidence devices is attributed to a fortuitous residual on-axis aberration dependence on short wavelengths; recent advances in mirror fabrication, including a new thin-film deposition technique to correct figure errors precisely in one dimension; and novel design. For even higher resolution, a means of creating precise aspherical mirrors of spheric-quality microroughness may be possible by use of the same deposition technique

  17. Intense ion beam research for inertial confinement fusion. Final technical report, 1 October 1981-31 October 1985

    International Nuclear Information System (INIS)

    Hammer, D.A.; Kusse, B.R.; Sudan, R.N.

    1986-01-01

    Theoretical and experimental research has been performed on the application of intense light ion beams to inertial confinement fusion. The following achievements are documented. A 1 TW accelerator (a module of the PBFA 1 device at Sandia National Laboratories, Albuquerque), has been installed at Cornell and it has been used to develop high power magnetically insulated ion diodes. Ion beams at 0.3 TW level have been produced. The use of spectroscopic techniques to diagnose conditions in detail with in magnetically insulated diodes was proposed, and preliminary experiments have been successfully performed. These have revealed the anode plasma density, transverse velocities of ions within the diode (from Doppler broadening of ion emission lines) and the electric field profile in the accelerating gap (from the Stark shifted line profile of especially selected emission lines). Theoretical studies on the effects of lack of symmetry in the electron drift direction on the leakage electron current in a magnetically insulated diode show that even very small perturbations can cause a substantial enhancement of the leakage current. Experiments involving electron flow in a magnetically insulated diode have shown cathode sheath losses to occur in local burst as well as in a smooth manner

  18. Nondestructive Inspection System for Special Nuclear Material Using Inertial Electrostatic Confinement Fusion Neutrons and Laser Compton Scattering Gamma-Rays

    Science.gov (United States)

    Ohgaki, H.; Daito, I.; Zen, H.; Kii, T.; Masuda, K.; Misawa, T.; Hajima, R.; Hayakawa, T.; Shizuma, T.; Kando, M.; Fujimoto, S.

    2017-07-01

    A Neutron/Gamma-ray combined inspection system for hidden special nuclear materials (SNMs) in cargo containers has been developed under a program of Japan Science and Technology Agency in Japan. This inspection system consists of an active neutron-detection system for fast screening and a laser Compton backscattering gamma-ray source in coupling with nuclear resonance fluorescence (NRF) method for precise inspection. The inertial electrostatic confinement fusion device has been adopted as a neutron source and two neutron-detection methods, delayed neutron noise analysis method and high-energy neutron-detection method, have been developed to realize the fast screening system. The prototype system has been constructed and tested in the Reactor Research Institute, Kyoto University. For the generation of the laser Compton backscattering gamma-ray beam, a race track microtron accelerator has been used to reduce the size of the system. For the NRF measurement, an array of LaBr3(Ce) scintillation detectors has been adopted to realize a low-cost detection system. The prototype of the gamma-ray system has been demonstrated in the Kansai Photon Science Institute, National Institutes for Quantum and Radiological Science and Technology. By using numerical simulations based on the data taken from these prototype systems and the inspection-flow, the system designed by this program can detect 1 kg of highly enriched 235U (HEU) hidden in an empty 20-ft container within several minutes.

  19. 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

  20. Effect of loss cone on confinement in toroidal helical device

    International Nuclear Information System (INIS)

    Itoh, K.; Itoh, S.-I.; Fukuyama, A.; Hanatani, K.

    1988-12-01

    Analytical estimation is given on the loss cone in the toroidal helical devices in the presence of the radial electric field and the modulation of the helical ripple. The minimum energy of particles entering the loss cone is calculated. The modulation is not always effective in reducing the loss in the presence of the radial electric field. The plasma loss due to the loss cone is estimated in the collisionless limit. The radial electric field is estimated in the presence of the loss cone. It is found that the transition to the solution with positive radial electric field, which is necessary to achieve the high-ion-temperature mode, becomes difficult. This difficulty is large for the systems with the small helical ripple. (author)

  1. Exploitation of a Breakthrough in Magnetic Confinement Fusion to Improve Transuranic Incineration

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Erich [Nuclear and Radiation Engineering Program, The University of Texas at Austin, Austin, TX 78712 (United States); Kotschenreuther, Mike; Mahajan, Swadesh; Valanju, Prashant [Institute for Fusion Studies, The University of Texas at Austin, Austin, TX 78712 (United States)

    2009-06-15

    A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal non-fissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The centerpiece of this fuel cycle, the high power density compact fusion neutron source (CFNS, 100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is about an order of magnitude below the corresponding number of critical fast spectrum reactors (FR) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades. The generic Hybrid, combining neutron-rich fusion with energy-rich fission, was first conceptualized several decades ago. However, it is only now that accumulated advances in fusion science and technology allow designing a neutron source like CFNS that is simultaneously compact and high power density, offering a neutron source an order of magnitude stronger than that obtained from accelerator driven systems. The former is essential for efficient coupling to the fission blanket, and the latter is key to efficient neutron production necessary to yield high neutron fluxes needed for effective transmutation. The recent invention of the SuperX-Divertor (SXD)1, a new magnetic configuration that allows the system to safely exhaust large heat and particle fluxes peculiar to CFNS-like devices, is a crucial addition to the underlying knowledge base. The subcritical FFTS acquires a definite advantage over the critical FR approach because of its ability to support an innovative fuel cycle that makes the cheaper LWR do the bulk (75%) of the transuranic transmutation via deep burn in an inert matrix fuel

  2. Development of new low activation aluminum alloys for fusion devices

    International Nuclear Information System (INIS)

    Kamada, Kohji; Kakihana, Hidetake.

    1985-01-01

    As the materials for the R facility (a tokamak nuclear fusion device in the R project intended for D-T burning) in the Institute of Plasma Physics, Nagoya University, Al-4 % Mg-0.2 % Bi (5083 improved type) and Al-4 % Mg-1 % Li, aimed at low radioactivability, high electric resistance and high strength, have been developed. The results of the nuclear properties evaluation with 14 MeV neutrons and of the measurements of electric resistance and mechanical properties were satisfactory. The possibility of producing large Al-4 % Mg-1 % Li plate (1 m x 2 m x 25 mm) in the existing factory was confirmed, with the properties retained. The electric resistances were higher than those in the conventional aluminum alloys, and still with feasibility for the further improvement. General properties of the fusion aluminum alloys and the 26 Al formation in (n, 2n) reaction were studied. (Mori, K.)

  3. 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

  4. 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

  5. Railgun pellet injection system for fusion experimental devices

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Oda, Y. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Azuma, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Satake, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Kasai, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan); Hasegawa, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan)

    1995-11-01

    A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s{sup -1} using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s{sup -1} using a 3 m long railgun. (orig.).

  6. Railgun pellet injection system for fusion experimental devices

    International Nuclear Information System (INIS)

    Onozuka, M.; Hasegawa, K.

    1995-01-01

    A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s -1 using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s -1 using a 3 m long railgun. (orig.)

  7. The Roles and Developments needed for Diagnostics in the ITER Fusion Device

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, Michael [ITER Organization, Route de Vinon-sur-Verdon - CS 90046, 13067 St Paul-lez-Durance Cedex (France)

    2015-07-01

    Harnessing the power from Fusion on earth is an important and challenging task. Excellent work has been carried out in this area over the years with several demonstrations of the ability to produce power. Now, a new large device is being constructed in the south of France. This is called ITER. ITER is a large-scale scientific experiment that aims to demonstrate a possibility to produce commercial energy from fusion. This project is now well underway with the many teams working on the construction and completing various aspects of the design. This device will carry up to 15 MA of plasma current and produce about 500 MW of power, 400 MW approximately in high energy neutrons. The typical temperatures of the electrons inside this device are in the region of a few hundred million Kelvin. It is maintained using a magnetic field. This device is pushing several boundaries from those currently existing. As a result of this, several technologies need to be developed or extended. This is especially true for the systems or diagnostics that measure the performance and provide the control signals for this device. A diagnostic set will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include amongst others, measurements of the plasma shape, temperature, density, impurity concentration, and particle and energy confinement times. The system will comprise about 45 individual measuring systems drawn from the full range of modern plasma diagnostic techniques, including magnetics, lasers, X-rays, neutron cameras, impurity monitors, particle spectrometers, radiation bolometers, pressure and gas analysis, and optical fibres. These devices will have to be made to work in the new and challenging environment inside the vacuum vessel. These systems will have to cope with a range of phenomena that extend the current knowledge in the Fusion field. One

  8. Device for plasma confinement and heating by high currents and nonclassical plasma transport properties

    Science.gov (United States)

    Coppi, B.; Montgomery, D.B.

    1973-12-11

    A toroidal plasma containment device having means for inducing high total plasma currents and current densities and at the same time emhanced plasma heating, strong magnetic confinement, high energy density containment, magnetic modulation, microwaveinduced heating, and diagnostic accessibility is described. (Official Gazette)

  9. Maximizing 1D “like” implosion performance for inertial confinement fusion science

    Energy Technology Data Exchange (ETDEWEB)

    Kline, John L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-07-15

    While the march towards achieving indirectly driven inertial confinement fusion at the NIF has made great progress, the experiments show that multi-dimensional effects still dominate the implosion performance. Low mode implosion symmetry and hydrodynamic instabilities seed by capsule mounting features appear to be two key limiting factors for implosion performance. One reason these factors have a large impact on the performance of ICF implosions is the high convergence required to achieve high fusion gains. To tackle these problems, a predictable implosion platform is needed meaning experiments must trade-off high gain for performance. To this end, LANL has adopted three main approaches to develop a 1D implosion platform where 1D means high yield over 1D clean calculations. Taking advantage of the properties of beryllium capsules, a high adiabat, low convergence platform is being developed. The higher drive efficiency for beryllium enables larger case-to-capsule ratios to improve symmetry at the expense of drive. Smaller capsules with a high adiabat drive are expected to reduce the convergence and thus increase predictability. The second approach is liquid fuel layers using wetted foam targets. With liquid fuel layers, the initial mass in the hot spot can be controlled via the target fielding temperature which changes the liquid vapor pressure. Varying the initial hot spot mass via the vapor pressure controls the implosion convergence and minimizes the need to vaporize the dense fuel layer during the implosion to achieve ignition relevant hot spot densities. The last method is double shell targets. Unlike hot spot ignition, double shells ignite volumetrically. The inner shell houses the DT fuel and the convergence of this cavity is relatively small compared to hot spot ignition. Radiation trapping and the longer confinement times relax the conditions required to ignite the fuel. Key challenges for double shell targets are coupling the momentum of the outer shell to

  10. HYPERFUSE: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1980-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with each other or a target block in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs or 90 Sr. The 14-MeV fusion neutrons released during the pellet burn cause transmutation reactions (e.g., (n, 2n), (n, α), etc.) that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  11. Hyper fuse: a novel inertial confinement system utilizing hypervelocity projectiles for fusion energy production and fission waste transmutation

    International Nuclear Information System (INIS)

    Makowitz, H.; Powell, J.R.; Wiswall, R.

    1979-01-01

    Parametric system studies of an inertial confinement fusion (ICF) reactor system to transmute fission products from an LWR economy have been carried out. The ICF reactors would produce net power in addition to transmuting fission products. The particular ICF concept examined is an impact fusion approach termed HYPERFUSE, in which hypervelocity pellets, traveling on the order of 100 to 300 km/sec, collide with a target in a reactor chamber and initiate a thermonuclear reaction. The DT fusion fuel is contained in a shell of the material to be transmuted, e.g., 137 Cs or 90 Sr. The 14 MeV fusion neutrons released during the pellet burn cause transmutation reactions [e.g., (n, 2n), (n, α), etc.] that convert the long lived fission products (FP's) either to stable products or to species that decay with a short half-life to a stable product

  12. Theoretical and simulation research of hydrodynamic instabilities in inertial-confinement fusion implosions

    Science.gov (United States)

    Wang, LiFeng; Ye, WenHua; He, XianTu; Wu, JunFeng; Fan, ZhengFeng; Xue, Chuang; Guo, HongYu; Miao, WenYong; Yuan, YongTeng; Dong, JiaQin; Jia, Guo; Zhang, Jing; Li, YingJun; Liu, Jie; Wang, Min; Ding, YongKun; Zhang, WeiYan

    2017-05-01

    Inertial fusion energy (IFE) has been considered a promising, nearly inexhaustible source of sustainable carbon-free power for the world's energy future. It has long been recognized that the control of hydrodynamic instabilities is of critical importance for ignition and high-gain in the inertial-confinement fusion (ICF) hot-spot ignition scheme. In this mini-review, we summarize the progress of theoretical and simulation research of hydrodynamic instabilities in the ICF central hot-spot implosion in our group over the past decade. In order to obtain sufficient understanding of the growth of hydrodynamic instabilities in ICF, we first decompose the problem into different stages according to the implosion physics processes. The decomposed essential physics pro- cesses that are associated with ICF implosions, such as Rayleigh-Taylor instability (RTI), Richtmyer-Meshkov instability (RMI), Kelvin-Helmholtz instability (KHI), convergent geometry effects, as well as perturbation feed-through are reviewed. Analyti- cal models in planar, cylindrical, and spherical geometries have been established to study different physical aspects, including density-gradient, interface-coupling, geometry, and convergent effects. The influence of ablation in the presence of preheating on the RTI has been extensively studied by numerical simulations. The KHI considering the ablation effect has been discussed in detail for the first time. A series of single-mode ablative RTI experiments has been performed on the Shenguang-II laser facility. The theoretical and simulation research provides us the physical insights of linear and weakly nonlinear growths, and nonlinear evolutions of the hydrodynamic instabilities in ICF implosions, which has directly supported the research of ICF ignition target design. The ICF hot-spot ignition implosion design that uses several controlling features, based on our current understanding of hydrodynamic instabilities, to address shell implosion stability, has

  13. Dynamic symmetry of indirectly driven inertial confinement fusion capsules on the National Ignition Facility

    International Nuclear Information System (INIS)

    Town, R. P. J.; Bradley, D. K.; Kritcher, A.; Jones, O. S.; Rygg, J. R.; Tommasini, R.; Barrios, M.; Benedetti, L. R.; Berzak Hopkins, L. F.; Celliers, P. M.; Döppner, T.; Dewald, E. L.; Eder, D. C.; Field, J. E.; Glenn, S. M.; Izumi, N.; Haan, S. W.; Khan, S. F.; Ma, T.; Milovich, J. L.

    2014-01-01

    In order to achieve ignition using inertial confinement fusion it is important to control the growth of low-mode asymmetries as the capsule is compressed. Understanding the time-dependent evolution of the shape of the hot spot and surrounding fuel layer is crucial to optimizing implosion performance. A design and experimental campaign to examine sources of asymmetry and to quantify symmetry throughout the implosion has been developed and executed on the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)]. We have constructed a large simulation database of asymmetries applied during different time intervals. Analysis of the database has shown the need to measure and control the hot-spot shape, areal density distribution, and symmetry swings during the implosion. The shape of the hot spot during final stagnation is measured using time-resolved imaging of the self-emission, and information on the shape of the fuel at stagnation can be obtained from Compton radiography [R. Tommasini et al., Phys. Plasmas 18, 056309 (2011)]. For the first time on NIF, two-dimensional inflight radiographs of gas-filled and cryogenic fuel layered capsules have been measured to infer the symmetry of the radiation drive on the capsule. These results have been used to modify the hohlraum geometry and the wavelength tuning to improve the inflight implosion symmetry. We have also expanded our shock timing capabilities by the addition of extra mirrors inside the re-entrant cone to allow the simultaneous measurement of shock symmetry in three locations on a single shot, providing asymmetry information up to Legendre mode 4. By diagnosing the shape at nearly every step of the implosion, we estimate that shape has typically reduced fusion yield by about 50% in ignition experiments

  14. Dynamic response of materials on subnanosecond time scales, and beryllium properties for inertial confinement fusion

    International Nuclear Information System (INIS)

    Swift, Damian C.; Tierney, Thomas E.; Luo Shengnian; Paisley, Dennis L.; Kyrala, George A.; Hauer, Allan; Greenfield, Scott R.; Koskelo, Aaron C.; McClellan, Kenneth J.; Lorenzana, Hector E.; Kalantar, Daniel; Remington, Bruce A.; Peralta, Pedro; Loomis, Eric

    2005-01-01

    During the past few years, substantial progress has been made in developing experimental techniques capable of investigating the response of materials to dynamic loading on nanosecond time scales and shorter, with multiple diagnostics probing different aspects of the behavior. These relatively short time scales are scientifically interesting because plastic flow and phase changes in common materials with simple crystal structures--such as iron--may be suppressed, allowing unusual states to be induced and the dynamics of plasticity and polymorphism to be explored. Loading by laser-induced ablation can be particularly convenient: this technique has been used to impart shocks and isentropic compression waves from ∼1 to 200 GPa in a range of elements and alloys, with diagnostics including line imaging surface velocimetry, surface displacement (framed area imaging), x-ray diffraction (single crystal and polycrystal), ellipsometry, and Raman spectroscopy. A major motivation has been the study of the properties of beryllium under conditions relevant to the fuel capsule in inertial confinement fusion: magnetically driven shock and isentropic compression shots at Z were used to investigate the equation of state and shock melting characteristics, complemented by laser ablation experiments to investigate plasticity and heterogeneous response from the polycrystalline microstructure. These results will help to constrain acceptable tolerances on manufacturing, and possible loading paths, for inertial fusion ignition experiments at the National Ignition Facility. Laser-based techniques are being developed further for future material dynamics experiments, where it should be possible to obtain high quality data on strength and phase changes up to at least 1 TPa

  15. The potential role of electric fields and plasma barodiffusion on the inertial confinement fusion databasea)

    Science.gov (United States)

    Amendt, Peter; Wilks, S. C.; Bellei, C.; Li, C. K.; Petrasso, R. D.

    2011-05-01

    The generation of strong, self-generated electric fields (GV/m) in direct-drive, inertial-confinement-fusion (ICF) capsules has been reported [Rygg et al., Science 319, 1223 (2008); Li et al., Phys. Rev. Lett. 100, 225001 (2008)]. A candidate explanation for the origin of these fields based on charge separation across a plasma shock front was recently proposed [Amendt et al., Plasma Phys. Controlled Fusion 51 124048 (2009)]. The question arises whether such electric fields in imploding capsules can have observable consequences on target performance. Two well-known anomalies come to mind: (1) an observed ≈2× greater-than-expected deficit of neutrons in an equimolar D3He fuel mixture compared with hydrodynamically equivalent D [Rygg et al., Phys. Plasmas 13, 052702 (2006)] and DT [Herrmann et al., Phys. Plasmas 16, 056312 (2009)] fuels, and (2) a similar shortfall of neutrons when trace amounts of argon are mixed with D in indirect-drive implosions [Lindl et al., Phys. Plasmas 11, 339 (2004)]. A new mechanism based on barodiffusion (or pressure gradient-driven diffusion) in a plasma is proposed that incorporates the presence of shock-generated electric fields to explain the reported anomalies. For implosions performed at the Omega laser facility [Boehly et al., Opt. Commun. 133, 495 (1997)], the (low Mach number) return shock has an appreciable scale length over which the lighter D ions can diffuse away from fuel center. The depletion of D fuel is estimated and found to lead to a corresponding reduction in neutrons, consistent with the anomalies observed in experiments for both argon-doped D fuels and D3He equimolar mixtures. The reverse diffusional flux of the heavier ions toward fuel center also increases the pressure from a concomitant increase in electron number density, resulting in lower stagnation pressures and larger imploded cores in agreement with gated, self-emission, x-ray imaging data.

  16. Single and nested tungsten-wire-array dynamics and applications to inertial confinement fusion

    Science.gov (United States)

    Cuneo, Michael

    2005-10-01

    Wire array z-pinches show great promise as x-ray sources for indirect-drive inertial confinement fusion (ICF). The double z-pinch hohlraum, for example, has produced capsule radiation drive symmetric to within 3%. This ICF concept will require that each of two 20-mm-diam arrays scale to x-ray powers ˜1 PW, to drive high-yield (>0.2 GJ) capsules to ignition. High-yield fusion will also require a temporally shaped radiation pulse to drive a low-entropy capsule implosion. Recently, improved understanding of high current (11-19 MA) single and nested wire-array dynamics has enabled significant progress towards these goals. As at lower currents, a single wire array (and both the outer and inner arrays of a nested system) shows a wire ablation phase, axial modulation of the ablation rate, a larger ablation rate for larger diameter wires, trailing mass and trailing current. These processes and others produce a broad mass profile that may impact the optimization of x-ray output for single and nested arrays. Our new insights into wire array physics have led to 20-mm-diam single and nested arrays with peak powers of 150-190 TW at implosion times of 55-90 ns, increased from 60-120 TW at 95-110 ns, improving power scaling. Radiation pulse shapes required for 3-shock isentropic compression of high-yield ICF capsules have also been demonstrated with nested wire arrays operating in current-transfer mode. In collaboration with: D.B. Sinars, R.A. Vesey, E.M. Waisman, W.A. Stygar, D.E. Bliss, S.V. Lebedev, J.P. Chittenden, P.V. Sasorov, R.W. Lemke, E.P. Yu, B.B. Afeyan, G.R. Bennett, M.G. Mazarakis, M.R. Lopez, M.E. Savage, J.L. Porter, T.A. Mehlhorn.

  17. Multilayer mirror based monitors for impurity controls in large fusion reactor type devices

    International Nuclear Information System (INIS)

    Regan, S.P.; May, M.J.; Soukhanovskii, V.; Finkenthal, M.; Moos, H.W.

    1995-01-01

    Multilayer Mirror (MLM) based monitors are compact, high throughput diagnostics capable of extracting XUV emissions (the wavelength range including the soft-x-ray and the extreme ultraviolet, 10 angstrom to 304 angstrom) of impurities from the harsh environment of large fusion reactor type devices. For several years the Plasma Spectroscopy Group at Johns Hopkins University has investigated the application of MLM based XUV spectroscopic diagnostics for magnetically confined fusion plasmas. MLM based monitors have been constructed for and extensively used on DIII-D, Alcator C-mod, TEXT, Phaedrus-T, and CDX-U tokamaks to study the impurity behavior of elements ranging from He to Mo. On ITER MLM based devices would be used to monitor the spectral line emissions from Li I-like to F I-like charge states of Fe, Cr, and Ni, as well as extractors for the bands of emissions from high Z elements such as Mo or W for impurity controls of the fusion plasma. In addition to monitoring the impurity emissions from the main plasma, MLM based devices can also be adapted for radiation measurements of low Z elements in the divertor. The concepts and designs of these MLM based monitors for impurity controls in ITER will be presented. The results of neutron irradiation experiments of the MLMs performed in the Los Alamos Spallation Radiation Effects Facility (LASREF) at the Los Alamos National Laboratory will also be discussed. These preliminary neutron exposure studies show that the dispersive and reflective qualities of the MLMs were not affected in a significant manner

  18. High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

    Science.gov (United States)

    Neu, R.

    2006-04-01

    In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures.

  19. High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

    International Nuclear Information System (INIS)

    Neu, R.

    2006-01-01

    In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures

  20. A study of hydrogen isotopes fuel control by wall effect in magnetic fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Motevalli, S.M., E-mail: motavali@umz.ac.ir; Safari, M.

    2016-11-15

    Highlights: • A particle balance model for the main plasma and wall inventory in magnetic fusion device has been represented. • The dependence of incident particles energy on the wall has been considered in 10–300 eV for the sputtering yield and recycling coefficient. • The effect of fueling methods on plasma density behavior has been studied. - Abstract: Determination of plasma density behavior in magnetic confinement system needs to study the plasma materials interaction in the facing components such as first wall, limiter and divertor. Recycling of hydrogen isotope is an effective parameter in plasma density rate and plasma fueling. Recycling coefficient over the long pulse operation, gets to the unity, so it has a significant effect on steady state in magnetic fusion devices. Typically, sputtered carbon atoms from the plasma facing components form hydrocarbons and they redeposit on the wall. In this case little rate of hydrogen loss occurs. In present work a zero dimensional particle equilibrium model has been represented to determine particles density rate in main plasma and wall inventory under recycling effect and codeposition of hydrogen in case of continues and discontinues fueling methods and effective parameters on the main plasma decay has been studied.

  1. Superconducting (radiation hardened) magnets for mirror fusion devices

    International Nuclear Information System (INIS)

    Henning, C.D.; Dalder, E.N.C.; Miller, J.R.; Perkins, J.R.

    1983-01-01

    Superconducting magnets for mirror fusion have evolved considerably since the Baseball II magnet in 1970. Recently, the Mirror Fusion Test Facility (MFTF-B) yin-yang has been tested to a full field of 7.7 T with radial dimensions representative of a full scale reactor. Now the emphasis has turned to the manufacture of very high field solenoids (choke coils) that are placed between the tandem mirror central cell and the yin-yang anchor-plug set. For MFTF-B the choke coil field reaches 12 T, while in future devices like the MFTF-Upgrade, Fusion Power Demonstration and Mirror Advanced Reactor Study (MARS) reactor the fields are doubled. Besides developing high fields, the magnets must be radiation hardened. Otherwise, thick neutron shields increase the magnet size to an unacceptable weight and cost. Neutron fluences in superconducting magnets must be increased by an order of magnitude or more. Insulators must withstand 10 10 to 10 11 rads, while magnet stability must be retained after the copper has been exposed to fluence above 10 19 neutrons/cm 2

  2. Development and application of charcoal sorbents for cryopumping fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Sedgley, D.W. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)

    1989-06-01

    Progress has been made in defining the capabilities of charcoal as the most promising absorbent to be used in cryopumps for fusion power application. The capabilities of alternative methods of cryopumping helium have been examined in a literature survey and by test, and the results are described here. Considerations include pumping speed, capacity to accumulate pumped gas, ease of reconditioning, use of alternative materials and tolerance to the fusion environment. Vacuum pumps for future fusion devices must handle large quantities of helium/hydrogen isotopes and other impurities. Cryopumps or turbomolecular pumps have demonstrated the capability on a small scale, and each has an important advantage: TMPs do not accumulate gases; cryopumps can separate helium from other effluents. This paper includes a review of a method for selecting charcoals for helium cryopumping, testing of a continuously operating cryopump system, and definition of a design that is based on the requirements of the Next European Torus. Tritium limits are satisfied. The pump design incorporates the charcoal sorbent system that has been recently developed and is based on a reasonable extrapolation of current state-of-the-art. Evaluation of alternative methods of separating helium and other gases led to selection of a movable barrier as the preferred solution. (orig.).

  3. Optical Comb Generation for Streak Camera Calibration for Inertial Confinement Fusion Experiments

    International Nuclear Information System (INIS)

    Ronald Justin; Terence Davies; Frans Janson; Bruce Marshall; Perry Bell; Daniel Kalantar; Joseph Kimbrough; Stephen Vernon; Oliver Sweningsen

    2008-01-01

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is coming on-line to support physics experimentation for the U.S. Department of Energy (DOE) programs in Inertial Confinement Fusion (ICF) and Stockpile Stewardship (SS). Optical streak cameras are an integral part of the experimental diagnostics instrumentation at NIF. To accurately reduce streak camera data a highly accurate temporal calibration is required. This article describes a technique for simultaneously generating a precise +/- 2 ps optical marker pulse (fiducial reference) and trains of precisely timed, short-duration optical pulses (so-called 'comb' pulse trains) that are suitable for the timing calibrations. These optical pulse generators are used with the LLNL optical streak cameras. They are small, portable light sources that, in the comb mode, produce a series of temporally short, uniformly spaced optical pulses, using a laser diode source. Comb generators have been produced with pulse-train repetition rates up to 10 GHz at 780 nm, and somewhat lower frequencies at 664 nm. Individual pulses can be as short as 25-ps FWHM. Signal output is via a fiber-optic connector on the front panel of the generator box. The optical signal is transported from comb generator to streak camera through multi-mode, graded-index optical fiber

  4. Investigation of natural frequencies of laser inertial confinement fusion capsules using resonant ultrasound spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Xiaojun [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Tang, Xing; Wang, Zongwei [Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Chen, Qian; Qian, Menglu [Institute of Acoustic, Tongji University, Shanghai 200433 (China); Meng, Jie [Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Tang, Yongjian [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Research Center of Laser Fusion, CAEP, Mianyang 621900 (China); Zou, Yaming; Shen, Hao [Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Gao, Dangzhong, E-mail: dgaocn@163.com [Research Center of Laser Fusion, CAEP, Mianyang 621900 (China)

    2017-01-15

    Highlights: • The frequency equation of isotropic multi-layer hollow spheres was derived using three-dimension (3D) elasticity theory and transfer matrix method. • The natural frequencies of the capsules with a millimeter-sized diameter are determined experimentally using resonant ultrasound spectrum (RUS) system. • The predicted natural frequencies of the frequency equation accord well with the observed results. • The theoretical and experimental investigation has proved the potential applicability of RUS to both metallic and non-metallic capsules. - Abstract: The natural frequency problem of laser inertial confinement fusion (ICF) capsules is one of the basic problems for determining non-destructively the elasticity modulus of each layer material using resonant ultrasound spectroscopy (RUS). In this paper, the frequency equation of isotropic one-layer hollow spheres was derived using three dimension (3D) elasticity theory and some simplified frequency equations were discussed under axisymmetric and spherical symmetry conditions. The corresponding equation of isotropic multi-layer hollow spheres was given employing transfer matrix method. To confirm the validity of the frequency equation and explore the feasibility of RUS for characterizing the ICF capsules, three representative capsules with a millimeter-sized diameter were determined by piezoelectric-based resonant ultrasound spectroscopy (PZT-RUS) and laser-based resonant ultrasound spectroscopy (LRUS) techniques. On the basis of both theoretical and experimental results, it is proved that the calculated and measured natural frequencies are accurate enough for determining the ICF capsules.

  5. New issues and direction in the U.S. Inertial Confinement Fusion Program

    International Nuclear Information System (INIS)

    Sluyter, M.M.

    1995-01-01

    In pursuit of its goal of developing a laboratory microfusion capability for both defense and other applications, the US Inertial Confinement Fusion (ICF) Program has developed a set of laser and light ion research facilities which, with their supporting components, represent the most advanced set of high energy density physics research capabilities in the world. The US ICF Program is considering both direct and indirect drive as a path to ignition. In the former, a deuterium-tritium (DT) filled spherical capsule is directly imploded using laser light, while in the latter driver (laser or light ion) energy is converted to x-rays which are then used to implode a capsule. The latter results in greater drive symmetry and reduced hydrodynamic instability at a cost in efficiency. The Program's major direct drive laser facilities are the NIKE KrF laser at the Naval Research Laboratory (NRL) and the Nd-glass OMEGA Upgrade laser at the University of Rochester. The 40-kJ Nova laser at the Lawrence Livermore National Laboratory (LLNL), completed in 1984, is the Program's major indirect drive facility. In addition to their Program related work, a number of these facilities are now available to outside users through a proposal submission process. The next few years are expected to be very exciting for ICF as experiments commence on OMEGA Upgrade and NIKE and efforts continue to construct the National Ignition Facility (NIF) a large glass laser which is designed to demonstrate ignition in the laboratory

  6. Effects of electron-ion temperature equilibration on inertial confinement fusion implosions.

    Science.gov (United States)

    Xu, Barry; Hu, S X

    2011-07-01

    The electron-ion temperature relaxation essentially affects both the laser absorption in coronal plasmas and the hot-spot formation in inertial confinement fusion (ICF). It has recently been reexamined for plasma conditions closely relevant to ICF implosions using either classical molecular-dynamics simulations or analytical methods. To explore the electron-ion temperature equilibration effects on ICF implosion performance, we have examined two Coulomb logarithm models by implementing them into our hydrocodes, and we have carried out hydrosimulations for ICF implosions. Compared to the Lee-More model that is currently used in our standard hydrocodes, the two models predict substantial differences in laser absorption, coronal temperatures, and neutron yields for ICF implosions at the OMEGA Laser Facility [Boehly et al. Opt. Commun. 133, 495 (1997)]. Such effects on the triple-picket direct-drive design at the National Ignition Facility (NIF) have also been explored. Based on the validity of the two models, we have proposed a combined model of the electron-ion temperature-relaxation rate for the overall ICF plasma conditions. The hydrosimulations using the combined model for OMEGA implosions have shown ∼6% more laser absorption, ∼6%-15% higher coronal temperatures, and ∼10% more neutron yield, when compared to the Lee-More model prediction. It is also noticed that the gain for the NIF direct-drive design can be varied by ∼10% among the different electron-ion temperature-relaxation models.

  7. A measurable Lawson criterion and hydro-equivalent curves for inertial confinement fusion

    International Nuclear Information System (INIS)

    Zhou, C. D.; Betti, R.

    2008-01-01

    It is shown that the ignition condition (Lawson criterion) for inertial confinement fusion (ICF) can be cast in a form dependent on the only two parameters of the compressed fuel assembly that can be measured with existing techniques: the hot spot ion temperature (T i h ) and the total areal density (ρR tot ), which includes the cold shell contribution. A marginal ignition curve is derived in the ρR tot , T i h plane and current implosion experiments are compared with the ignition curve. On this plane, hydrodynamic equivalent curves show how a given implosion would perform with respect to the ignition condition when scaled up in the laser-driver energy. For 3 i h > n i h > n 2.6 · tot > n >50 keV 2.6 · g/cm 2 , where tot > n and i h > n are the burn-averaged total areal density and hot spot ion temperature, respectively. Both quantities are calculated without accounting for the alpha-particle energy deposition. Such a criterion can be used to determine how surrogate D 2 and subignited DT target implosions perform with respect to the one-dimensional ignition threshold.

  8. Fast-shock ignition: a new approach to inertial confinement fusion

    Directory of Open Access Journals (Sweden)

    AH Farahbod

    2013-03-01

    Full Text Available  A new concept for inertial confinement fusion called fast-shock ignition (FSI is introduced as a credible scheme in order to obtain high target gain. In the proposed model, the separation of fuel ignition into two successive steps, under the suitable conditions, reduces required ignitor energy for the fuel ignition. The main procedure in FSI concept is compressing the fuel up to stagnation. Then, two high intensity short pulse laser spikes with energy and power lower than those required for shock ignition (SI and fast ignition (FI with a proper delay time are launched at the fuel which increases the central hot-spot temperature and completes the ignition of the precompressed fuel. The introduced semi-analytical model indicates that with fast-shock ignition, the total required energy for compressing and igniting the fuel can be slightly reduced in comparison to pure shock ignition. Furthermore, for fuel mass greater than , the target energy gain increases up to 15 percent and the contribution of fast ignitor under the proper conditions could be decreased about 20 percent compared with pure fast ignition. The FSI scheme is beneficial from technological considerations for the construction of short pulse high power laser drivers. The general advantages of fast-shock ignition over pure shock ignition in terms of figure of merit can be more than 1.3.

  9. Picosecond imaging of inertial confinement fusion plasmas using electron pulse-dilation

    Science.gov (United States)

    Hilsabeck, T. J.; Nagel, S. R.; Hares, J. D.; Kilkenny, J. D.; Bell, P. M.; Bradley, D. K.; Dymoke-Bradshaw, A. K. L.; Piston, K.; Chung, T. M.

    2017-02-01

    Laser driven inertial confinement fusion (ICF) plasmas typically have burn durations on the order of 100 ps. Time resolved imaging of the x-ray self emission during the hot spot formation is an important diagnostic tool which gives information on implosion symmetry, transient features and stagnation time. Traditional x-ray gated imagers for ICF use microchannel plate detectors to obtain gate widths of 40-100 ps. The development of electron pulse-dilation imaging has enabled a 10X improvement in temporal resolution over legacy instruments. In this technique, the incoming x-ray image is converted to electrons at a photocathode. The electrons are accelerated with a time-varying potential that leads to temporal expansion as the electron signal transits the tube. This expanded signal is recorded with a gated detector and the effective temporal resolution of the composite system can be as low as several picoseconds. An instrument based on this principle, known as the Dilation X-ray Imager (DIXI) has been constructed and fielded at the National Ignition Facility. Design features and experimental results from DIXI will be presented.

  10. Reactor potential of the Magnetically Insulated Inertial Confinement Fusion (MICF) system

    International Nuclear Information System (INIS)

    Kammash, T.; Galbraith, D.L.

    1987-01-01

    In this paper a quasi one dimensional, time dependent set of particle and energy balance equations for the thermal species, namely, electrons, ions and thermal alphas which also allows for an appropriate set of fast alpha groups is utilized to assess the reactor prospects of a DT-burning Magnetically Insulated Inertial Confinement Fusion (MICF) system. A reference reactor consisting of an initial plasma with density of 10 21 cm -3 , temperature of keV, a radius of 0.25 cm is shown to ignite and yield an energy multiplication factor ''Q'' of about 60 when the plasma is allowed to burn for 2 microseconds. When the burntime is extended to 9 microseconds for the same initial conditions our calculations show that Q almost doubles just before the final radius becomes equal to the inner radius of the shell. These preliminary results seem to indicate that MICF does indeed have the potential for a reactor although some relevant physics issues need to be addressed first. 42 refs., 6 figs

  11. Studies on the feasibility of heavy ion beams for inertial confinement fusion

    International Nuclear Information System (INIS)

    1983-05-01

    This Annual Report summarizes the scientific results of work carried out in 1982 in the framework of a feasibility study for inertial confinement fusion with heavy ion beams funded by the German Ministry of Research and Technology. The principal aim of this basic research program is the investigation of key problems and the identification of critical issues of the heavy ion ICF concept in the fields of accelerator research, atomic physics, target physics, and reactor design. The research is carried out by about ten working groups at various research centers and universities. One of the highlights in 1982 was a symposium held end of March at GSI which focussed on a critical analysis of the HIBALL accelerator concept. Whereas technical issues and hardware parameters were found feasible the beam dynamics in the storage rings turned out to be beyond the so far believed stability limits. As a consequence a revised accelerator scenario based on a lower charge state and a higher linac current has been investigated during the last year. First considerations were made on an experimental facility necessary for the study of high-intensity beam dynamics and of beam target interaction. Experimental studies of this kind will be of increasing importance for the future of the project. (orig.) [de

  12. Plasma viscosity with mass transport in spherical inertial confinement fusion implosion simulations

    Energy Technology Data Exchange (ETDEWEB)

    Vold, E. L.; Molvig, K. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Joglekar, A. S. [University of Michigan, Ann Arbor, Michigan 48109 (United States); Ortega, M. I. [University of New Mexico, Albuquerque, New Mexico 87131 (United States); Moll, R. [University of California, Santa Cruz, California 95064 (United States); Fenn, D. [Florida State University, Tallahassee, Florida 32306 (United States)

    2015-11-15

    The effects of viscosity and small-scale atomic-level mixing on plasmas in inertial confinement fusion (ICF) currently represent challenges in ICF research. Many current ICF hydrodynamic codes ignore the effects of viscosity though recent research indicates viscosity and mixing by classical transport processes may have a substantial impact on implosion dynamics. We have implemented a Lagrangian hydrodynamic code in one-dimensional spherical geometry with plasma viscosity and mass transport and including a three temperature model for ions, electrons, and radiation treated in a gray radiation diffusion approximation. The code is used to study ICF implosion differences with and without plasma viscosity and to determine the impacts of viscosity on temperature histories and neutron yield. It was found that plasma viscosity has substantial impacts on ICF shock dynamics characterized by shock burn timing, maximum burn temperatures, convergence ratio, and time history of neutron production rates. Plasma viscosity reduces the need for artificial viscosity to maintain numerical stability in the Lagrangian formulation and also modifies the flux-limiting needed for electron thermal conduction.

  13. Analysis of the neutron time-of-flight spectra from inertial confinement fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hatarik, R., E-mail: hatarik1@llnl.gov; Sayre, D. B.; Caggiano, J. A.; Phillips, T.; Eckart, M. J.; Bond, E. J.; Cerjan, C.; Grim, G. P.; Hartouni, E. P.; Mcnaney, J. M.; Munro, D. H. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Knauer, J. P. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States)

    2015-11-14

    Neutron time-of-flight diagnostics have long been used to characterize the neutron spectrum produced by inertial confinement fusion experiments. The primary diagnostic goals are to extract the d + t → n + α (DT) and d + d → n + {sup 3}He (DD) neutron yields and peak widths, and the amount DT scattering relative to its unscattered yield, also known as the down-scatter ratio (DSR). These quantities are used to infer yield weighted plasma conditions, such as ion temperature (T{sub ion}) and cold fuel areal density. We report on novel methodologies used to determine neutron yield, apparent T{sub ion}, and DSR. These methods invoke a single temperature, static fluid model to describe the neutron peaks from DD and DT reactions and a spline description of the DT spectrum to determine the DSR. Both measurements are performed using a forward modeling technique that includes corrections for line-of-sight attenuation and impulse response of the detection system. These methods produce typical uncertainties for DT T{sub ion} of 250 eV, 7% for DSR, and 9% for the DT neutron yield. For the DD values, the uncertainties are 290 eV for T{sub ion} and 10% for the neutron yield.

  14. Studying Validity of Single-Fluid Model in Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Gu Jian-Fa; Fan Zheng-Feng; Dai Zhen-Sheng; Ye Wen-Hua; Pei Wen-Bing; Zhu Shao-Ping

    2014-01-01

    The validity of single-fluid model in inertial confinement fusion simulations is studied by comparing the results of the multi- and single-fluid models. The multi-fluid model includes the effects of collision and interpenetration between fluid species. By simulating the collision of fluid species, steady-state shock propagation into the thin DT gas and expansion of hohlraum Au wall heated by lasers, the results show that the validity of single-fluid model is strongly dependent on the ratio of the characteristic length of the simulated system to the particle mean free path. When the characteristic length L is one order larger than the mean free path λ, the single-fluid model's results are found to be in good agreement with the multi-fluid model's simulations, and the modeling of single-fluid remains valid. If the value of L/λ is lower than 10, the interpenetration between fluid species is significant, and the single-fluid simulations show some unphysical results; while the multi-fluid model can describe well the interpenetration and mix phenomena, and give more reasonable results. (physics of gases, plasmas, and electric discharges)

  15. Recent progress on the Los Alamos Aurora ICF [inertial confinement fusion] laser system

    International Nuclear Information System (INIS)

    Rosocha, L.A.; Blair, L.S.

    1987-01-01

    Aurora is the Los Alamos short-pulse, high-power, krypton-fluoride laser system. It serves as an end-to-end technology demonstration prototype for large-scale ultraviolet laser systems for short wavelength inertial confinement fusion (ICF) investigations. The system is designed to employ optical angular multiplexing and serial amplification by electron-beam-driven KrF laser amplifiers to deliver stacked, 248-nm, 5-ns duration multikilojoule laser pulses to ICF-relevant targets. This paper presents a summary of the Aurora system and a discussion of the progress achieved in the construction and integration of the laser system. We concentrate on the main features of the following major system components: front-end lasers, amplifier train, multiplexer, optical relay train, demultiplexer, and the associated optical alignment system. During the past year, two major construction and integration tasks have been accomplished. The first task is the demonstration of 96-beam multiplexing and amplified energy extraction, as evidenced by the integrated operation of the front end, the multiplexer (12-fold and 8-fold encoders), the optical relay train, and three electron-beam-driven amplifiers. The second task is the assembly and installation of the demultiplexer optical hardware, which consists of over 300 optical components ranging in size from several centimeters square to over a meter square. 13 refs., 13 figs

  16. Three-dimensional hydrodynamics of the deceleration stage in inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C. R., E-mail: weber30@llnl.gov; Clark, D. S.; Cook, A. W.; Eder, D. C.; Haan, S. W.; Hammel, B. A.; Hinkel, D. E.; Jones, O. S.; Marinak, M. M.; Milovich, J. L.; Patel, P. K.; Robey, H. F.; Salmonson, J. D.; Sepke, S. M.; Thomas, C. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2015-03-15

    The deceleration stage of inertial confinement fusion implosions is modeled in detail using three-dimensional simulations designed to match experiments at the National Ignition Facility. In this final stage of the implosion, shocks rebound from the center of the capsule, forming the high-temperature, low-density hot spot and slowing the incoming fuel. The flow field that results from this process is highly three-dimensional and influences many aspects of the implosion. The interior of the capsule has high-velocity motion, but viscous effects limit the range of scales that develop. The bulk motion of the hot spot shows qualitative agreement with experimental velocity measurements, while the variance of the hot spot velocity would broaden the DT neutron spectrum, increasing the inferred temperature by 400–800 eV. Jets of ablator material are broken apart and redirected as they enter this dynamic hot spot. Deceleration stage simulations using two fundamentally different rad-hydro codes are compared and the flow field is found to be in good agreement.

  17. 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.

  18. Sensitivity of inertial confinement fusion hot spot properties to the deuterium-tritium fuel adiabat

    Energy Technology Data Exchange (ETDEWEB)

    Melvin, J.; Lim, H.; Rana, V.; Glimm, J. [Department of Applied Mathematics and Statistics, Stony Brook University, Stony Brook, New York 11794-3600 (United States); Cheng, B.; Sharp, D. H.; Wilson, D. C. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2015-02-15

    We determine the dependence of key Inertial Confinement Fusion (ICF) hot spot simulation properties on the deuterium-tritium fuel adiabat, here modified by addition of energy to the cold shell. Variation of this parameter reduces the simulation to experiment discrepancy in some, but not all, experimentally inferred quantities. Using simulations with radiation drives tuned to match experimental shots N120321 and N120405 from the National Ignition Campaign (NIC), we carry out sets of simulations with varying amounts of added entropy and examine the sensitivities of important experimental quantities. Neutron yields, burn widths, hot spot densities, and pressures follow a trend approaching their experimentally inferred quantities. Ion temperatures and areal densities are sensitive to the adiabat changes, but do not necessarily converge to their experimental quantities with the added entropy. This suggests that a modification to the simulation adiabat is one of, but not the only explanation of the observed simulation to experiment discrepancies. In addition, we use a theoretical model to predict 3D mix and observe a slight trend toward less mixing as the entropy is enhanced. Instantaneous quantities are assessed at the time of maximum neutron production, determined dynamically within each simulation. These trends contribute to ICF science, as an effort to understand the NIC simulation to experiment discrepancy, and in their relation to the high foot experiments, which features a higher adiabat in the experimental design and an improved neutron yield in the experimental results.

  19. Neutron imaging development for megajoule scale inertial confinement fusion experiments{sup 1}

    Energy Technology Data Exchange (ETDEWEB)

    Grim, G P; Bradley, P A; Day, R D; Clark, D D; Fatherley, V E; Finch, J P; Garcia, F P; Jaramillo, S A; Montoya, A J; Morgan, G L; Oertel, J A; Ortiz, T A; Payton, J R; Pazuchanics, P; Schmidt, D W; Valdez, A C; Wilde, C H; Wilke, M D; Wilson, D C [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)], E-mail: gpgrim@lanl.gov

    2008-05-15

    Neutron imaging of Inertial Confinement Fusion (ICF) targets is useful for understanding the implosion conditions of deuterium and tritium filled targets at Mega-Joule/Tera-Watt scale laser facilities. The primary task for imaging ICF targets at the National Ignition Facility, Lawrence Livermore National Laboratory, Livermore CA, is to determine the asymmetry of the imploded target. The image data, along with other nuclear information, are to be used to provide insight into target drive conditions. The diagnostic goal at the National Ignition Facility is to provide neutron images with 10 {mu}m resolution and peak signal-to-background values greater than 20 for neutron yields of {approx} 10{sup 15}. To achieve this requires signal multiplexing apertures with good resolution. In this paper we present results from imaging system development efforts aimed at achieving these requirements using neutron pinholes. The data were collected using directly driven ICF targets at the Omega Laser, University of Rochester, Rochester, NY., and include images collected from a 3 x 3 array of 15.5 {mu}m pinholes. Combined images have peak signal-to-background values greater than 30 at neutron yields of {approx} 10{sup 13}.

  20. Sensitivity of inertial confinement fusion hot spot properties to the deuterium-tritium fuel adiabat

    International Nuclear Information System (INIS)

    Melvin, J.; Lim, H.; Rana, V.; Glimm, J.; Cheng, B.; Sharp, D. H.; Wilson, D. C.

    2015-01-01

    We determine the dependence of key Inertial Confinement Fusion (ICF) hot spot simulation properties on the deuterium-tritium fuel adiabat, here modified by addition of energy to the cold shell. Variation of this parameter reduces the simulation to experiment discrepancy in some, but not all, experimentally inferred quantities. Using simulations with radiation drives tuned to match experimental shots N120321 and N120405 from the National Ignition Campaign (NIC), we carry out sets of simulations with varying amounts of added entropy and examine the sensitivities of important experimental quantities. Neutron yields, burn widths, hot spot densities, and pressures follow a trend approaching their experimentally inferred quantities. Ion temperatures and areal densities are sensitive to the adiabat changes, but do not necessarily converge to their experimental quantities with the added entropy. This suggests that a modification to the simulation adiabat is one of, but not the only explanation of the observed simulation to experiment discrepancies. In addition, we use a theoretical model to predict 3D mix and observe a slight trend toward less mixing as the entropy is enhanced. Instantaneous quantities are assessed at the time of maximum neutron production, determined dynamically within each simulation. These trends contribute to ICF science, as an effort to understand the NIC simulation to experiment discrepancy, and in their relation to the high foot experiments, which features a higher adiabat in the experimental design and an improved neutron yield in the experimental results

  1. Sensitivity of inertial confinement fusion hot spot properties to the deuterium-tritium fuel adiabat

    Science.gov (United States)

    Melvin, J.; Lim, H.; Rana, V.; Cheng, B.; Glimm, J.; Sharp, D. H.; Wilson, D. C.

    2015-02-01

    We determine the dependence of key Inertial Confinement Fusion (ICF) hot spot simulation properties on the deuterium-tritium fuel adiabat, here modified by addition of energy to the cold shell. Variation of this parameter reduces the simulation to experiment discrepancy in some, but not all, experimentally inferred quantities. Using simulations with radiation drives tuned to match experimental shots N120321 and N120405 from the National Ignition Campaign (NIC), we carry out sets of simulations with varying amounts of added entropy and examine the sensitivities of important experimental quantities. Neutron yields, burn widths, hot spot densities, and pressures follow a trend approaching their experimentally inferred quantities. Ion temperatures and areal densities are sensitive to the adiabat changes, but do not necessarily converge to their experimental quantities with the added entropy. This suggests that a modification to the simulation adiabat is one of, but not the only explanation of the observed simulation to experiment discrepancies. In addition, we use a theoretical model to predict 3D mix and observe a slight trend toward less mixing as the entropy is enhanced. Instantaneous quantities are assessed at the time of maximum neutron production, determined dynamically within each simulation. These trends contribute to ICF science, as an effort to understand the NIC simulation to experiment discrepancy, and in their relation to the high foot experiments, which features a higher adiabat in the experimental design and an improved neutron yield in the experimental results.

  2. X-ray crystal imagers for inertial confinement fusion experiments (invited)

    International Nuclear Information System (INIS)

    Aglitskiy, Y.; Lehecka, T.; Obenschain, S.; Pawley, C.; Brown, C.M.; Seely, J.

    1999-01-01

    We report on our continued development of high resolution monochromatic x-ray imaging system based on spherically curved crystals. This system can be extensively used in the relevant experiments of the inertial confinement fusion (ICF) program. The system is currently used, but not limited to diagnostics of the targets ablatively accelerated by the Nike KrF laser. A spherically curved quartz crystal (2d=6.68703 Angstrom, R=200mm) has been used to produce monochromatic backlit images with the He-like Si resonance line (1865 eV) as the source of radiation. Another quartz crystal (2d=8.5099 Angstrom, R=200mm) with the H-like Mg resonance line (1473 eV) has been used for backlit imaging with higher contrast. The spatial resolution of the x-ray optical system is 1.7 μm in selected places and 2 - 3 μm over a larger area. A second crystal with a separate backlighter was added to the imaging system. This makes it possible to make use of all four strips of the framing camera. Time resolved, 20x magnified, backlit monochromatic images of CH planar targets driven by the Nike facility have been obtained with spatial resolution of 2.5 μm in selected places and 5 μm over the focal spot of the Nike laser. We are exploring the enhancement of this technique to the higher and lower backlighter energies. copyright 1999 American Institute of Physics

  3. Refraction-enhanced backlit imaging of axially symmetric inertial confinement fusion plasmas.

    Science.gov (United States)

    Koch, Jeffrey A; Landen, Otto L; Suter, Laurence J; Masse, Laurent P; Clark, Daniel S; Ross, James S; Mackinnon, Andrew J; Meezan, Nathan B; Thomas, Cliff A; Ping, Yuan

    2013-05-20

    X-ray backlit radiographs of dense plasma shells can be significantly altered by refraction of x rays that would otherwise travel straight-ray paths, and this effect can be a powerful tool for diagnosing the spatial structure of the plasma being radiographed. We explore the conditions under which refraction effects may be observed, and we use analytical and numerical approaches to quantify these effects for one-dimensional radial opacity and density profiles characteristic of inertial-confinement fusion (ICF) implosions. We also show how analytical and numerical approaches allow approximate radial plasma opacity and density profiles to be inferred from point-projection refraction-enhanced radiography data. This imaging technique can provide unique data on electron density profiles in ICF plasmas that cannot be obtained using other techniques, and the uniform illumination provided by point-like x-ray backlighters eliminates a significant source of uncertainty in inferences of plasma opacity profiles from area-backlit pinhole imaging data when the backlight spatial profile cannot be independently characterized. The technique is particularly suited to in-flight radiography of imploding low-opacity shells surrounding hydrogen ice, because refraction is sensitive to the electron density of the hydrogen plasma even when it is invisible to absorption radiography. It may also provide an alternative approach to timing shockwaves created by the implosion drive, that are currently invisible to absorption radiography.

  4. Pulse*Star Inertial Confinement Fusion Reactor: heat transfer loop and balance of plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Murray, K.A.

    1984-01-01

    A conceptual heat transfer loop and balance of plant design for the Pulse*Star Inertial Confinement Fusion Reactor has been investigated and results are presented. The Pulse*Star reaction vessel, a perforated steel bell jar approximately 11 m in diameter, is immersed in Li 17 Pb 83 coolant which flows through the perforations and forms a 1.5 m thick plenum of droplets around an 8 m diameter inner chamber. The reactor and associated pumps, piping, and steam generators are contained within a 17 m diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops with flow rates of 5.5 m 3 /s each are necessary to transfer 3300 MWt of power. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. Power balance calculations based on an improved electrical conversion efficiency revealed a net electrical output of 1260 MWe to the bus bar and a resulting net efficiency of 39%. Suggested balance-of-plant layouts are also presented

  5. Three-dimensional hydrodynamics of the deceleration stage in inertial confinement fusion

    International Nuclear Information System (INIS)

    Weber, C. R.; Clark, D. S.; Cook, A. W.; Eder, D. C.; Haan, S. W.; Hammel, B. A.; Hinkel, D. E.; Jones, O. S.; Marinak, M. M.; Milovich, J. L.; Patel, P. K.; Robey, H. F.; Salmonson, J. D.; Sepke, S. M.; Thomas, C. A.

    2015-01-01

    The deceleration stage of inertial confinement fusion implosions is modeled in detail using three-dimensional simulations designed to match experiments at the National Ignition Facility. In this final stage of the implosion, shocks rebound from the center of the capsule, forming the high-temperature, low-density hot spot and slowing the incoming fuel. The flow field that results from this process is highly three-dimensional and influences many aspects of the implosion. The interior of the capsule has high-velocity motion, but viscous effects limit the range of scales that develop. The bulk motion of the hot spot shows qualitative agreement with experimental velocity measurements, while the variance of the hot spot velocity would broaden the DT neutron spectrum, increasing the inferred temperature by 400–800 eV. Jets of ablator material are broken apart and redirected as they enter this dynamic hot spot. Deceleration stage simulations using two fundamentally different rad-hydro codes are compared and the flow field is found to be in good agreement

  6. Terrestrial and extraterrestrial superresonators as drivers for an inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Vath, W.

    1992-01-01

    This paper reports on the recirculating power fraction of a laser-driven inertial confinement fusion (ICF) reactor which can be reduced by using laser diodes to pump a neodymium solid-state laser. To overcome the high costs of two-dimensional arrays of laser diodes, two types of superresonators are proposed: a terrestrially based one and an extraterrestrially based one on a geostationary orbit. Both are designed in such a way that a sequence of short laser pulses (10 to 20 ns wide), each with an energy of 5 to 10 MJ and a frequency of 10 Hz, are produced to trigger a deuterium-tritium ICF reactor. The terrestrial superresonator needs a much smaller number of two-dimensional laser diode arrays than a conventionally pumped once-through solid-state laser system, and the extraterrestrial resonator is pumped by means of concentrated solar radiation. In practice, at least an order of magnitude fewer laser diodes and crystalline calcium fluoride gain media are needed to meet the requirements of a laser driver for an ICF reactor. If, finally, a liquid neodymium laser system could be used for an ICF reactor, the cooling of the gain slabs would be facilitated substantially

  7. Advances in HYDRA and its application to simulations of Inertial Confinement Fusion targets

    Science.gov (United States)

    Marinak, M. M.; Kerbel, G. D.; Koning, J. M.; Patel, M. V.; Sepke, S. M.; Brown, P. N.; Chang, B.; Procassini, R.; Veitzer, S. A.

    2008-11-01

    We will outline new capabilities added to the HYDRA 2D/3D multiphysics ICF simulation code. These include a new SN multigroup radiation transport package (1D), constitutive models for elastic-plastic (strength) effects, and a mix model. A Monte Carlo burn package is being incorporated to model diagnostic signatures of neutrons, gamma rays and charged particles. A 3D MHD package that treats resistive MHD is available. Improvements to HYDRA's implicit Monte Carlo photonics package, including the addition of angular biasing, now enable integrated hohlraum simulations to complete in substantially shorter time. The heavy ion beam deposition package now includes a new model for ion stopping power developed by the Tech-X Corporation, with improved accuracy below the Bragg peak. Examples will illustrate HYDRA's enhanced capabilities to simulate various aspects of inertial confinement fusion targets.This work was performed under the auspices of the Lawrence Livermore National Security, LLC, (LLNS) under Contract No. DE-AC52-07NA27344. The work of Tech-X personnel was funded by the Department of Energy under Small Business Innovation Research Contract No. DE-FG02-03ER83797.

  8. A high-energy x-ray microscope for inertial confinement fusion

    International Nuclear Information System (INIS)

    Marshall, F.J.; Bennett, G.R.

    1999-01-01

    We have developed a microscope capable of imaging x-ray emission from inertial confinement fusion targets in the range of 7 - 9 keV. Imaging is accomplished with a Kirkpatrick-Baez type, four-image microscope coated with a WB 4 C multilayer having a 2d period of 140 Angstrom. This microscope design (a standard used on the University of Rochester close-quote s OMEGA laser system) is capable of 5 μm resolution over a region large enough to image an imploded target (∼400 μm). This design is capable of being extended to ∼40 keV if state-of-the-art, short-spacing, multilayer coatings are used (∼25 Angstrom), and has been configured to obtain 3 μm resolution with the appropriate choice of mirror size. As such, this type of microscope could serve as a platform for multiframe, hard x-ray imaging on the National Ignition Facility. Characterization of the microscope and laboratory measurements of the energy response made with a cw x-ray source will be shown. copyright 1999 American Institute of Physics

  9. One-dimensional Lagrangian implicit hydrodynamic algorithm for Inertial Confinement Fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Ramis, Rafael, E-mail: rafael.ramis@upm.es

    2017-02-01

    A new one-dimensional hydrodynamic algorithm, specifically developed for Inertial Confinement Fusion (ICF) applications, is presented. The scheme uses a fully conservative Lagrangian formulation in planar, cylindrical, and spherically symmetric geometries, and supports arbitrary equations of state with separate ion and electron components. Fluid equations are discretized on a staggered grid and stabilized by means of an artificial viscosity formulation. The space discretized equations are advanced in time using an implicit algorithm. The method includes several numerical parameters that can be adjusted locally. In regions with low Courant–Friedrichs–Lewy (CFL) number, where stability is not an issue, they can be adjusted to optimize the accuracy. In typical problems, the truncation error can be reduced by a factor between 2 to 10 in comparison with conventional explicit algorithms. On the other hand, in regions with high CFL numbers, the parameters can be set to guarantee unconditional stability. The method can be integrated into complex ICF codes. This is demonstrated through several examples covering a wide range of situations: from thermonuclear ignition physics, where alpha particles are managed as an additional species, to low intensity laser–matter interaction, where liquid–vapor phase transitions occur.

  10. Hydrodynamic stability theory of double ablation front structures in inertial confinement fusion

    International Nuclear Information System (INIS)

    Yanez Vico, C.

    2012-11-01

    For moderate-Z materials, the hydrodynamic structure of the ablation region formed by the irradiation of high intensity laser beams differs from that of low-Z materials (hydrogenic ablators). In particular, the role played by the radiative energy flux becomes non-negligible for increasing atomic number material and ended up forming a second ablation front. This structure of two separated ablation fronts, called double ablation (DA) front, was confirmed in the simulations carried out by Fujioka et al. In this work a linear stability theory of DA fronts is developed for direct-drive inertial confinement fusion targets. Two models are proposed. First, a sharp boundary model where the thin front approximation is assumed for both ablation fronts. The information about the corona region that permits to close the sharp boundary model is obtained from a prior self-consistent analysis of the electronic-radiative ablation (ERA) front. Numerical results are presented as well as an analytical approach for the radiation dominated regime of very steep double ablation front structure. Second, a self-consistent numerical method where the finite length of the ablation fronts is considered. Accurate hydrodynamic profiles are taken into account in the theoretical model by means of a fitting parameters method using one-dimensional simulation results. Numerical dispersion relation is compared to the analytical sharp boundary model showing an excellent agreement for the radiation dominated regime, and the stabilization due to smooth profiles. 2D simulations are presented to validate the linear stability theory

  11. Demonstration of radiation pulse shaping with nested-tungsten-wire-array pinches for high-yield inertial confinement fusion.

    Science.gov (United States)

    Cuneo, M E; Vesey, R A; Sinars, D B; Chittenden, J P; Waisman, E M; Lemke, R W; Lebedev, S V; Bliss, D E; Stygar, W A; Porter, J L; Schroen, D G; Mazarakis, M G; Chandler, G A; Mehlhorn, T A

    2005-10-28

    Nested wire-array pinches are shown to generate soft x-ray radiation pulse shapes required for three-shock isentropic compression and hot-spot ignition of high-yield inertial confinement fusion capsules. We demonstrate a reproducible and tunable foot pulse (first shock) produced by interaction of the outer and inner arrays. A first-step pulse (second shock) is produced by inner array collision with a central CH2 foam target. Stagnation of the inner array at the axis produces the third shock. Capsules optimized for several of these shapes produce 290-900 MJ fusion yields in 1D simulations.

  12. Potential formation in the plasma confinement region of a radio-frequency plugged linear device

    International Nuclear Information System (INIS)

    Fujita, Hideki; Kumazawa, Ryuhei; Howald, A.M.; Okamura, Shoichi; Sato, Teruyuki; Adati, Keizo; Garner, H.R.; Nishimura, Kiyohiko.

    1987-08-01

    Plasma potential formation in an open-ended plasma confinement system with RF plugging (the RFC-XX-M device) is investigated. The plasma potential in the central confinement region is measured with a heavy ion beam probe system and potentials at the RF plug section are measured with multi-grid energy analyzers. The measured plasma potential is compared with that deduced from the generalized Pastukhov formula. Results show that the plasma potential develops as an ambipolar potential to equate ion and electron end losses. During RF plugging, electrons are heated by Landau damping, while ions are not heated since adiabatic conditions apply during ion plugging in this experiment. (author)

  13. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    Karpenko, V.N.

    1983-01-01

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

  14. Electrical insulation and conduction coating for fusion experimental devices

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori; Tsujimura, Seiji; Toyoda, Masahiko; Inoue, Masahiko [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Abe, Tetsuya; Murakami, Yoshio [Japan Atomic Energy Research Inst., Naka (Japan)

    1996-01-01

    The development of electrical insulation and conduction coating methods that can be applied to large components of fusion experimental devices has been investigated. A thermal spraying method is used to coat the insulation or conduction materials on the structural components because of its applicability for large surfaces. The insulation material chosen was Al{sub 2}O{sub 3}, while Cr{sub 3}C{sub 2}-NiCr and WC-NiCr were chosen as conduction materials. These materials were coated on stainless steel substrates to examine the basic characteristics of the coated layers, such as their adhesive strength to the substrate, thermal shock resistance, electrical resistance, dielectric breakdown voltage, and thermal conductivity. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed adequate frictional properties. The spraying method was tested on a 100- x 1000-mm surface and found to be applicable for large surfaces of experimental fusion devices. 9 refs., 6 figs., 15 tabs.

  15. Electrical insulation and conduction coating for fusion experimental devices

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Tsujimura, Seiji; Toyoda, Masahiko; Inoue, Masahiko; Abe, Tetsuya; Murakami, Yoshio

    1996-01-01

    The development of electrical insulation and conduction coating methods that can be applied to large components of fusion experimental devices has been investigated. A thermal spraying method is used to coat the insulation or conduction materials on the structural components because of its applicability for large surfaces. The insulation material chosen was Al 2 O 3 , while Cr 3 C 2 -NiCr and WC-NiCr were chosen as conduction materials. These materials were coated on stainless steel substrates to examine the basic characteristics of the coated layers, such as their adhesive strength to the substrate, thermal shock resistance, electrical resistance, dielectric breakdown voltage, and thermal conductivity. It was found that they have sufficient electrical insulation and conduction properties, respectively. In addition, the sliding tests of the coated layers showed adequate frictional properties. The spraying method was tested on a 100- x 1000-mm surface and found to be applicable for large surfaces of experimental fusion devices. 9 refs., 6 figs., 15 tabs

  16. Design and construction of a fast (neutron) spectrometer for magnetic confinement fusion studies

    International Nuclear Information System (INIS)

    Lagattu, A.

    1980-09-01

    Neutron emission in conjunction with the 'Tokamak' devices studied in order to carry out controlled fusion is done at present on a short time scale (a few hundred milliseconds). Obtaining spectrometric information on this emission made it necessary to design sufficiently fast equipment to adapt to this situation. The basis of the equipment selected is conventional: NE213 liquid scintillator in conjunction with a fast photomultiplier. A commercialized fast neutron gamma discriminator (approximately 210 5 S -1 ) was used. The data acquisition and processing system was specially made for this occasion. This assembly, so designed and constructed, was tested on a particle accelerator and on the TFR Tokamak. In the best conditions a spectrum of 60 milliseconds was produced. Hence several spectra can be recorded during an experiment. The rejection of gamma radiations by the discriminator exceeds 99% in such conditions [fr

  17. Stable confinement of toroidal electron plasma in an internal conductor device Prototype-Ring Trap

    International Nuclear Information System (INIS)

    Saitoh, H.; Yoshida, Z.; Watanabe, S.

    2005-01-01

    A pure electron plasma has been produced in an internal conductor device Prototype-Ring Trap (Proto-RT). The temporal evolution of the electron plasma was investigated by the measurement of electrostatic fluctuations. Stable confinement was realized when the potential profile adjusted to match the magnetic surfaces. The confinement time varies as a function of the magnetic field strength and the neutral gas pressure, and is comparable to the diffusion time of electrons determined by the classical collisions with neutral gas. Although the addition of a toroidal magnetic field stabilized the electrostatic fluctuation of the plasma, the effects of the magnetic shear shortened the stable confinement time, possibly because of the obstacles of coil support structures

  18. On the Utility of Antiprotons as Drivers for Inertial Confinement Fusion

    International Nuclear Information System (INIS)

    Perkins, L J; Orth, C D; Tabak, M

    2003-01-01

    By contrast to the large mass, complexity and recirculating power of conventional drivers for inertial confinement fusion (ICF), antiproton annihilation offers a specific energy of 90MJ/(micro)g and thus a unique form of energy packaging and delivery. In principle, antiproton drivers could provide a profound reduction in system mass for advanced space propulsion by ICF. We examine the physics underlying the use of antiprotons ((bar p)) to drive various classes of high-yield ICF targets by the methods of volumetric ignition, hotspot ignition and fast ignition. The useable fraction of annihilation deposition energy is determined for both (bar p)-driven ablative compression and (bar p)-driven fast ignition, in association with 0-D and 1-D target burn models. Thereby, we deduce scaling laws for the number of injected antiprotons required per capsule, together with timing and focal spot requirements. The kinetic energy of the injected antiproton beam required to penetrate to the desired annihilation point is always small relative to the deposited annihilation energy. We show that heavy metal seeding of the fuel and/or ablator is required to optimize local deposition of annihilation energy and determine that a minimum of ∼3x10 15 injected antiprotons will be required to achieve high yield (several hundred megajoules) in any target configuration. Target gains - i.e., fusion yields divided by the available p - (bar p) annihilation energy from the injected antiprotons (1.88GeV/(bar p)) - range from ∼3 for volumetric ignition targets to ∼600 for fast ignition targets. Antiproton-driven ICF is a speculative concept, and the handling of antiprotons and their required injection precision - temporally and spatially - will present significant technical challenges. The storage and manipulation of low-energy antiprotons, particularly in the form of antihydrogen, is a science in its infancy and a large scale-up of antiproton production over present supply methods would be

  19. On the utility of antiprotons as drivers for inertial confinement fusion

    International Nuclear Information System (INIS)

    Perkins, L. John; Orth, Charles D.; Tabak, Max

    2004-01-01

    In contrast to the large mass, complexity and recirculating power of conventional drivers for inertial confinement fusion (ICF), antiproton annihilation offers a specific energy of 90 MJ μg -1 and thus a unique form of energy packaging and delivery. In principle, antiproton drivers could provide a profound reduction in system mass for advanced space propulsion by ICF. We examine the physics underlying the use of antiprotons (p-bar) to drive various classes of high-yield ICF targets by the methods of volumetric ignition, hotspot ignition and fast ignition. The useable fraction of annihilation deposition energy is determined for both p-bar-driven ablative compression and p-bar-driven fast ignition, in association with zero- and one-dimensional target burn models. Thereby, we deduce scaling laws for the number of injected antiprotons required per capsule, together with timing and focal spot requirements. The kinetic energy of the injected antiproton beam required to penetrate to the desired annihilation point is always small relative to the deposited annihilation energy. We show that heavy metal seeding of the fuel and/or ablator is required to optimize local deposition of annihilation energy and determine that a minimum of ∼3 x 10 15 injected antiprotons will be required to achieve high yield (several hundred megajoules) in any target configuration. Target gains-i.e. fusion yields divided by the available p-p-bar annihilation energy from the injected antiprotons (1.88 GeV/p-bar)-range from ∼3 for volumetric ignition targets to ∼600 for fast ignition targets. Antiproton-driven ICF is a speculative concept, and the handling of antiprotons and their required injection precision - temporally and spatially - will present significant technical challenges. The storage and manipulation of low-energy antiprotons, particularly in the form of antihydrogen, is a science in its infancy and a large scale-up of antiproton production over present supply methods would be

  20. VISTA -- A Vehicle for Interplanetary Space Transport Application Powered by Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Orth, C D

    2005-03-31

    Inertial Confinement Fusion (ICF) is an ideal technology to power self-contained single-stage piloted (manned) spacecraft within the solar system because of its inherently high power/mass ratios and high specific impulses (i.e., high exhaust velocities). These technological advantages are retained when ICF is utilized with a magnetic thrust chamber, which avoids the plasma thermalization and resultant degradation of specific impulse that are unavoidable with the use of mechanical thrust chambers. We started with Rod Hyde's 1983 description of an ICF-powered engine concept using a magnetic thrust chamber, and conducted a more detailed systems study to develop a viable, realistic, and defensible spacecraft concept based on ICF technology projected to be available in the first half of the 21st century. The results include an entirely new conical spacecraft conceptual design utilizing near-existing radiator technology. We describe the various vehicle systems for this new concept, estimate the missions performance capabilities for general missions to the planets within the solar system, and describe in detail the performance for the baseline mission of a piloted roundtrip to Mars with a 100-ton payload. For this mission, we show that roundtrips totaling {ge}145 days are possible with advanced DT fusion technology and a total (wet) spacecraft mass of about 6000 metric tons. Such short-duration missions are advantageous to minimize the known cosmic-radiation hazards to astronauts, and are even more important to minimize the physiological deteriorations arising from zero gravity. These ICF-powered missions are considerably faster than those available using chemical or nuclear-electric-propulsion technologies with minimum-mass vehicle configurations. VISTA also offers onboard artificial gravity and propellant-based shielding from cosmic rays, thus reducing the known hazards and physiological deteriorations to insignificant levels. We emphasize, however, that the degree

  1. On the Utility of Antiprotons as Drivers for Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, L J; Orth, C D; Tabak, M

    2003-10-20

    By contrast to the large mass, complexity and recirculating power of conventional drivers for inertial confinement fusion (ICF), antiproton annihilation offers a specific energy of 90MJ/{micro}g and thus a unique form of energy packaging and delivery. In principle, antiproton drivers could provide a profound reduction in system mass for advanced space propulsion by ICF. We examine the physics underlying the use of antiprotons ({bar p}) to drive various classes of high-yield ICF targets by the methods of volumetric ignition, hotspot ignition and fast ignition. The useable fraction of annihilation deposition energy is determined for both {bar p}-driven ablative compression and {bar p}-driven fast ignition, in association with 0-D and 1-D target burn models. Thereby, we deduce scaling laws for the number of injected antiprotons required per capsule, together with timing and focal spot requirements. The kinetic energy of the injected antiproton beam required to penetrate to the desired annihilation point is always small relative to the deposited annihilation energy. We show that heavy metal seeding of the fuel and/or ablator is required to optimize local deposition of annihilation energy and determine that a minimum of {approx}3x10{sup 15} injected antiprotons will be required to achieve high yield (several hundred megajoules) in any target configuration. Target gains - i.e., fusion yields divided by the available p - {bar p} annihilation energy from the injected antiprotons (1.88GeV/{bar p}) - range from {approx}3 for volumetric ignition targets to {approx}600 for fast ignition targets. Antiproton-driven ICF is a speculative concept, and the handling of antiprotons and their required injection precision - temporally and spatially - will present significant technical challenges. The storage and manipulation of low-energy antiprotons, particularly in the form of antihydrogen, is a science in its infancy and a large scale-up of antiproton production over present supply

  2. Detailed high-resolution three-dimensional simulations of OMEGA separated reactants inertial confinement fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Haines, Brian M., E-mail: bmhaines@lanl.gov; Fincke, James R.; Shah, Rahul C.; Boswell, Melissa; Fowler, Malcolm M.; Gore, Robert A.; Hayes-Sterbenz, Anna C.; Jungman, Gerard; Klein, Andreas; Rundberg, Robert S.; Steinkamp, Michael J.; Wilhelmy, Jerry B. [Los Alamos National Laboratory, MS T087, Los Alamos, New Mexico 87545 (United States); Grim, Gary P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Forrest, Chad J.; Silverstein, Kevin; Marshall, Frederic J. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States)

    2016-07-15

    We present results from the comparison of high-resolution three-dimensional (3D) simulations with data from the implosions of inertial confinement fusion capsules with separated reactants performed on the OMEGA laser facility. Each capsule, referred to as a “CD Mixcap,” is filled with tritium and has a polystyrene (CH) shell with a deuterated polystyrene (CD) layer whose burial depth is varied. In these implosions, fusion reactions between deuterium and tritium ions can occur only in the presence of atomic mix between the gas fill and shell material. The simulations feature accurate models for all known experimental asymmetries and do not employ any adjustable parameters to improve agreement with experimental data. Simulations are performed with the RAGE radiation-hydrodynamics code using an Implicit Large Eddy Simulation (ILES) strategy for the hydrodynamics. We obtain good agreement with the experimental data, including the DT/TT neutron yield ratios used to diagnose mix, for all burial depths of the deuterated shell layer. Additionally, simulations demonstrate good agreement with converged simulations employing explicit models for plasma diffusion and viscosity, suggesting that the implicit sub-grid model used in ILES is sufficient to model these processes in these experiments. In our simulations, mixing is driven by short-wavelength asymmetries and longer-wavelength features are responsible for developing flows that transport mixed material towards the center of the hot spot. Mix material transported by this process is responsible for most of the mix (DT) yield even for the capsule with a CD layer adjacent to the tritium fuel. Consistent with our previous results, mix does not play a significant role in TT neutron yield degradation; instead, this is dominated by the displacement of fuel from the center of the implosion due to the development of turbulent instabilities seeded by long-wavelength asymmetries. Through these processes, the long

  3. Laser - driven high - energy ions and their application to inertial confinement fusion

    International Nuclear Information System (INIS)

    Borghesi, M.

    2007-01-01

    The acceleration of high-energy ion beams (up to several tens of MeV per nucleon) following the interaction of short and intense laser pulses with solid targets has been one of the most important results of recent laser-plasma research [1]. The acceleration is driven by relativistic electrons, which acquire energy directly from the laser pulse and set up extremely large (∼TV/m) space charge fields at the target interfaces. The properties of laser-driven ion beams (high brightness and laminarity, high-energy cut-off, ultrashort burst duration) distinguish them from lower energy ions accelerated in earlier experiments at moderate laser intensities, and compare favourably with those of 'conventional' accelerator beams. In view of these properties, laser-driven ion beams can be employed in a number of innovative applications in the scientific, technological and medical areas. We will discuss in particular aspects of interest to their application in an Inertial Confinement Fusion context. Laser-driven protons are indeed being considered as a possible trigger for Fast Ignition of a precompressed fuel.[2] Recent results relating to the optimization of beam energy and focusing will be presented. These include the use of laser-driven impulsive fields for proton beam collimation and focusing [3], and the investigation of acceleration in presence of finite-scale plasma gradient. Proposed target developments enabling proton production at high repetition rate will also be discussed. Another important area of application of proton beams is diagnostic use in a particle probing arrangement for detection of density non-homogeneities [4] and electric/magnetic fields [5]. We will discuss the use of laser-driven proton beams for the diagnosis of magnetic and electric fields in planar and hohlraum targets and for the detection of fields associated to relativistic electron propagation through dense matter, an issue of high relevance for electron driven Fast Ignition. [1] M

  4. Simultaneous usage of pinhole and penumbral apertures for imaging small scale neutron sources from inertial confinement fusion experiments.

    Science.gov (United States)

    Guler, N; Volegov, P; Danly, C R; Grim, G P; Merrill, F E; Wilde, C H

    2012-10-01

    Inertial confinement fusion experiments at the National Ignition Facility are designed to understand the basic principles of creating self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT) filled cryogenic plastic capsules. The neutron imaging diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by observing neutron images in two different energy bands for primary (13-17 MeV) and down-scattered (6-12 MeV) neutrons. From this, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. These experiments provide small sources with high yield neutron flux. An aperture design that includes an array of pinholes and penumbral apertures has provided the opportunity to image the same source with two different techniques. This allows for an evaluation of these different aperture designs and reconstruction algorithms.

  5. The effects of convergence ratio on the implosion behavior of DT layered inertial confinement fusion capsules

    Science.gov (United States)

    Haines, Brian M.; Yi, S. A.; Olson, R. E.; Khan, S. F.; Kyrala, G. A.; Zylstra, A. B.; Bradley, P. A.; Peterson, R. R.; Kline, J. L.; Leeper, R. J.; Shah, R. C.

    2017-07-01

    The wetted foam capsule design for inertial confinement fusion capsules, which includes a foam layer wetted with deuterium-tritium liquid, enables layered capsule implosions with a wide range of hot-spot convergence ratios (CR) on the National Ignition Facility. We present a full-scale wetted foam capsule design that demonstrates high gain in one-dimensional simulations. In these simulations, increasing the convergence ratio leads to an improved capsule yield due to higher hot-spot temperatures and increased fuel areal density. High-resolution two-dimensional simulations of this design are presented with detailed and well resolved models for the capsule fill tube, support tent, surface roughness, and predicted asymmetries in the x-ray drive. Our modeling of these asymmetries is validated by comparisons with available experimental data. In 2D simulations of the full-scale wetted foam capsule design, jetting caused by the fill tube is prevented by the expansion of the tungsten-doped shell layer due to preheat. While the impacts of surface roughness and predicted asymmetries in the x-ray drive are enhanced by convergence effects, likely underpredicted in 2D at high CR, simulations predict that the capsule is robust to these features. Nevertheless, the design is highly susceptible to the effects of the capsule support tent, which negates all of the one-dimensional benefits of increasing the convergence ratio. Indeed, when the support tent is included in simulations, the yield decreases as the convergence ratio is increased for CR > 20. Nevertheless, the results suggest that the full-scale wetted foam design has the potential to outperform ice layer capsules given currently achievable levels of asymmetries when fielded at low convergence ratios (CR < 20).

  6. Condensation of ablated first-wall materials in the cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Ladd, A.J.C.

    1985-01-01

    This report concerns problems involved in recondensing first-wall materials vaporized by x rays and pellet debris in the Cascade inertial confinement fusion reactor. It examines three proposed first-wall materials, beryllium oxide (BeO), silicon carbide (SiO), and pyrolytic graphite (C), paying particular attention to the chemical equilibrium and kinetics of the vaporized gases. The major results of this study are as follows. Ceramic materials composed of diatomic molecules, such as BeO and SiC, exist as highly dissociated species after vaporization. The low gas density precludes significant recombination during times of interest (i.e., less than 0.1 s). The dissociated species (Be, O, Si, and C) are, except for carbon, quite volatile and are thermodynamically stable as a vapor under the high temperature and low density found in Cascade. These materials are thus unsuitable as first-wall materials. This difficulty is avoided with pyrolytic graphite. Since the condensation coefficient of monatomic carbon vapor (approx. 0.5) is greater than that of the polyatomic vapor (<0.1), recondensation is assisted by the expected high degree of dissociation. The proposed 10-layer granular carbon bed is sufficient to condense all the carbon vapor before it penetrates to the BeO layer below. The effective condensation coefficient of the porous bed is about 50% greater than that of a smooth wall. An estimate of the mass flux leaving the chamber results in a condensation time for a carbon first wall of about 30 to 50 ms. An experiment to investigate condensation in a Cascade-like chamber is proposed

  7. Inertial confinement fusion driver enhancements: Final focusing systems and compact heavy-ion driver designs

    International Nuclear Information System (INIS)

    Bieri, R.L.

    1991-01-01

    Required elements of an inertial confinement fusion power plant are modeled and discussed. A detailed analysis of two critical elements of candidate drivers is done, and new component designs are proposed to increase the credibility and feasibility of each driver system. An analysis of neutron damage to the final elements of a laser focusing system is presented, and multilayer -- dielectric mirrors are shown to have damage lifetimes which axe too short to be useful in a commercial power plant. A new final-focusing system using grazing incidence metal mirrors to protect sensitive laser optics is designed and shown to be effective in extending the lifetime of the final focusing system. The reflectivities and damage limits of grazing incidence metal mirrors are examined in detail, and the required mirror sizes are shown to be compatible with the beam sizes and illumination geometries currently envisioned for laser drivers. A detailed design and analysis is also done for compact arrays of superconducting magnetic quadrupoles, which are needed in a multi-beam heavy-ion driver. The new array model is developed in more detail than some previous conceptual designs and models arrays which are more compact than arrays scaled from existing single -- quadrupole designs. The improved integrated model for compact arrays is used to compare the effects of various quadrupole array design choices on the size and cost of a heavy-ion driver. Array design choices which significantly affect the cost of a heavy-ion driver include the choice of superconducting material and the thickness of the collar used to support the winding stresses. The effect of these array design choices on driver size and cost is examined and the array model is used to estimate driver cost savings and performance improvements attainable with aggressive quadrupole array designs with high-performance superconductors

  8. On the synchrotron emission in kinetic simulations of runaway electrons in magnetic confinement fusion plasmas

    Science.gov (United States)

    Carbajal, L.; del-Castillo-Negrete, D.

    2017-12-01

    Developing avoidance or mitigation strategies of runaway electrons (REs) in magnetic confinement fusion (MCF) plasmas is of crucial importance for the safe operation of ITER. In order to develop these strategies, an accurate diagnostic capability that allows good estimates of the RE distribution function in these plasmas is needed. Synchrotron radiation (SR) of RE in MCF, besides of being one of the main damping mechanisms for RE in the high energy relativistic regime, is routinely used in current MCF experiments to infer the parameters of RE energy and pitch angle distribution functions. In the present paper we address the long standing question about what are the relationships between different REs distribution functions and their corresponding synchrotron emission simultaneously including: full-orbit effects, information of the spectral and angular distribution of SR of each electron, and basic geometric optics of a camera. We study the spatial distribution of the SR on the poloidal plane, and the statistical properties of the expected value of the synchrotron spectra of REs. We observe a strong dependence of the synchrotron emission measured by the camera on the pitch angle distribution of runaways, namely we find that crescent shapes of the spatial distribution of the SR as measured by the camera relate to RE distributions with small pitch angles, while ellipse shapes relate to distributions of runaways with larger the pitch angles. A weak dependence of the synchrotron emission measured by the camera with the RE energy, value of the q-profile at the edge, and the chosen range of wavelengths is observed. Furthermore, we find that oversimplifying the angular dependence of the SR changes the shape of the synchrotron spectra, and overestimates its amplitude by approximately 20 times for avalanching runaways and by approximately 60 times for mono-energetic distributions of runaways1.

  9. Alternative hot spot formation techniques using liquid deuterium-tritium layer inertial confinement fusion capsules

    International Nuclear Information System (INIS)

    Olson, R. E.; Leeper, R. J.

    2013-01-01

    The baseline DT ice layer inertial confinement fusion (ICF) ignition capsule design requires a hot spot convergence ratio of ∼34 with a hot spot that is formed from DT mass originally residing in a very thin layer at the inner DT ice surface. In the present paper, we propose alternative ICF capsule designs in which the hot spot is formed mostly or entirely from mass originating within a spherical volume of DT vapor. Simulations of the implosion and hot spot formation in two DT liquid layer ICF capsule concepts—the DT wetted hydrocarbon (CH) foam concept and the “fast formed liquid” (FFL) concept—are described and compared to simulations of standard DT ice layer capsules. 1D simulations are used to compare the drive requirements, the optimal shock timing, the radial dependence of hot spot specific energy gain, and the hot spot convergence ratio in low vapor pressure (DT ice) and high vapor pressure (DT liquid) capsules. 2D simulations are used to compare the relative sensitivities to low-mode x-ray flux asymmetries in the DT ice and DT liquid capsules. It is found that the overall thermonuclear yields predicted for DT liquid layer capsules are less than yields predicted for DT ice layer capsules in simulations using comparable capsule size and absorbed energy. However, the wetted foam and FFL designs allow for flexibility in hot spot convergence ratio through the adjustment of the initial cryogenic capsule temperature and, hence, DT vapor density, with a potentially improved robustness to low-mode x-ray flux asymmetry

  10. SIRIUS-P: An inertially confined direct drive laser fusion power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.; Kulcinski, G.L.; Moses, G.A.; Bruggink, D.; Engelstad, R.L.; Khater, H.Y.; Larsen, E.M.; Lovell, E.G.; MacFarlane, J.J.; Mogahed, E.A.; Peterson, R.R.; Sawan, M.E.; Wang, P.; Wittenberg, L.J.

    1993-03-01

    The SIRIUS-P conceptual design study is of a 1000 MWe laser driven inertial confinement fusion power reactor utilizing near symmetric illumination of direct drive targets. The reference driver is a KrF laser; however, any other laser capable of delivering short wavelength energy can be substituted. Sixty beams providing a total of 3.4 MJ of energy are used at a repetition rate of 6.7 Hz and a target gain of 118. The spherical chamber has an internal diameter of 6.5 m and consists of two independent components, a first wall assembly fabricated from a c/c composite and a blanket assembly made of SiC. First wall protection is provided by a xenon buffer gas at a pressure of 0.5 torr. The chamber is cooled by a flowing granular bed of solid ceramic material, TiO 2 for the first wall assembly and Li 2 O for the blanket assembly. The chamber is housed within a 42 m radius cylindrical reactor building which is 86 m high and which shares the same vacuum space as the chamber. All the laser beams are brought in at the bottom of the building, first onto a dielectrically coated final focusing mirror and finally onto a metallic grazing incidence mirror which reflects them into the chamber through beam ports open to the building. Neutron traps behind the grazing incidence mirrors are used to prolong the lifetimes of the final focusing optics. The nominal cost of electricity from this system is 65 mills/kwh assuming an 8% interest rate on capital

  11. A hybrid-drive nonisobaric-ignition scheme for inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    He, X. T., E-mail: xthe@iapcm.ac.cn [Institute of Applied Physics and Computational Mathematics, P. O. Box 8009, Beijing 100094 (China); Center for Applied Physics and Technology, HEDPS, Peking University, Beijing 100871 (China); IFSA Collaborative Innovation Center of MoE, Shanghai Jiao-Tong University, Shanghai 200240 (China); Institute of Fusion Theory and Simulation, Zhejiang University, Hangzhou 310027 (China); Li, J. W.; Wang, L. F.; Liu, J.; Lan, K.; Ye, W. H. [Institute of Applied Physics and Computational Mathematics, P. O. Box 8009, Beijing 100094 (China); Center for Applied Physics and Technology, HEDPS, Peking University, Beijing 100871 (China); IFSA Collaborative Innovation Center of MoE, Shanghai Jiao-Tong University, Shanghai 200240 (China); Fan, Z. F.; Wu, J. F. [Institute of Applied Physics and Computational Mathematics, P. O. Box 8009, Beijing 100094 (China)

    2016-08-15

    A new hybrid-drive (HD) nonisobaric ignition scheme of inertial confinement fusion (ICF) is proposed, in which a HD pressure to drive implosion dynamics increases via increasing density rather than temperature in the conventional indirect drive (ID) and direct drive (DD) approaches. In this HD (combination of ID and DD) scheme, an assembled target of a spherical hohlraum and a layered deuterium-tritium capsule inside is used. The ID lasers first drive the shock to perform a spherical symmetry implosion and produce a large-scale corona plasma. Then, the DD lasers, whose critical surface in ID corona plasma is far from the radiation ablation front, drive a supersonic electron thermal wave, which slows down to a high-pressure electron compression wave, like a snowplow, piling up the corona plasma into high density and forming a HD pressurized plateau with a large width. The HD pressure is several times the conventional ID and DD ablation pressure and launches an enhanced precursor shock and a continuous compression wave, which give rise to the HD capsule implosion dynamics in a large implosion velocity. The hydrodynamic instabilities at imploding capsule interfaces are suppressed, and the continuous HD compression wave provides main pdV work large enough to hotspot, resulting in the HD nonisobaric ignition. The ignition condition and target design based on this scheme are given theoretically and by numerical simulations. It shows that the novel scheme can significantly suppress implosion asymmetry and hydrodynamic instabilities of current isobaric hotspot ignition design, and a high-gain ICF is promising.

  12. Experimental techniques for measuring Rayleigh-Taylor instability in inertial confinement fusion (ICF)

    Energy Technology Data Exchange (ETDEWEB)

    Smalyuk, V A

    2012-06-07

    Rayleigh-Taylor (RT) instability is one of the major concerns in inertial confinement fusion (ICF) because it amplifies target modulations in both acceleration and deceleration phases of implosion, which leads to shell disruption and performance degradation of imploding targets. This article reviews experimental results of the RT growth experiments performed on OMEGA laser system, where targets were driven directly with laser light. RT instability was studied in the linear and nonlinear regimes. The experiments were performed in acceleration phase, using planar and spherical targets, and in deceleration phase of spherical implosions, using spherical shells. Initial target modulations consisted of 2-D pre-imposed modulations, and 2-D and 3-D modulations imprinted on targets by the non-uniformities in laser drive. In planar geometry, the nonlinear regime was studied using 3-D modulations with broadband spectra near nonlinear saturation levels. In acceleration-phase, the measured modulation Fourier spectra and nonlinear growth velocities are in good agreement with those predicted by Haan's model [Haan S W 1989 Phys. Rev. A 39 5812]. In a real-space analysis, the bubble merger was quantified by a self-similar evolution of bubble size distributions [Oron D et al 2001 Phys. Plasmas 8, 2883]. The 3-D, inner-surface modulations were measured to grow throughout the deceleration phase of spherical implosions. RT growth rates are very sensitive to the drive conditions, therefore they can be used to test and validate drive physics in hydrodynamic codes used to design ICF implosions. Measured growth rates of pre-imposed 2-D target modulations below nonlinear saturation levels were used to validate non-local thermal electron transport model in laser-driven experiments.

  13. First-principles investigations on ionization and thermal conductivity of polystyrene for inertial confinement fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Hu, S. X., E-mail: shu@lle.rochester.edu; Goncharov, V. N.; McCrory, R. L.; Skupsky, S. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Collins, L. A.; Kress, J. D. [Theoretical Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2016-04-15

    Using quantum molecular-dynamics (QMD) methods based on the density functional theory, we have performed first-principles investigations of the ionization and thermal conductivity of polystyrene (CH) over a wide range of plasma conditions (ρ = 0.5 to 100 g/cm{sup 3} and T = 15 625 to 500 000 K). The ionization data from orbital-free molecular-dynamics calculations have been fitted with a “Saha-type” model as a function of the CH plasma density and temperature, which gives an increasing ionization as the CH density increases even at low temperatures (T < 50 eV). The orbital-free molecular dynamics method is only used to gauge the average ionization behavior of CH under the average-atom model in conjunction with the pressure-matching mixing rule. The thermal conductivities (κ{sub QMD}) of CH, derived directly from the Kohn–Sham molecular-dynamics calculations, are then analytically fitted with a generalized Coulomb logarithm [(lnΛ){sub QMD}] over a wide range of plasma conditions. When compared with the traditional ionization and thermal conductivity models used in radiation–hydrodynamics codes for inertial confinement fusion simulations, the QMD results show a large difference in the low-temperature regime in which strong coupling and electron degeneracy play an essential role in determining plasma properties. Hydrodynamic simulations of cryogenic deuterium–tritium targets with CH ablators on OMEGA and the National Ignition Facility using the QMD-derived ionization and thermal conductivity of CH have predicted ∼20% variation in target performance in terms of hot-spot pressure and neutron yield (gain) with respect to traditional model simulations.

  14. Pulse Star Inertial Confinement Fusion Reactor: Heat transfer loop and balance-of-plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Blink, J.A.; Curlander, K.A.

    1983-01-01

    A conceptual heat transfer loop and balance-of-plant design for the Pulse Star Inertial Confinement Fusion Reactor has been investigated and the results are presented. The Pulse Star reaction vessel, a perforated steel bell jar about11 m in diameter, is immersed in Li 17 Pb 83 coolant, which flows through the perforations and forms a 1.5-m-thick plenum of droplets around a 8-m-diameter inner chamber. The bell jar and associated pumps, piping, and steam generators are contained within a 17-m-diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops, each with a flow rate of 5.5 m 3 /s, are necessary to transfer 3300 MWt of power. Liquid metal is pumped to the top of the pool, where it flows downward through eight vertical steam generators. Double-walled tubes are used in the steam generators to assure tritium containment without intermediate heat transfer loops. Each pump is a mixed flow type and has a required NPSH of 3.4 m, a speed of 278 rpm, and an impeller diameter of 1.2 m. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. The design minimizes the total cost (heat exchanger area plus pumping) for the plant lifetime. The power required for the pumps is 36 MWe. Each resulting steam generator is 12 m high and 1.6 m in diameter, with 2360 tubes. The steam generators and pumps fit easily in the pool between the reactor chamber and the pool wall

  15. Dense plasma chemistry of hydrocarbons at conditions relevant to planetary interiors and inertial confinement fusion

    Science.gov (United States)

    Kraus, Dominik

    2017-10-01

    Carbon-hydrogen demixing and subsequent diamond precipitation has been predicted to strongly participate in shaping the internal structure and evolution of icy giant planets like Neptune and Uranus. The very same dense plasma chemistry is also a potential concern for CH plastic ablator materials in inertial confinement fusion (ICF) experiments where similar conditions are present during the first compression stage of the imploding capsule. Here, carbon-hydrogen demixing may enhance the hydrodynamic instabilities occurring in the following compression stages. First experiments applying dynamic compression and ultrafast in situ X-ray diffraction at SLAC's Linac Coherent Light Source demonstrated diamond formation from polystyrene (CH) at 150 GPa and 5000 K. Very recent experiments have now investigated the influence of oxygen, which is highly abundant in icy giant planets on the phase separation process. Compressing PET (C5H4O2) and PMMA(C5H8O2), we find again diamond formation at pressures above 150 GPa and temperatures of several thousand kelvins, showing no strong effect due to the presence of oxygen. Thus, diamond precipitation deep inside icy giant planets seems very likely. Moreover, small-angle X-ray scattering (SAXS) was added to the platform, which determines an upper limit for the diamond particle size, while the width of the diffraction features provides a lower limit. We find that diamond particles of several nanometers in size are formed on a nanosecond timescale. Finally, spectrally resolved X-ray scattering is used to scale amorphous diffraction signals and allows for determining the amount of carbon-hydrogen demixing inside the compressed samples even if no crystalline diamond is formed. This whole set of diagnostics provides unprecedented insights into the nanosecond kinetics of dense plasma chemistry.

  16. 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

  17. Tritium inventory and recovery in next-step fusion devices

    International Nuclear Information System (INIS)

    Causey, R.A.; Brooks, J.N.; Federici, G.

    2002-01-01

    Future fusion devices will use tritium and deuterium fuel. Because tritium is both radioactive and expensive, it is absolutely necessary that there be an understanding of the tritium retention characteristics of the materials used in these devices as well as how to recover the tritium. There are three materials that are strong candidates for plasma-facing-material use in next-step fusion devices. These are beryllium, tungsten, and carbon. While beryllium has the disadvantage of high sputtering and low melting point (which limits its power handling capabilities in divertor areas), it has the advantages of being a low-Z material with a good thermal conductivity and the ability to get oxygen from the plasma. Due to beryllium's very low solubility for hydrogen, implantation of beryllium with deuterium and tritium results in a saturated layer in the very near-surface with limited inventory (J. Nucl. Mater. 273 (1999) 1). Unfortunately, there are nuclear reactions generated by neutrons that will breed tritium and helium in the material bulk (J. Nucl. Mater. 179 (1991) 329). This process will lead to a substantial tritium inventory in the bulk of the beryllium after long-term neutron exposure (i.e. well beyond the operation life time of a next-step reactor like ITER). Tungsten is a high-Z material that will be used in the divertor region of next-step devices (e.g. ITER) and possibly as a first wall material in later devices. The divertor is the preferred location for tungsten use because net erosion is very low there due to low sputtering and high redeposition. While experiments are still continuing on tritium retention in tungsten, present data suggest that relatively low tritium inventories will result with this material (J. Nucl. Mater. 290-293 (2001) 505). For tritium inventories, carbon is the problem material. Neutron damage to the graphite can result in substantial bulk tritium retention (J. Nucl. Mater. 191-194 (1992) 368), and codeposition of the sputtered carbon

  18. Enhanced confinement phenomenology in magnetic fusion plasmas: Is it unique in physics?

    International Nuclear Information System (INIS)

    Dendy, R.O.

    2002-01-01

    There is substantial experimental evidence that simple diffusive models for turbulent transport are insufficient to produce all the confinement phenomena observed in tokamaks. This paper reports on the emerging linkage between rapid, nonlocal, nondiffusive transport and overall confinement phenomenology including edge pedestals, enhanced confinement, ELMs, and internal transport barriers. Modern statistical physics techniques are used to construct simple models that generate many of the distinctive elements of global tokamak confinement phenomenology. The similarities are deep and are quantified. These results imply that current observations of avalanching transport in tokamaks may be deeply linked to the fundamental global features of tokamak plasma confinement. (author)

  19. Profile consistency, anomalous electron thermal conduction, and confinement analysis of tokamak devices

    International Nuclear Information System (INIS)

    Qu Wenxiao

    1992-01-01

    Assuming that there exists a position in the tokamak plasma where the energy transport is dominated by local anomalous electron thermal conduction and taking advantage of the basic experimental result usually referred to as profile consistency, the authors obtain a more convincing approach to the description of the confinement property of tokamak devices without touching upon the physical mechanism of global plasma energy transport. 8 refs

  20. Reducing the tritium inventory in waste produced by fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Pamela, J., E-mail: jerome.pamela@cea.fr [CEA, Agence ITER-France, F-13108 Saint-Paul-lez-Durance (France); Decanis, C. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Canas, D. [CEA, DEN/DADN, Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Liger, K.; Gaune, F. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2015-04-15

    Highlights: • Fusion devices including ITER will generate tritiated waste, some of which will need to be detritiated before disposal. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Incineration is very attractive for VLLW and possibly SL-LILW soft housekeeping waste, since it offers higher tritium and waste volume reduction than the alternative thermal treatment technique. • For metallic waste, further R&D efforts should be made to optimize tritium release management and minimize the need for interim storage. - Abstract: The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted.

  1. Current fusion plasma theory grant: Task I, Magnetic confinement fusion plasma theory: Final report, December 1, 1987--November 14, 1988

    International Nuclear Information System (INIS)

    Callen, J.D.

    1988-07-01

    The research performed under this grant over the current 11-1/2 month period has concentrated on key tokamak plasma confinement and heating theory issues: extensions of neoclassical MHD; viscosity coefficients and transport; nonlinear resistive MHD simulations of Tokapole II plasmas; ICRF and edge plasma interactions; energy confinement degradation due to macroscopic phenomena; and coordination of a new transport initiative. Progress and publications in these areas are briefly summarized in this report. 21 refs

  2. OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

    Science.gov (United States)

    Batistoni, P; Villari, R; Obryk, B; Packer, L W; Stamatelatos, I E; Popovichev, S; Colangeli, A; Colling, B; Fonnesu, N; Loreti, S; Klix, A; Klosowski, M; Malik, K; Naish, J; Pillon, M; Vasilopoulou, T; De Felice, P; Pimpinella, M; Quintieri, L

    2017-10-05

    The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  3. Interrupter and hybrid-switch testing for fusion devices

    International Nuclear Information System (INIS)

    Parsons, W.M.; Warren, R.W.; Honig, E.M.; Lindsay, J.D.G.; Bellamo, P.; Cassel, R.L.

    1979-01-01

    This paper discusses recent and ongoing switch testing for fusion devices. The first part describes testing for the TFTR ohmic-heating circuit. In this set of tests, which simulated the stresses produced during a plasma initiation pulse, circuit breakers were required to interrupt a current of 24 kA with an associated recovery voltage of 25 kV. Two interrupter systems were tested for over 1000 operations each, and both appear to satisfy TFTR requirements. The second part discusses hybrid-switch development for superconducting coil protection. These switching systems must be capable of carrying large currents on a continuous basis as well as performing interruption duties. The third part presents preliminary results on an early-counterpulse technique applied to vacuum interrupters. Implementation of this technique has resulted in large increases in interruptible current as well as a marked reduction in contact erosion

  4. Speckle interferometry application for erosion measurements in fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Gauthier, E.; Roupillard, R. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2003-07-01

    In order to measure erosion/redeposition in fusion devices, a new diagnostic based on speckle interferometry is investigated. First experiments performed on carbon fibre composite (CFC) materials have shown that this technique is able to measure a modification of the surface in the range of 1 {mu}m. Further experiments have been performed on different materials using a second wavelength in order to carry out 3-dimensional measurements of the surface and to increase the dynamic range of the depth measurement. A diagnostic, based on two-wavelength TV-holography to measure in situ erosion/redeposition during long duration discharges on the CIEL limiter in Tore Supra, is under development at CEA Cadarache. (authors)

  5. 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

  6. Erosion and redeposition at the vessel walls in fusion devices

    International Nuclear Information System (INIS)

    Naujoks, D.; Behrisch, R.

    1995-01-01

    The plasma induced erosion and redeposition at the vessel walls in today's fusion devices have been investigated both with the computer simulation code ERO, and in experiments. Well prepared carbon probes with implanted and evaporated markers in the surface layers have been exposed in the scrape-off layer (SOL) of several tokamaks such as JET, TEXTOR and ASDEX-Upgrade. The main plasma parameters (electron density and temperature, impurity concentration in the SOL) are simultaneously determined. After exposure to single plasma discharges, erosion and redeposition of the marker material were measured by surface layer analysis with MeV ion beam techniques. The experimental results were compared with the results from the ERO code. The measured erosion/redeposition could be described with ERO, which takes into account the impurity concentration in the SOL, the dynamical change of the surface composition (causing a modification of the sputtering yield during the exposure) and ExB drift effects. ((orig.))

  7. 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.

  8. Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

    Science.gov (United States)

    Woolley, Robert D.

    2002-01-01

    A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

  9. 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

  10. Suppression of Stimulated Brillouin Scattering in multiple-ion species inertial confinement fusion Hohlraum Plasmas

    International Nuclear Information System (INIS)

    Neumayer, P

    2007-01-01

    A long-standing problem in the field of laser-plasma interactions is to successfully employ multiple-ion species plasmas to reduce stimulated Brillouin scattering (SBS) in inertial confinement fusion (ICF) hohlraum conditions. Multiple-ion species increase significantly the linear Landau damping for acoustic waves. Consequently, recent hohlraum designs for indirect-drive ignition on the National Ignition Facility investigate wall liner material options so that the liner gain for parametric instabilities will be below threshold for the onset SBS. Although the effect of two-ion species plasmas on Landau damping has been directly observed with Thomson scattering, early experiments on SBS in these plasmas have suffered from competing non-linear effects or laser beam filamentation. In this study, a reduction of SBS scattering to below the percent level has been observed in hohlraums at Omega that emulate the plasma conditions in an indirect drive ICF experiments. These experiments have measured the laser-plasma interaction processes in ignition-relevant high-electron temperature regime demonstrating Landau damping as a controlling process for SBS. The hohlraums have been filled with various fractions of CO 2 and C 3 H 8 varying the ratio of the light (H) to heavy (C and O) ion density from 0 to 2.6. They have been heated by 14.5 kJ of 351-nm light, thus increasing progressively Landau damping by an order of magnitude at constant electron density and temperature. A delayed 351-nm interaction beam, spatially smoothed to produce a 200-(micro)m laser spot at best focus, has propagated along the axis of the hohlraum. The backscattered light, both into the lens and outside, the transmitted light through the hohlraum plasma and the radiation temperature of the hohlraum has been measured. For ignition relevant laser intensities (3-9 10 14 Wcm -2 ), we find that the SBS reflectivity scales as predicted with Landau damping from >30% to <1%. Simultaneously, the hohlraum radiation

  11. 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

  12. First-principles thermal conductivity of warm-dense deuterium plasmas for inertial confinement fusion applications.

    Science.gov (United States)

    Hu, S X; Collins, L A; Boehly, T R; Kress, J D; Goncharov, V N; Skupsky, S

    2014-04-01

    Thermal conductivity (κ) of both the ablator materials and deuterium-tritium (DT) fuel plays an important role in understanding and designing inertial confinement fusion (ICF) implosions. The extensively used Spitzer model for thermal conduction in ideal plasmas breaks down for high-density, low-temperature shells that are compressed by shocks and spherical convergence in imploding targets. A variety of thermal-conductivity models have been proposed for ICF hydrodynamic simulations of such coupled and degenerate plasmas. The accuracy of these κ models for DT plasmas has recently been tested against first-principles calculations using the quantum molecular-dynamics (QMD) method; although mainly for high densities (ρ > 100 g/cm3), large discrepancies in κ have been identified for the peak-compression conditions in ICF. To cover the wide range of density-temperature conditions undergone by ICF imploding fuel shells, we have performed QMD calculations of κ for a variety of deuterium densities of ρ = 1.0 to 673.518 g/cm3, at temperatures varying from T = 5 × 103 K to T = 8 × 106 K. The resulting κQMD of deuterium is fitted with a polynomial function of the coupling and degeneracy parameters Γ and θ, which can then be used in hydrodynamic simulation codes. Compared with the "hybrid" Spitzer-Lee-More model currently adopted in our hydrocode lilac, the hydrosimulations using the fitted κQMD have shown up to ∼20% variations in predicting target performance for different ICF implosions on OMEGA and direct-drive-ignition designs for the National Ignition Facility (NIF). The lower the adiabat of an imploding shell, the more variations in predicting target performance using κQMD. Moreover, the use of κQMD also modifies the shock conditions and the density-temperature profiles of the imploding shell at early implosion stage, which predominantly affects the final target performance. This is in contrast to the previous speculation that κQMD changes mainly the

  13. Inertial Confinement Fusion quarterly report April-June 1999, volume 9, number 3

    International Nuclear Information System (INIS)

    MacGowan, B

    1999-01-01

    enhance significantly our understanding of the details of how radiation and matter interact with one another in contexts relevant both to inertial confinement fusion (ICF) and to other applications. The first of these two articles describes these experiments. The second one describes the application of such radiative heating techniques to development of a ''piston'' for shocklessly accelerating materials. This new experimental technique, first developed on Nova, shows promise as a way to diagnose the development of acceleration-driven hydrodynamic instabilities in the compressible regime, a longstanding ICF problem that is currently only poorly understood

  14. Hydrodynamic modelling of the shock ignition scheme for inertial confinement fusion

    International Nuclear Information System (INIS)

    Vallet, Alexandra

    2014-01-01

    The shock ignition concept in inertial confinement fusion uses an intense power spike at the end of an assembly laser pulse. The key features of shock ignition are the generation of a high ablation pressure, the shock pressure amplification by at least a factor of a hundred in the cold fuel shell and the shock coupling to the hot-spot. In this thesis, new semi-analytical hydrodynamic models are developed to describe the ignitor shock from its generation up to the moment of fuel ignition. A model is developed to describe a spherical converging shock wave in a pre-heated hot spot. The self-similar solution developed by Guderley is perturbed over the shock Mach number Ms ≥≥1. The first order correction accounts for the effects of the shock strength. An analytical ignition criterion is defined in terms of the shock strength and the hot-spot areal density. The ignition threshold is higher when the initial Mach number of the shock is lower. A minimal shock pressure of 20 Gbar is needed when it enters the hot-spot. The shock dynamics in the imploding shell is then analyzed. The shock is propagating into a non inertial medium with a high radial pressure gradient and an overall pressure increase with time. The collision with a returning shock coming from the assembly phase enhances further the ignitor shock pressure. The analytical theory allows to describe the shock pressure and strength evolution in a typical shock ignition implosion. It is demonstrated that, in the case of the HiPER target design, a generation shock pressure near the ablation zone on the order of 300-400 Mbar is needed. An analysis of experiments on the strong shock generation performed on the OMEGA laser facility is presented. It is shown that a shock pressure close to 300 Mbar near the ablation zone has been reached with an absorbed laser intensity up to 2 * 10 15 W:cm -2 and a laser wavelength of 351 nm. This value is two times higher than the one expected from collisional laser absorption only

  15. Theoretical studies of turbulence and anomalous transport in toroidal confinement devices

    International Nuclear Information System (INIS)

    Terry, P.W.

    1990-01-01

    The research performed under this grant during the current year has focused on key issues with respect to turbulence and transport in toroidal confinement devices. This work includes theoretical and computational studies of electron thermal confinement which have concentrated on the role of sheared poloidal flow in suppressing turbulence and transport, trapped ion convective cell turbulence and microtearing turbulence; analytical studies of anomalous particle transport and pinch mechanisms, and comparison with experimental measurement; development of the theory of self-consistent radial transport of field-aligned momentum in the tokamak and RFP; and work on other topics (ion temperature gradient driven turbulence, RFP fluctuation theory, coherent structures). Progress and publications in these areas are briefly summarized in this report. 20 refs

  16. Spherical ion oscillations in a positive polarity gridded inertial-electrostatic confinement device

    Energy Technology Data Exchange (ETDEWEB)

    Bandara, R.; Khachan, J. [Plasma Physics, School of Physics, University of Sydney, Camperdown, New South Wales 2006 (Australia)

    2013-07-15

    A pulsed, positive polarity gridded inertial electrostatic confinement device has been investigated experimentally, using a differential emissive probe and potential traces as primary diagnostics. Large amplitude oscillations in the plasma current and plasma potential were observed within a microsecond of the discharge onset, which are indicative of coherent ion oscillations about a temporarily confined excess of recirculating electron space charge. The magnitude of the depth of the potential well in the established virtual cathode was determined using a differential emissive Langmuir probe, which correlated well to the potential well inferred from the ion oscillation frequency for both hydrogen and argon experiments. It was found that the timescale for ion oscillation dispersion is strongly dependent on the neutral gas density, and weakly dependent on the peak anode voltage. The cessation of the oscillations was found to be due to charge exchange processes converting ions to high velocity neutrals, causing the abrupt de-coherence of the oscillations through an avalanche dispersion in phase space.

  17. Spherical ion oscillations in a positive polarity gridded inertial-electrostatic confinement device

    Science.gov (United States)

    Bandara, R.; Khachan, J.

    2013-07-01

    A pulsed, positive polarity gridded inertial electrostatic confinement device has been investigated experimentally, using a differential emissive probe and potential traces as primary diagnostics. Large amplitude oscillations in the plasma current and plasma potential were observed within a microsecond of the discharge onset, which are indicative of coherent ion oscillations about a temporarily confined excess of recirculating electron space charge. The magnitude of the depth of the potential well in the established virtual cathode was determined using a differential emissive Langmuir probe, which correlated well to the potential well inferred from the ion oscillation frequency for both hydrogen and argon experiments. It was found that the timescale for ion oscillation dispersion is strongly dependent on the neutral gas density, and weakly dependent on the peak anode voltage. The cessation of the oscillations was found to be due to charge exchange processes converting ions to high velocity neutrals, causing the abrupt de-coherence of the oscillations through an avalanche dispersion in phase space.

  18. Electromagnetic device for confining a liquid metal and regulating the flow rate

    International Nuclear Information System (INIS)

    Garnier, Marcel; Moreau, R.J.

    1977-01-01

    The description is given of a device for confining a liquid metal jet, characterized in that it comprises in combination, at the jet outlet nozzle, (a) means for producing a high pressure in the jet composed of a coil around the nozzle and located on its outlet, in combination with facilities for passing a high frequency alternating current through the coil and (b) means for suppressing this high pressure. It is stated that this device has many uses, particularly for allowing the use of a relatively large diameter orifice, hence not subject to the risk of clogging, in order to produce a jet with a relatively small diameter. This invention particularly concerns the application of this device for regulating a flow of liquid metal at an outlet orifice located at the lower end of a receptacle containing this liquid metal [fr

  19. Safety analysis and evaluation of the next fusion device

    International Nuclear Information System (INIS)

    Kobayashi, Shigetada; Honda, Tsutomu; Ohmura, Hiroshi; Kawai, Masayoshi; Shimizu, Takeshi; Yamaoka, Mitsuaki; Nakahara, Katsuhiko; Seki, Yasushi.

    1988-12-01

    As a part of safety evaluation, a probabilistic risk assessment (PRA) has been attempted for the Next Fusion Device system. Among the various events related to safety, a number of representative events have been selected for assessment, from the events in normal operation state, repair and maintenance state and accidental state. In the first chapter, in order to conduct the probabilistic risk assessment of the whole Fusion Experimental Reactor (FER), the data base required for the analysis was investigated in 1.1, the results on the failure mode and effects analysis (FMEA), accident sequence, radioactive inventory leakage flow path, event tree analysis (ETA) and fault tree analysis (FTA) were summarized in 1.2 to 1.5, respectively. Based on these results, accident initiating events were evaluated in 1.6, and overall risk was assessed in 1.7 and the tasks for the future were summarized in 1.8. It is important to analyze and evaluate various events during normal operations, repair and maintenance and accidents. However, due to the large uncertainties in the modeling of phenomena or the data base, there are many events for which realistic analyses are difficult. Three such events were selected and studied in chapter two. In 2.1, the temperature rise in the reactor structure after the Loss-of-Coolant-Accident caused by the decay heat under various heat removal conditions were investigated. In 2.2, the radiation dose of personnel during repair and maintenance period caused by the release of activated dust were estimated. Lastly, in 2.3 tritium behavior in the stainless steel first wall and graphite armour were studied. (author)

  20. Fusion plasma theory. Task I. Magnetic confinement fusion plasma theory. Annual progress report, January 1, 1982-December 31, 1982

    International Nuclear Information System (INIS)

    Callen, J.D.

    1982-08-01

    The research on this contract over the past year has concentrated on some key tandem mirror confinement and heating issues (barrier trapping current, rf heating, low mode number stability) and on developing a comprehensive neoclassical transport theory for nonaxisymmetric toroidal plasmas (e.g., stellarators). Progress in these and some other miscellaneous areas are summarized briefly in this progress report