WorldWideScience

Sample records for confinement fusion power

  1. Volume ignition for inertial confinement fusion tested by different stopping power models

    Energy Technology Data Exchange (ETDEWEB)

    Hora, H.; Eliezer, S.; Honrubia, J.J.; Hoepfl, R.; Martinez-Val, J.M.; Peria, M.; Velarde, G. [European Network High Energy Density Matter-Polytechnic University of Madrid (Spain)

    1996-05-01

    Within the low cost energy alternatives to carbon burning, nuclear fusion based on inertial confinement offers the earliest solution. The existing difficulties with spark (hot spot) ignition are reduced by volume ignition. Two basically different computations with different stopping power models are compared and only minor differences in the fusion gains established. An effective density-radius criterion results in comparably very high values if the strong effect of self-heat is included. {copyright} {ital 1996 American Institute of Physics.}

  2. Lasers and power systems for inertial confinement fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stark, E.E. Jr.

    1978-01-01

    After discussing the role of lasers in ICF and the candidate lasers, several important areas of technology requirements are discussed. These include the beam transport system, the pulsed power system and the gas flow system. The system requirements, state of the art, as well as needs and prospects for new technology developments are given. Other technology issues and promising developments are described briefly.

  3. Inertial confinement fusion reaction chamber and power conversion system study

    Energy Technology Data Exchange (ETDEWEB)

    Maya, I.; Schultz, K.R.; Battaglia, J.M.; Buksa, J.J.; Creedson, R.L.; Erlandson, O.D.; Levine, H.E.; Roelant, D.F.; Sanchez, H.W.; Schrader, S.A.

    1984-09-01

    GA Technologies has developed a conceptual ICF reactor system based on the Cascade rotating-bed reaction chamber concept. Unique features of the system design include the use of low activation SiC in a reaction chamber constructed of box-shaped tiles held together in compression by prestressing tendons to the vacuum chamber. Circulating Li/sub 2/O granules serve as the tritium breeding and energy transport material, cascading down the sides of the reaction chamber to the power conversion system. The total tritium inventory of the system is 6 kg; tritium recovery is accomplished directly from the granules via the vacuum system. A system for centrifugal throw transport of the hot Li/sub 2/O granules from the reaction chamber to the power conversion system has been developed. A number of issues were evaluated during the course of this study. These include the response of first-layer granules to the intense microexplosion surface heat flux, cost effective fabrication of Li/sub 2/O granules, tritium inventory and recovery issues, the thermodynamics of solids-flow options, vacuum versus helium-medium heat transfer, and the tradeoffs of capital cost versus efficiency for alternate heat exchange and power conversion system option. The resultant design options appear to be economically competitive, safe, and environmentally attractive.

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

    Science.gov (United States)

    Orth, Charles D.; Klein, Gail; Sercel, Joel; Hoffman, Nathan; Murray, Kathy; Chang-Diaz, Franklin

    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. In this paper a concept is produced 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.

  5. Magnetic-confinement fusion

    Science.gov (United States)

    Ongena, J.; Koch, R.; Wolf, R.; Zohm, H.

    2016-05-01

    Our modern society requires environmentally friendly solutions for energy production. Energy can be released not only from the fission of heavy nuclei but also from the fusion of light nuclei. Nuclear fusion is an important option for a clean and safe solution for our long-term energy needs. The extremely high temperatures required for the fusion reaction are routinely realized in several magnetic-fusion machines. Since the early 1990s, up to 16 MW of fusion power has been released in pulses of a few seconds, corresponding to a power multiplication close to break-even. Our understanding of the very complex behaviour of a magnetized plasma at temperatures between 150 and 200 million °C surrounded by cold walls has also advanced substantially. This steady progress has resulted in the construction of ITER, a fusion device with a planned fusion power output of 500 MW in pulses of 400 s. ITER should provide answers to remaining important questions on the integration of physics and technology, through a full-size demonstration of a tenfold power multiplication, and on nuclear safety aspects. Here we review the basic physics underlying magnetic fusion: past achievements, present efforts and the prospects for future production of electrical energy. We also discuss questions related to the safety, waste management and decommissioning of a future fusion power plant.

  6. Inertial Confinement fusion targets

    Science.gov (United States)

    Hendricks, C. D.

    1982-01-01

    Inertial confinement fusion (ICF) targets are made as simple flat discs, as hollow shells or as complicated multilayer structures. Many techniques were 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.

  7. 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

  8. Inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Powers, L.; Condouris, R.; Kotowski, M.; Murphy, P.W. (eds.)

    1992-01-01

    This issue of the ICF Quarterly contains seven articles that describe recent progress in Lawrence Livermore National Laboratory's ICF program. The Department of Energy recently initiated an effort to design a 1--2 MJ glass laser, the proposed National Ignition Facility (NIF). These articles span various aspects of a program which is aimed at moving forward toward such a facility by continuing to use the Nova laser to gain understanding of NIF-relevant target physics, by developing concepts for an NIF laser driver, and by envisioning a variety of applications for larger ICF facilities. This report discusses research on the following topics: Stimulated Rotational Raman Scattering in Nitrogen; A Maxwell Equation Solver in LASNEX for the Simulation of Moderately Intense Ultrashort Pulse Experiments; Measurements of Radial Heat-Wave Propagation in Laser-Produced Plasmas; Laser-Seeded Modulation Growth on Directly Driven Foils; Stimulated Raman Scattering in Large-Aperture, High-Fluence Frequency-Conversion Crystals; Fission Product Hazard Reduction Using Inertial Fusion Energy; Use of Inertial Confinement Fusion for Nuclear Weapons Effects Simulations.

  9. Overview of the VISTA Spacecraft Concept Powered by Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Orth, C D

    2000-11-21

    VISTA was conceived through a detailed systems analysis as a viable, realistic, and defensible spacecraft concept based on advanced ICF technology but existing or near-term technology for other systems. It is a conical self-contained single-stage piloted spacecraft in which a magnetic thrust chamber directs the plasma emissions from inertial confinement fusion (ICF) targets into a rearward exhaust. VISTA's propulsion system is therefore unique because it is based on (1) a rather mature technology (ICF), which is known to work with sufficient driver input; (2) direct heating of all expellant by the fusion process, thus providing high mass flow rates without significant degradation of jet efficiency; and (3) 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. VISTA therefore has inherently high power/mass ratios and high specific impulses. With advanced ICF technology, ultra-fast roundtrips (RTs) to objects within the solar system are possible (e.g., {ge}145 days RT to Mars, {ge}7 years RT to Pluto). Such short-duration missions are imperative to minimize the human physiological deteriorations arising from zero gravity and the cosmic-radiation. In addition, VISTA offers on-board artificial gravity and propellant-based shielding from cosmic rays, thus reducing the physiological deteriorations to insignificant levels. In this paper, we give an overview of the various vehicle systems for this concept, estimate the general missions performance capabilities for interplanetary missions, and describe in detail the performance for the baseline mission of a piloted roundtrip to Mars with a 100-ton payload. Items requiring further research include a reduction of the wet mass from its baseline value of 6,000 metric tons, and the development of fast ignition or its equivalent to provide target gains in excess of several hundred. With target gains well

  10. Physics of magnetic confinement fusion

    Directory of Open Access Journals (Sweden)

    Wagner F.

    2013-06-01

    Full Text Available Fusion is the energy source of the universe. The local conditions in the core of the Sun allow the transfer of mass into energy, which is finally released in the form of radiation. Technical fusion melts deuterons and tritons to helium releasing large amounts of energy per fusion process. Because of the conditions for fusion, which will be deduced, the fusion fuel is in the plasma state. Here we report on the confinement of fusion plasmas by magnetic fields. Different confinement concepts — tokamaks and stellarators — will be introduced and described. The first fusion reactor, ITER, and the most modern stellarator, Wendelstein 7-X, are under construction. Their basic features and objectives will be presented.

  11. Magnetic confinement fusion energy research

    Energy Technology Data Exchange (ETDEWEB)

    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/sup 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.

  12. Effects of alpha stopping power modelling on the ignition threshold in a directly-driven inertial confinement fusion capsule

    Science.gov (United States)

    Temporal, Mauro; Canaud, Benoit; Cayzac, Witold; Ramis, Rafael; Singleton, Robert L.

    2017-05-01

    The alpha-particle energy deposition mechanism modifies the ignition conditions of the thermonuclear Deuterium-Tritium fusion reactions, and constitutes a key issue in achieving high gain in Inertial Confinement Fusion implosions. One-dimensional hydrodynamic calculations have been performed with the code Multi-IFE [R. Ramis, J. Meyer-ter-Vehn, Comput. Phys. Commun. 203, 226 (2016)] to simulate the implosion of a capsule directly irradiated by a laser beam. The diffusion approximation for the alpha energy deposition has been used to optimize three laser profiles corresponding to different implosion velocities. A Monte-Carlo package has been included in Multi-IFE to calculate the alpha energy transport, and in this case the energy deposition uses both the LP [C.K. Li, R.D. Petrasso, Phys. Rev. Lett. 70, 3059 (1993)] and the BPS [L.S. Brown, D.L. Preston, R.L. Singleton Jr., Phys. Rep. 410, 237 (2005)] stopping power models. Homothetic transformations that maintain a constant implosion velocity have been used to map out the transition region between marginally-igniting and high-gain configurations. The results provided by the two models have been compared and it is found that - close to the ignition threshold - in order to produce the same fusion energy, the calculations performed with the BPS model require about 10% more invested energy with respect to the LP model.

  13. Diagnostics for magnetic confinement fusion research

    Science.gov (United States)

    Weller, Arthur

    2010-11-01

    Significant progress towards the development of an attractive fusion energy source based on magnetic or inertial plasma confinement has been achieved within the international fusion energy program. High-level diagnostics capabilities are required to characterize fusion plasmas and to achieve a sound physics basis to design a fusion power plant. A large variety of different measuring techniques is used, most of them based on the detection of electromagnetic radiation in a wide range of wavelengths or of particles emitted from the plasma. Active probing by laser and particle beams permits to measure local plasma parameters directly, whereas passive measurements and imaging methods require unfolding and tomographic reconstruction techniques in order to obtain the spatial source distribution. Most diagnostics systems are limited in the accessible parameter range, in accuracy, temporal and spatial resolution, energy resolution and hardiness in a harsh environment, so that redundancy and complementarity of different methods is desirable. A considerable synergy exists between plasma diagnostics for fusion and astrophysics research. In particular, novel imaging detectors developed for the observation of astrophysical objects may be applied to fusion devices, too. An overview of diagnostics requirements, measuring techniques and selected results are presented with an emphasis of imaging diagnostics in toroidal magnetic fusion devices.

  14. Inertial Confinement Fusion and the National Ignition Facility (NIF)

    Energy Technology Data Exchange (ETDEWEB)

    Ross, P.

    2012-08-29

    Inertial confinement fusion (ICF) seeks to provide sustainable fusion energy by compressing frozen deuterium and tritium fuel to extremely high densities. The advantages of fusion vs. fission are discussed, including total energy per reaction and energy per nucleon. The Lawson Criterion, defining the requirements for ignition, is derived and explained. Different confinement methods and their implications are discussed. The feasibility of creating a power plant using ICF is analyzed using realistic and feasible numbers. The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is shown as a significant step forward toward making a fusion power plant based on ICF. NIF is the world’s largest laser, delivering 1.8 MJ of energy, with a peak power greater than 500 TW. NIF is actively striving toward the goal of fusion energy. Other uses for NIF are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-01

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

  16. An antiproton catalyst for inertial confinement fusion propulsion

    Science.gov (United States)

    Lewis, Raymond A.; Newton, Richard; Smith, Gerald A.; Toothacker, William S.; Kanzleiter, Randall J.

    1990-01-01

    This paper discusses the concept of an inertial confinement fusion propulsion system involving an antiproton catalyst (for antiproton-induced fission). It is argued that, when the two processes, fusion and antimatter annihilation, are combined into one system, a viable candidate propulsion system for planetary exploration emerges. It is shown that as much as 7.6 GW of power, well within the requrements for interplanetary travel, can be achieved using existing driver technologies and available quantities of antiprotons.

  17. Inertial-Electrostatic Confinement (IEC) Fusion for Space Propulsion

    Science.gov (United States)

    Nadler, Jon

    1999-01-01

    An Inertial-Electrostatic Confinement (IEC) device was assembled at the Marshall Space Flight Center (MSFC) Propulsion Research Center (PRC) to study the possibility of using EEC technology for deep space propulsion and power. Inertial-Electrostatic Confinement is capable of containing a nuclear fusion plasma in a series of virtual potential wells. These wells would substantially increase plasma confinement, possibly leading towards a high-gain, breakthrough fusion device. A one-foot in diameter IEC vessel was borrowed from the Fusion Studies Laboratory at the University of Illinois@Urbana-Champaign for the summer. This device was used in initial parameterization studies in order to design a larger, actively cooled device for permanent use at the PRC.

  18. Inertial Confinement Fusion R&D and Nuclear Proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Robert J. Goldston

    2011-04-28

    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&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.

  19. 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. Particula...... are highly relevant for developing predictive models of the transition, essential for understanding and optimizing future fusion power reactors........ Particularly, the slow transition with an intermediate dithering phase is well reproduced at proper parameters. The model recovers the power threshold for the L-H transition as well as the decrease in power threshold switching from single to double null configuration observed experimentally. The results...

  20. 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...

  1. 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.

  2. Ultra-Compact Electrostatic Confinement Fusion Device

    Science.gov (United States)

    Young, Garrett

    2017-10-01

    A unique, linear dual-beam configuration with an internal volume of 144 cc was simulated and operated. Deuteron ion paths were simulated using Mathematica and the electric field distribution was optimized relative to convergence density, potential well efficiency, and confinement time. The resulting cathode design is a departure from conventional systems, with gradual conical surfaces. The simulated trajectories correlated well to the observed operation, evidenced by two principle factors. First, the high transparency of the cathode due to the focused beams allowed for >1 kW operation without duration-limiting temperature rise. Second, when compared to inertial electrostatic configurations, the constructed device achieved record steady-state D-D fusion rates per internal volume including 3.7E +4 fusions/sec/cc at 52 kV applied potential and 28 mTorr operating pressure.

  3. Diamond Ablators for Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Biener, J; Mirkarimi, P B; Tringe, J W; Baker, S L; Wang, Y M; Kucheyev, S O; Teslich, N E; Wu, K J; Hamza, A V; Wild, C; Woerner, E; Koidl, P; Bruehne, K; Fecht, H

    2005-06-21

    Diamond has a unique combination of physical properties for the inertial confinement fusion ablator application, such as appropriate optical properties, high atomic density, high yield strength, and high thermal conductivity. Here, we present a feasible concept to fabricate diamond ablator shells. The fabrication of diamond capsules is a multi-step process, which involves diamond chemical vapor deposition on silicon mandrels followed by polishing, microfabrication of holes, and removing of the silicon mandrel by an etch process. We also discuss the pros and cons of coarse-grained optical quality and nanocrystalline chemical vapor deposition diamond films for the ablator application.

  4. Indirect drive targets for fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Amendt, Peter A.; Miles, Robin R.

    2016-10-11

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

  5. An effective electrostatic-confinement based fusion approach

    CERN Document Server

    Paul, R K

    2013-01-01

    The paper reports a new electrostatic-confinement based fusion approach, where, a new non-equilibrium distribution function for an ion-beam, compressed by an external electric force, has been derived. This distribution function allows the system to possess appreciably low and insignificant thermal energy irrespective of the energy per particle. The spread in the energy among the particles is attributed to the collisions in presence of the external force, whereas; for equilibrium, the spreading in energy is due to the absence of the force. The reactivity for a deuterium-deuterium fusion, using the proposed distribution function, has been computed. It is shown that the fusion time is comparable to the energy confinement time, collision time and transit time of the ion for beam energy greater than 160 keV. The estimated energy gain Q (ratio of fusion power to the power consumed by the system) is around 10 for beam energy 160 keV and ion density 1018 cm-3. The energy loss due to particle scattering is estimated a...

  6. 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.

  7. High-gain volume ignition for inertial confinement fusion (ICF)

    Energy Technology Data Exchange (ETDEWEB)

    Hora, H. [University of New South Wales, Kensington 2033 (Australia); Eliezer, S. [University of New South Wales, Kensington 2033 (Australia)]|[SOREQ Nuclear Research Centre, Yavne (Israel); Martinez-Val, J.M. [University of New South Wales, Kensington 2033 (Australia)]|[Instituto de Fusion Nuclear, ETSI Industriales, Madrid (Spain); Miley, G.H. [University of New South Wales, Kensington 2033 (Australia)]|[Fusion Studies Laboratory, University of Illinois, Urbana, Illinois (United States)

    1994-10-05

    Since 1976, volume ignition calculations of pellet-fueled inertial confinement fusion have shown very high fusion gains due to the strong temperature increases caused by self-heating. This phenomenon was first reported in 1978 (Hora & Ray) and subsequently named the ``Wheeler modes.`` The very low optimum initial temperatures ({approx}1 keV) and the fuel burn of up to 80% permit gains of {gt}100, for compression of D-T to 2000--10,000 times the solid state for total input driver energy of a few MJ per pulse. The simple, smooth density and temperature profiles---unlike those of spark ignition---indicate it may be a faster and easier route to pellet fueling for a fusion power reactor capable of low-cost energy production. {copyright} 1994 {ital American} {ital Institute} {ital of} {ital Physics}

  8. The VISTA spacecraft: Advantages of ICF (Inertial Confinement Fusion) for interplanetary fusions propulsion applications

    Science.gov (United States)

    Orth, Charles D.; Klein, Gail; Sercel, Joel; Hoffman, Nate; Murray, Kathy; Chang-Diaz, Franklin

    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.

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

    Science.gov (United States)

    Fillo, J. A.

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

  10. BOOK REVIEW: Advanced Diagnostics for Magnetic and Inertial Confinement Fusion

    Science.gov (United States)

    Stott, PE; Wootton, A.; Gorini, G.; Sindoni, E.; Batani, D.

    2003-02-01

    parameters approach ignition. Spectroscopic systems and their recent developments are well represented, whereas edge diagnostics are somewhat thin on the ground. A dedicated section is devoted to the latest tests on radiation effects and technological issues. The problems of damage to optical components and the difficulties presented by the determination of the tritium inventory are described. In the last part, the new diagnostic systems of the most recent experiments (under construction or recently operated) are reported. Various aspects of some diagnostics not included in the three previous sections are also covered, with particular emphasis on microwaves and infrared diagnostics. The book is well suited for specialists and, more generally, for people involved in nuclear fusion, who need information about the most recent developments in the field of plasma diagnostics. The papers cover many aspects of the challenges and possible solutions for performing measurements in fusion machines approaching reactor conditions. On the other hand, the contributions are in general quite advanced and would be challenging for people without a significant background in plasma diagnostics and nuclear fusion. The quality of the paper is more than satisfactory both from the point of view of clarity and of graphics. Moreover, at the beginning of the book, several papers make a considerable effort to put diagnostic issues in the wider context of present day nuclear fusion research. For those topics, which are too involved to be completely described in a conference contribution, in general adequate references are provided for deeper investigation. A Murari Approximately one third of the papers included in this volume deal with diagnostics related to inertial confinement fusion plasmas (i.e., laser-produced plasmas and pulsed-power). These papers discuss recent developments in charged particle diagnostics, neutron diagnostics, optical and x-ray measurements along with laser and particle probing

  11. Inertially confined fusion using heavy ion drivers

    Energy Technology Data Exchange (ETDEWEB)

    Herrmannsfeldt, W.B. (Stanford Linear Accelerator Center, Menlo Park, CA (United States)); Bangerter, R.O. (Lawrence Berkeley Lab., CA (United States)); Bock, R. (Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany)); Hogan, W.J.; Lindl, J.D. (Lawrence Livermore National Lab., CA (United States))

    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.

  12. Inertially confined fusion using heavy ion drivers

    Energy Technology Data Exchange (ETDEWEB)

    Herrmannsfeldt, W.B. [Stanford Linear Accelerator Center, Menlo Park, CA (United States); Bangerter, R.O. [Lawrence Berkeley Lab., CA (United States); Bock, R. [Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany); Hogan, W.J.; Lindl, J.D. [Lawrence Livermore National Lab., CA (United States)

    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.

  13. 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)

  14. 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.

  15. Fusion Propulsion and Power for Future Flight

    Science.gov (United States)

    Froning, H. D., Jr.

    1996-01-01

    There are innovative magnetic and electric confinement fusion power and propulsion system designs with potential for: vacuum specific impulses of 1500-2000 seconds with rocket engine thrust/mass ratios of 5-10 g's; environmentally favorable exhaust emissions if aneutronic fusion propellants can be used; a 2 to 3-fold reduction in the mass of hypersonic airliners and SSTO aerospace planes; a 10 to 20 fold reduction in Mars expedition mass and cost (if propellant from planetary atmospheres is used); and feasibility or in-feasibility of these systems could be confirmed with a modest applied research and exploratory development cost.

  16. Zonal flow generation in inertial confinement fusion implosions

    Science.gov (United States)

    Peterson, J. L.; Humbird, K. D.; Field, J. E.; Brandon, S. T.; Langer, S. H.; Nora, R. C.; Spears, B. K.; Springer, P. T.

    2017-03-01

    A supervised machine learning algorithm trained on a multi-petabyte dataset of inertial confinement fusion simulations has identified a class of implosions that robustly achieve high yield, even in the presence of drive variations and hydrodynamic perturbations. These implosions are purposefully driven with a time-varying asymmetry, such that coherent flow generation during hotspot stagnation forces the capsule to self-organize into an ovoid, a shape that appears to be more resilient to shell perturbations than spherical designs. This new class of implosions, whose configurations are reminiscent of zonal flows in magnetic fusion devices, may offer a path to robust inertial fusion.

  17. Reference commercial fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Young, J.R.; Gore, B.F.

    1976-09-01

    Currently available conceptual designs for commercial fusion power plants are for first generation plants using deuterium-tritium (D-T) fuel, and are all functionally similar. This similarity has been used as a basis for defining an envelope of D-T fusion power plant characteristics which encompasses the characteristics of the available designs. A description of this envelope, including general process descriptions, proposed materials uses and a tabulation of numerical ranges of plant parameters is presented in this document.

  18. Conical targets and pinch confinement for inertial fusion: Opacity calculations of plasmas by using parametric potentials

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, P.M.; Martinez-Val, J.M.; Eliezer, S.; Piera, M.; Chacon, L. [Universidad Politecnica de Madrid (Spain). Inst. de Fusion Nuclear

    1996-12-31

    Conical microducts and minithrottles can be used to accelerate micropellets of fusionable fuel up to very high speeds ({approx}10{sup 8} cm/s). The central collision of two pellets flying in opposite directions can produce a hot plasma where fusion reactions are triggered. The main drawback of this scheme is the short confinement time provided by the external guide tube (throttle). To obtain high yield, an extra force of confinement is advisable. In this paper, the performance of fuel implosions within conical targets and the effect of ultrashort magnetic fields and pinch forces are analyzed. Although very high currents are needed to stretch the confinement time, modern technologies based on pulse-power machines and fast discharges induced by ultrashort lasers can provide a solution to this problem. (Author).

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

    Energy Technology Data Exchange (ETDEWEB)

    Barboza, Nigel Oswald [Univ. of California, Berkeley, CA (United States)

    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/cm3 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.

  20. EDITORIAL: Safety aspects of fusion power plants

    Science.gov (United States)

    Kolbasov, B. N.

    2007-07-01

    importance for the fusion power plant research programmes. The objective of this Technical Meeting was to examine in an integrated way all the safety aspects anticipated to be relevant to the first fusion power plant prototype expected to become operational by the middle of the century, leading to the first generation of economically viable fusion power plants with attractive S&E features. After screening by guest editors and consideration by referees, 13 (out of 28) papers were accepted for publication. They are devoted to the following safety topics: power plant safety; fusion specific operational safety approaches; test blanket modules; accident analysis; tritium safety and inventories; decommissioning and waste. The paper `Main safety issues at the transition from ITER to fusion power plants' by W. Gulden et al (EU) highlights the differences between ITER and future fusion power plants with magnetic confinement (off-site dose acceptance criteria, consequences of accidents inside and outside the design basis, occupational radiation exposure, and waste management, including recycling and/or final disposal in repositories) on the basis of the most recent European fusion power plant conceptual study. Ongoing S&E studies within the US inertial fusion energy (IFE) community are focusing on two design concepts. These are the high average power laser (HAPL) programme for development of a dry-wall, laser-driven IFE power plant, and the Z-pinch IFE programme for the production of an economically-attractive power plant using high-yield Z-pinch-driven targets. The main safety issues related to these programmes are reviewed in the paper `Status of IFE safety and environmental activities in the US' by S. Reyes et al (USA). The authors propose future directions of research in the IFE S&E area. In the paper `Recent accomplishments and future directions in the US Fusion Safety & Environmental Program' D. Petti et al (USA) state that the US fusion programme has long recognized that the S

  1. Fast ignition of asymmetrically compressed targets for inertial confinement fusion

    Science.gov (United States)

    Gus'kov, S. Yu.; Demchenko, N. N.; Zmitrenko, N. V.; Kuchugov, P. A.; Rozanov, V. B.; Stepanov, R. V.; Yakhin, R. A.

    2017-03-01

    It is shown that fast ignition can ensure the combustion of asymmetrically compressed targets for inertial confinement fusion with an efficiency close to the combustion of one-dimensionally compressed targets. This statement is valid not only for targets specially designed for fast ignition. Fast heating by an external energy source can ensure the ignition of a target designed for spark ignition, but where this ignition does not occur because inhomogeneities are formed in the temperature and density distributions owing to the development of hydrodynamic instabilities. The condition for ignition is the fast heating of the plasma in the combustion initiation region whose size is comparable with the sizes of compression inhomogeneities. Thus, fast ignition not only significantly reduces the ignition energy, but also is possibly a necessary stage in the inertial confinement fusion scheme when the spherically symmetric compression of a target requires very high engineering and financial expenses. The studies are based on the numerical simulation of the compression and combustion of inertial confinement fusion targets with one- and two-dimensional hydrodynamic codes.

  2. Bouillabaisse sushi fusion power

    CERN Multimedia

    2004-01-01

    "If avant-garde cuisine is any guide, Japanese-French fusion does not work all that well. And the interminable discussions over the International Thermonuclear Experimental Reactor (ITER) suggest that what is true of cooking is true of physics" (1 page)

  3. 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.

  4. Measuring time of flight of fusion products in an inertial electrostatic confinement fusion device for spatial profiling of fusion reactions.

    Science.gov (United States)

    Donovan, D C; Boris, D R; Kulcinski, G L; Santarius, J F; Piefer, G R

    2013-03-01

    A new diagnostic has been developed that uses the time of flight (TOF) of the products from a nuclear fusion reaction to determine the location where the fusion reaction occurred. The TOF diagnostic uses charged particle detectors on opposing sides of the inertial electrostatic confinement (IEC) device that are coupled to high resolution timing electronics to measure the spatial profile of fusion reactions occurring between the two charged particle detectors. This diagnostic was constructed and tested by the University of Wisconsin-Madison Inertial Electrostatic Confinement Fusion Group in the IEC device, HOMER, which accelerates deuterium ions to fusion relevant energies in a high voltage (∼100 kV), spherically symmetric, electrostatic potential well [J. F. Santarius, G. L. Kulcinski, R. P. Ashley, D. R. Boris, B. B. Cipiti, S. K. Murali, G. R. Piefer, R. F. Radel, T. E. Radel, and A. L. Wehmeyer, Fusion Sci. Technol. 47, 1238 (2005)]. The TOF diagnostic detects the products of D(d,p)T reactions and determines where along a chord through the device the fusion event occurred. The diagnostic is also capable of using charged particle spectroscopy to determine the Doppler shift imparted to the fusion products by the center of mass energy of the fusion reactants. The TOF diagnostic is thus able to collect spatial profiles of the fusion reaction density along a chord through the device, coupled with the center of mass energy of the reactions occurring at each location. This provides levels of diagnostic detail never before achieved on an IEC device.

  5. Innovative Technology for an Inertial Electrostatic Confinement (IEC) Fusion Propulsion Unit

    Science.gov (United States)

    Satsangi, Ann J.; Miley, George H.; Javedani, Jalal B.; Nakashima, Hideki; Yamamoto, Yasushi

    1994-07-01

    An Inertial Electrostatic Confinement (IEC) fusion power source has the potential to support a very efficient nuclear propulsion system. High power-to-weight ratios, necessary for manned travel beyond our moon, are obtainable with an IEC power source because of the grid structure is relatively light weight. A preliminary design study of a system using an IEC power source has been done by integrating an IEC device, designed to burn d-3He, with a Direct Energy Converter (DEC), an electrical system, and an electrical thruster unit. (Miley 1993a) Further investigation of this design will be presented. These subsequent studies were aimed at addressing several key issues which arose during the course of the original design work. The most important result is the discovery that a pulsed power version of the IEC offers a high fusion energy gain, giving an improved specific power for the design.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sweeney, M.A. [ed.

    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.

  7. Diagnosing inertial confinement fusion gamma ray physics (invited).

    Science.gov (United States)

    Herrmann, H W; Hoffman, N; Wilson, D C; Stoeffl, W; Dauffy, L; Kim, Y H; McEvoy, A; Young, C S; Mack, J M; Horsfield, C J; Rubery, M; Miller, E K; Ali, Z A

    2010-10-01

    The gamma reaction history (GRH) diagnostic is a multichannel, time-resolved, energy-thresholded γ-ray spectrometer that provides a high-bandwidth, direct-measurement of fusion reaction history in inertial confinement fusion implosion experiments. 16.75 MeV deuterium+tritium (DT) fusion γ-rays, with a branching ratio of the order of 10(-5)γ/(14 MeV n), are detected to determine fundamental burn parameters, such as nuclear bang time and burn width, critical to achieving ignition at the National Ignition Facility. During the tritium/hydrogen/deuterium ignition tuning campaign, an additional γ-ray line at 19.8 MeV, produced by hydrogen+tritium fusion with a branching ratio of unity, will increase the available γ-ray signal and may allow measurement of reacting fuel composition or ion temperature. Ablator areal density measurements with the GRH are also made possible by detection of 4.43 MeV γ-rays produced by inelastic scatter of DT fusion neutrons on (12)C nuclei in the ablating plastic capsule material.

  8. Diagnosing inertial confinement fusion gamma ray physics (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, H. W.; Hoffman, N.; Wilson, D. C.; Kim, Y. H.; McEvoy, A.; Young, C. S.; Mack, J. M. [Los Alamos National Laboratory, P.O. Box 1663, M/S E526, Los Alamos, New Mexico 87545 (United States); Stoeffl, W.; Dauffy, L. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Horsfield, C. J.; Rubery, M. [Atomic Weapons Establishment, Aldermaston, Reading RG7 4PR (United Kingdom); Miller, E. K. [Special Technologies Laboratory, NSTec, Santa Barbara, California 93111 (United States); Ali, Z. A. [Livermore Operations, NSTec, Livermore, California 94550 (United States)

    2010-10-15

    The gamma reaction history (GRH) diagnostic is a multichannel, time-resolved, energy-thresholded {gamma}-ray spectrometer that provides a high-bandwidth, direct-measurement of fusion reaction history in inertial confinement fusion implosion experiments. 16.75 MeV deuterium+tritium (DT) fusion {gamma}-rays, with a branching ratio of the order of 10{sup -5}{gamma}/(14 MeV n), are detected to determine fundamental burn parameters, such as nuclear bang time and burn width, critical to achieving ignition at the National Ignition Facility. During the tritium/hydrogen/deuterium ignition tuning campaign, an additional {gamma}-ray line at 19.8 MeV, produced by hydrogen+tritium fusion with a branching ratio of unity, will increase the available {gamma}-ray signal and may allow measurement of reacting fuel composition or ion temperature. Ablator areal density measurements with the GRH are also made possible by detection of 4.43 MeV {gamma}-rays produced by inelastic scatter of DT fusion neutrons on {sup 12}C nuclei in the ablating plastic capsule material.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    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.

  10. Characterization of inertial confinement fusion (ICF) targets using PIXE, RBS, and STIM analysis.

    Science.gov (United States)

    Li, Yongqiang; Liu, Xue; Li, Xinyi; Liu, Yiyang; Zheng, Yi; Wang, Min; Shen, Hao

    2013-08-01

    Quality control of the inertial confinement fusion (ICF) target in the laser fusion program is vital to ensure that energy deposition from the lasers results in uniform compression and minimization of Rayleigh-Taylor instabilities. The technique of nuclear microscopy with ion beam analysis is a powerful method to provide characterization of ICF targets. Distribution of elements, depth profile, and density image of ICF targets can be identified by particle-induced X-ray emission, Rutherford backscattering spectrometry, and scanning transmission ion microscopy. We present examples of ICF target characterization by nuclear microscopy at Fudan University in order to demonstrate their potential impact in assessing target fabrication processes.

  11. Energy gain of ignitable targets in inertial confinement fusion (ICF

    Directory of Open Access Journals (Sweden)

    A. Parvazian,J Jafari

    2002-06-01

    Full Text Available   In order to determine the fusion energy gain in a target due to inertial confinement fusion, it is necessary to solve hydrodynamic equations governed on plasma behavior during confinement time. To compress spherical multilayer targets having fuel in the central part, they are irradiated by laser or heavy ion beams. A suitable mass ratio of a pusher is used to ignite the central part of the target. When compression is maximum, fuel density exceeds from 500 to 1000 times of the cold density. Temperature in the cold fuel region rises rapidly and cause the plasma and fusion reaction to take place. Calculations of density, temperature and pressure profiles in the plasma are necessary to obtain the energy flux of neurons, electrons and radiations coming out from the target. Using numerical solutions for continuity, the momentum and energy equations based on a defined continuity equation we prepared a computer program to calculate density, temperature and pressure profiles. The gain of the target as output to input energy is determined. Using this procedure to a designed target with deuterium-tritium (DT fuel derived by heavy ion beams gives an energy gain over 400.

  12. Multibeam Stimulated Raman Scattering in Inertial Confinement Fusion Conditions.

    Science.gov (United States)

    Michel, P; Divol, L; Dewald, E L; Milovich, J L; Hohenberger, M; Jones, O S; Hopkins, L Berzak; Berger, R L; Kruer, W L; Moody, J D

    2015-07-31

    Stimulated Raman scattering from multiple laser beams arranged in a cone sharing a common daughter wave is investigated for inertial confinement fusion (ICF) conditions in a inhomogeneous plasma. It is found that the shared electron plasma wave (EPW) process, where the lasers collectively drive the same EPW, can lead to an absolute instability when the electron density reaches a matching condition dependent on the cone angle of the laser beams. This mechanism could explain recent experimental observations of hot electrons at early times in ICF experiments, at densities well below quarter critical when two plasmon decay is not expected to occur.

  13. Multibeam seeded brillouin sidescatter in inertial confinement fusion experiments.

    Science.gov (United States)

    Turnbull, D; Michel, P; Ralph, J E; Divol, L; Ross, J S; Berzak Hopkins, L F; Kritcher, A L; Hinkel, D E; Moody, J D

    2015-03-27

    We present the first observations of multibeam weakly seeded Brillouin sidescatter in indirect-drive inertial confinement fusion (ICF) experiments. Two seeding mechanisms have been identified and quantified: specular reflections ("glint") from opposite hemisphere beams, and Brillouin backscatter from neighboring beams with a different angle of incidence. Seeded sidescatter can dominate the overall coupling losses, so understanding this process is crucial for proper accounting of energy deposition and drive symmetry. Glint-seeded scattered light could also be used to probe hydrodynamic conditions inside ICF targets.

  14. Open-ended magnetic confinement systems for fusion

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Turbulent Transport at High Reynolds Numbers in an Inertial Confinement Fusion Context

    Science.gov (United States)

    2014-09-01

    ABSTRACT Turbulent Transport at High Reynolds Numbers in an Inertial Confinement Fusion Context Report Title Mix is a critical input to hydro... inertial confinement fusion (ICF) targets. Mix contributes to numerical solution uncertainty through its dependence on turbulent transport coefficients...lanl.gov Turbulent Transport at High Reynolds Numbers in an Inertial Confinement Fusion Context Mix is a critical input to hydro simulations used in

  16. Shock ignition: a new approach to high gain inertial confinement fusion on the national ignition facility.

    Science.gov (United States)

    Perkins, L J; Betti, R; LaFortune, K N; Williams, W H

    2009-07-24

    Shock ignition, an alternative concept for igniting thermonuclear fuel, is explored as a new approach to high gain, inertial confinement fusion targets for the National Ignition Facility (NIF). Results indicate thermonuclear yields of approximately 120-250 MJ may be possible with laser drive energies of 1-1.6 MJ, while gains of approximately 50 may still be achievable at only approximately 0.2 MJ drive energy. The scaling of NIF energy gain with laser energy is found to be G approximately 126E (MJ);{0.510}. This offers the potential for high-gain targets that may lead to smaller, more economic fusion power reactors and a cheaper fusion energy development path.

  17. 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.

  18. Strong coupling and degeneracy effects in inertial confinement fusion implosions.

    Science.gov (United States)

    Hu, S X; Militzer, B; Goncharov, V N; Skupsky, S

    2010-06-11

    Accurate knowledge about the equation of state (EOS) of deuterium is critical to inertial confinement fusion (ICF). Low-adiabat ICF implosions routinely access strongly coupled and degenerate plasma conditions. Using the path integral Monte Carlo method, we have derived a first-principles EOS (FPEOS) table of deuterium. It is the first ab initio EOS table which completely covers typical ICF implosion trajectory in the density and temperature ranges of ρ=0.002-1596  g/cm3 and T=1.35  eV-5.5  keV. Discrepancies in internal energy and pressure have been found in strongly coupled and degenerate regimes with respect to SESAME EOS. Hydrodynamics simulations of cryogenic ICF implosions using the FPEOS table have indicated significant differences in peak density, areal density (ρR), and neutron yield relative to SESAME simulations.

  19. Generalized measurable ignition criterion for inertial confinement fusion.

    Science.gov (United States)

    Chang, P Y; Betti, R; Spears, B K; Anderson, K S; Edwards, J; Fatenejad, M; Lindl, J D; McCrory, R L; Nora, R; Shvarts, D

    2010-04-02

    A multidimensional measurable criterion for central ignition of inertial-confinement-fusion capsules is derived. The criterion accounts for the effects of implosion nonuniformities and depends on three measurable parameters: the neutron-averaged total areal density (rhoR(n)(tot)), the ion temperature (T(n)), and the yield over clean (YOC=ratio of the measured neutron yield to the predicted one-dimensional yield). The YOC measures the implosion uniformity. The criterion can be approximated by chi=(rhoR(n)(tot))(0.8) x (T(n)/4.7)(1.7)YOC(mu)>1 (where rhoR is in g cm(-2), T in keV, and mu approximately 0.4-0.5) and can be used to assess the performance of cryogenic implosions on the NIF and OMEGA. Cryogenic implosions on OMEGA have achieved chi approximately 0.02-0.03.

  20. Progress in direct-drive inertial confinement fusion

    Directory of Open Access Journals (Sweden)

    McCrory R.L.

    2013-11-01

    Full Text Available Significant progress has been made in direct-drive inertial confinement fusion research at the Laboratory for Laser Energetics since the 2009 IFSA Conference [R.L. McCrory et al., J. Phys.: Conf. Ser. 244, 012004 (2010]. Areal densities of 300mg/cm2 have been measured in cryogenic target implosions with neutron yields 15% of 1-D predictions. A model of crossed-beam energy transfer has been developed to explain the observed scattered-light spectrum and laser–target coupling. Experiments show that its impact can be mitigated by changing the ratio of the laser beam to target diameter. Progress continues in the development of the polar-drive concept that will allow direct-drive–ignition experiments to be conducted on the National Ignition Facility using the indirect-drive-beam layout.

  1. Polyvinyl alcohol coating of polystyrene inertial confinement fusion targets

    Science.gov (United States)

    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 microns) on the target (outside diameter of about 350-850 microns). Thicker coatings are obtained by repeated applications of the coating solutions. Preliminary results indicate that uniform coatings may be achievable on the targets with a background surface smoothness in the order of 1000 A.

  2. Development of Compton Radiography Diagnostics for Inertial Confinement Fusion Implosions

    Energy Technology Data Exchange (ETDEWEB)

    Tommasini, R; Hatchett, S P; Hey, D S; Izumi, N; Koch, J A; Landen, O L; Mackinnon, A J; Delettrez, J; Glebov, V; Stoeckl, C

    2010-11-16

    An important diagnostic tool for inertial confinement fusion will be time-resolved radiographic imaging of the dense cold fuel surrounding the hot spot. The measurement technique is based on point-projection radiography at photon energies from 60-200 keV where the Compton effect is the dominant contributor to the opacity of the fuel or pusher. We have successfully applied this novel Compton Radiography technique to the study of the final compression of directly driven plastic capsules at the OMEGA facility. The radiographs have a spatial and temporal resolution of {approx}10 {micro}m and {approx}10ps, respectively. A statistical accuracy of {approx}0.5% in transmission per resolution element is achieved, allowing localized measurements of areal mass densities to 7% accuracy. The experimental results show 3D non-uniformities and lower than 1D expected areal densities attributed to drive asymmetries and hydroinstabilities.

  3. Integrated diagnostic analysis of inertial confinement fusion capsule performancea)

    Science.gov (United States)

    Cerjan, Charles; Springer, Paul T.; Sepke, Scott M.

    2013-05-01

    A conceptual model is developed for typical inertial confinement fusion implosion conditions that integrates available diagnostic information to determine the stagnation properties of the interior fill and surrounding shell. Assuming pressure equilibrium at peak compression and invoking radiative and equation-of-state relations, the pressure, density, and electron temperature are obtained by optimized fitting of the experimental output to smooth, global functional forms. Typical observational data that may be used includes x-ray self-emission, directional neutron time-of-flight signals, neutron yield, high-resolution x-ray spectra, and radiographic images. This approach has been validated by comparison with radiation-hydrodynamic simulations, producing semi-quantitative agreement. Model results implicate poor kinetic energy coupling to the hot core as the primary cause of the observed low thermonuclear burn yields.

  4. 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.

  5. 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.

  6. Demonstration of ignition radiation temperatures in indirect-drive inertial confinement fusion hohlraums.

    Science.gov (United States)

    Glenzer, S H; MacGowan, B J; Meezan, N B; Adams, P A; Alfonso, J B; Alger, E T; Alherz, Z; Alvarez, L F; Alvarez, S S; Amick, P V; Andersson, K S; Andrews, S D; Antonini, G J; Arnold, P A; Atkinson, D P; Auyang, L; Azevedo, S G; Balaoing, B N M; Baltz, J A; Barbosa, F; Bardsley, G W; Barker, D A; Barnes, A I; Baron, A; Beeler, R G; Beeman, B V; Belk, L R; Bell, J C; Bell, P M; Berger, R L; Bergonia, M A; Bernardez, L J; Berzins, L V; Bettenhausen, R C; Bezerides, L; Bhandarkar, S D; Bishop, C L; Bond, E J; Bopp, D R; Borgman, J A; Bower, J R; Bowers, G A; Bowers, M W; Boyle, D T; Bradley, D K; Bragg, J L; Braucht, J; Brinkerhoff, D L; Browning, D F; Brunton, G K; Burkhart, S C; Burns, S R; Burns, K E; Burr, B; Burrows, L M; Butlin, R K; Cahayag, N J; Callahan, D A; Cardinale, P S; Carey, R W; Carlson, J W; Casey, A D; Castro, C; Celeste, J R; Chakicherla, A Y; Chambers, F W; Chan, C; Chandrasekaran, H; Chang, C; Chapman, R F; Charron, K; Chen, Y; Christensen, M J; Churby, A J; Clancy, T J; Cline, B D; Clowdus, L C; Cocherell, D G; Coffield, F E; Cohen, S J; Costa, R L; Cox, J R; Curnow, G M; Dailey, M J; Danforth, P M; Darbee, R; Datte, P S; Davis, J A; Deis, G A; Demaret, R D; Dewald, E L; Di Nicola, P; Di Nicola, J M; Divol, L; Dixit, S; Dobson, D B; Doppner, T; Driscoll, J D; Dugorepec, J; Duncan, J J; Dupuy, P C; Dzenitis, E G; Eckart, M J; Edson, S L; Edwards, G J; Edwards, M J; Edwards, O D; Edwards, P W; Ellefson, J C; Ellerbee, C H; Erbert, G V; Estes, C M; Fabyan, W J; Fallejo, R N; Fedorov, M; Felker, B; Fink, J T; Finney, M D; Finnie, L F; Fischer, M J; Fisher, J M; Fishler, B T; Florio, J W; Forsman, A; Foxworthy, C B; Franks, R M; Frazier, T; Frieder, G; Fung, T; Gawinski, G N; Gibson, C R; Giraldez, E; Glenn, S M; Golick, B P; Gonzales, H; Gonzales, S A; Gonzalez, M J; Griffin, K L; Grippen, J; Gross, S M; Gschweng, P H; Gururangan, G; Gu, K; Haan, S W; Hahn, S R; Haid, B J; Hamblen, J E; Hammel, B A; Hamza, A V; Hardy, D L; Hart, D R; Hartley, R G; Haynam, C A; Heestand, G M; Hermann, M R; Hermes, G L; Hey, D S; Hibbard, R L; Hicks, D G; Hinkel, D E; Hipple, D L; Hitchcock, J D; Hodtwalker, D L; Holder, J P; Hollis, J D; Holtmeier, G M; Huber, S R; Huey, A W; Hulsey, D N; Hunter, S L; Huppler, T R; Hutton, M S; Izumi, N; Jackson, J L; Jackson, M A; Jancaitis, K S; Jedlovec, D R; Johnson, B; Johnson, M C; Johnson, T; Johnston, M P; Jones, O S; Kalantar, D H; Kamperschroer, J H; Kauffman, R L; Keating, G A; Kegelmeyer, L M; Kenitzer, S L; Kimbrough, J R; King, K; Kirkwood, R K; Klingmann, J L; Knittel, K M; Kohut, T R; Koka, K G; Kramer, S W; Krammen, J E; Krauter, K G; Krauter, G W; Krieger, E K; Kroll, J J; La Fortune, K N; Lagin, L J; Lakamsani, V K; Landen, O L; Lane, S W; Langdon, A B; Langer, S H; Lao, N; Larson, D W; Latray, D; Lau, G T; Le Pape, S; Lechleiter, B L; Lee, Y; Lee, T L; Li, J; Liebman, J A; Lindl, J D; Locke, S F; Loey, H K; London, R A; Lopez, F J; Lord, D M; Lowe-Webb, R R; Lown, J G; Ludwigsen, A P; Lum, N W; Lyons, R R; Ma, T; MacKinnon, A J; Magat, M D; Maloy, D T; Malsbury, T N; Markham, G; Marquez, R M; Marsh, A A; Marshall, C D; Marshall, S R; Maslennikov, I L; Mathisen, D G; Mauger, G J; Mauvais, M -Y; McBride, J A; McCarville, T; McCloud, J B; McGrew, A; McHale, B; MacPhee, A G; Meeker, J F; Merill, J S; Mertens, E P; Michel, P A; Miller, M G; Mills, T; Milovich, J L; Miramontes, R; Montesanti, R C; Montoya, M M; Moody, J; Moody, J D; Moreno, K A; Morris, J; Morriston, K M; Nelson, J R; Neto, M; Neumann, J D; Ng, E; Ngo, Q M; Olejniczak, B L; Olson, R E; Orsi, N L; Owens, M W; Padilla, E H; Pannell, T M; Parham, T G; Patterson, R W; Pavel, G; Prasad, R R; Pendlton, D; Penko, F A; Pepmeier, B L; Petersen, D E; Phillips, T W; Pigg, D; Piston, K W; Pletcher, K D; Powell, C L; Radousky, H B; Raimondi, B S; Ralph, J E; Rampke, R L; Reed, R K; Reid, W A; Rekow, V V; Reynolds, J L; Rhodes, J J; Richardson, M J; Rinnert, R J; Riordan, B P; Rivenes, A S; Rivera, A T; Roberts, C J; Robinson, J A; Robinson, R B; Robison, S R; Rodriguez, O R; Rogers, S P; Rosen, M D; Ross, G F; Runkel, M; Runtal, A S; Sacks, R A; Sailors, S F; Salmon, J T; Salmonson, J D; Saunders, R L; Schaffer, J R; Schindler, T M; Schmitt, M J; Schneider, M B; Segraves, K S; Shaw, M J; Sheldrick, M E; Shelton, R T; Shiflett, M K; Shiromizu, S J; Shor, M; Silva, L L; Silva, S A; Skulina, K M; Smauley, D A; Smith, B E; Smith, L K; Solomon, A L; Sommer, S; Soto, J G; Spafford, N I; Speck, D E; Springer, P T; Stadermann, M; Stanley, F; Stone, T G; Stout, E A; Stratton, P L; Strausser, R J; Suter, L J; Sweet, W; Swisher, M F; Tappero, J D; Tassano, J B; Taylor, J S; Tekle, E A; Thai, C; Thomas, C A; Thomas, A; Throop, A L; Tietbohl, G L; Tillman, J M; Town, R P J; Townsend, S L; Tribbey, K L; Trummer, D; Truong, J; Vaher, J; Valadez, M; Van Arsdall, P; Van Prooyen, A J; Vergel de Dios, E O; Vergino, M D; Vernon, S P; Vickers, J L; Villanueva, G T; Vitalich, M A; Vonhof, S A; Wade, F E; Wallace, R J; Warren, C T; Warrick, A L; Watkins, J; Weaver, S; Wegner, P J; Weingart, M A; Wen, J; White, K S; Whitman, P K; Widmann, K; Widmayer, C C; Wilhelmsen, K; Williams, E A; Williams, W H; Willis, L; Wilson, E F; Wilson, B A; Witte, M C; Work, K; Yang, P S; Young, B K; Youngblood, K P; Zacharias, R A; Zaleski, T; Zapata, P G; Zhang, H; Zielinski, J S; Kline, J L; Kyrala, G A; Niemann, C; Kilkenny, J D; Nikroo, A; Van Wonterghem, B M; Atherton, L J; Moses, E I

    2011-02-25

    We demonstrate the hohlraum radiation temperature and symmetry required for ignition-scale inertial confinement fusion capsule implosions. Cryogenic gas-filled hohlraums with 2.2 mm-diameter capsules are heated with unprecedented laser energies of 1.2 MJ delivered by 192 ultraviolet laser beams on the National Ignition Facility. Laser backscatter measurements show that these hohlraums absorb 87% to 91% of the incident laser power resulting in peak radiation temperatures of T(RAD)=300 eV and a symmetric implosion to a 100 μm diameter hot core. © 2011 American Physical Society

  7. Numerical simulation on a new cylindrical target for Z-pinch driven inertial confinement fusion

    Science.gov (United States)

    Chu, Y. Y.; Wang, Z.; Qi, J. M.; Wu, F. Y.; Li, Z. H.

    2017-06-01

    A new indirectly driven cylindrical target is proposed for Z-pinch inertial confinement fusion, and the target implosion dynamics is simulated with a combination of the mass-point model and the radiation hydrodynamic model. Driven by a current waveform with the peak value of 60 MA and 10-90% rising time of 180 ns, the shell kinetic energy of 5 MJ cm-1 can be obtained when the 60 mg cm-1 liner with initial radius 5 cm is imploded to radius of 5 mm. The simulated kinetic energy is loaded to compress the multi-layer cylindrical target, and 24.6 MJ fusion energy can be released according to the radiation hydrodynamic simulation. The power balance relationship is analyzed for the fusion fuel, and the fuel is ignited in the volume-ignition style. The target here can avoid the problem of coupling between the cylindrical Z-pinch and spherical fusion capsule, and can make use of dynamics hohlraum to weaken the influence of Z-pinch instability on the fuel compression. The implosion dynamics of the cylindrical fusion target is easy to diagnose from the axial direction, which makes it suitable to be investigated in future experiments.

  8. Safety review of conceptual fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Clark, R.G.

    1976-11-01

    The potential public safety impacts from accidents in conceptual fusion power plants were investigated. Fusion was found to have some potential for accidents, as does any energy generating system. Functions of fusion power plants were identified that possess sufficient potential for an accidental release of toxic materials to the environment. An assessment was made of the impact of the potential accidents and recommendations are included for R and D that will allow incorporation of safety concerns in fusion power plant design. This work was based on a review of information available in conceptual design documents of fusion reactor systems.

  9. Fuel Target Implosion in Ion beam Inertial Confinement Fusion

    CERN Document Server

    Kawata, Shigeo

    2015-01-01

    The numerical results for the fuel target implosion are presented in order to clarify the target physics in ion beam inertial fusion. The numerical analyses are performed for a direct-driven ion beam target. In the paper the following issues are studied: the beam obliquely incidence on the target surface, the plasma effect on the beam-stopping power, the beam particle energy, the beam time duration, the target radius, the beam input energy and the non-uniformity effect on the fuel target performance. In this paper the beam ions are protons.

  10. Transmission Grating Imaging Spectrometer for Magnetically Confined Fusion Plasmas

    Science.gov (United States)

    Blagojevic, B.; Stutman, D.; Vero, R.; Finkenthal, M.; Moos, H. W.

    2001-10-01

    The Johns Hopkins Plasma Spectroscopy Group is developing a transmission grating (TG) based imaging spectrometer for the soft and ultrasoft X-ray (USXR) ranges. The spectrometer will be integrated into a multi-purpose impurity diagnostic package for Magnetically Confined Fusion experiments, which will provide time and space resolved information about radiation losses, Zeff profiles and particle transport. The package will also include 2-D filtered USXR diode arrays and atomic physics and impurity transport computational capability. The spectrometer has a very simple layout, consisting of two collimating and space resolving slits, a TG and a 2-D imaging detector. As detector we are developing phosphor (P45) coated fiber optic plates with CCD and intensified CCD image readout. The performance of a test 5000 l/mm, 2:1 bar to open area ratio TG has been evaluated in the laboratory using a K-alpha Manson source and the emission from a Penning Discharge. The incident and diffracted photon flux was recorded in the 10-300 Å range with a gas flow proportional counter. The measurements show that spectral resolution and efficiency agree well with the predicted values. A device optimized for spectral resolution and higher order suppression will be tested on the CDX-U and NSTX tokamak at Princeton Plasma Physics Laboratory. Work supported by DoE grant No. DE-FG02-86ER52314ATDoE

  11. 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.

  12. An accelerator based fusion-product source for development of inertial confinement fusion nuclear diagnostics.

    Science.gov (United States)

    McDuffee, S C; Frenje, J A; Séguin, F H; Leiter, R; Canavan, M J; Casey, D T; Rygg, J R; Li, C K; Petrasso, R D

    2008-04-01

    A fusion-product source, utilizing a 150 kV Cockraft-Walton linear accelerator, has been refurbished to provide a reliable nuclear diagnostic development tool to the national inertial confinement fusion (ICF) research program. The accelerator is capable of routinely generating DD reaction rates at approximately 10(7)/s when using a 150 kV, 150 microA deuterium (D) beam onto an erbium (Er) or titanium (Ti) target doped with D, and D(3)He reaction rates at approximately 5 x 10(5)/s when using a using a 120 kV, approximately 100 microA D beam onto a Er or Ti target doped with (3)He. The new accelerator is currently being used in a number of projects related to the national ICF program at the OMEGA Laser Fusion Facility [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)], which includes the wedge range filter charged-particle spectrometry program [F. H. Seguin et al., Rev. Sci Instrum. 75, 3520 (2004)] and the magnetic recoil neutron spectrometer [J. A. Frenje et al., Rev. Sci. Instrum. 72, 854 (2001)].

  13. 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.

  14. An Indispensable Truth How Fusion Power Can Save the Planet

    CERN Document Server

    Chen, Francis F

    2011-01-01

    Both global warming and oil shortage can be solved by controlled fusion, a clean power source that will serve mankind for millennia.� The idea of hydrogen fusion as well as its difficulties are presented in non-technical language to dispel the notion that fusion is always 50 years away.� This book also summarizes the evidence for climate change and explains the principles of both fossil and "green" energy sources to show that fusion is the best alternative for central-station power in the near term as well as the far future. Praise for An Indispensable Truth: How Fusion Power Can Save the Planet: "In this study Professor Chen outlines the underlying physics, recent progress in achieving advanced plasmas and magnetic confinement, and hopes for the future. He recognizes the difficulties that remain in engineering a fusion reactor, but he remains optimistic regarding ultimate success, yet fearful of the consequences were we to fail."- James R. Schlesinger, former Chairman, Atomic Energy Commission; Director,...

  15. 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.

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    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.

  18. 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.

  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. Motivation and fabrication methods for inertial confinement fusion and inertial fusion energy targets

    Science.gov (United States)

    Borisenko, N. G.; Akunets, A. A.; Bushuev, V. S.; Dorogotovtsev, V. M.; Merkuliev, Yu. A.

    2003-10-01

    Popular target designs are reviewed. Possible methods of fusion target fabrication are discussed and the equipment and samples are demonstrated. The properties of the uniform and structured (cluster) materials are considered, showing the advantage of cluster material for energy conversion into soft X rays. The target materials with high content of hydrogen isotopes (BeD2, LiBeD3, or ND3BD3) prove to be more effective for high-power drivers in comparison with beryllium or polyimide.

  1. 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.

  2. Inertial confinement fusion for energy: overview of the ongoing experimental, theoretical and numerical studies

    Science.gov (United States)

    Jacquemot, S.

    2017-10-01

    This paper provides an overview of the results presented at the 26th IAEA Fusion Energy Conference in the field of inertial confinement fusion for energy, covering its various experimental, numerical/theoretical and technological facets, as well as the different paths towards ignition that are currently followed worldwide.

  3. Fatigue cracking of a bare steel first wall in an inertial confinement fusion chamber

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, R. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Abbott, R. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Havstad, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunne, A. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-06-01

    Inertial confinement fusion power plants will deposit high energy X-rays onto the outer surfaces of the first wall many times a second for the lifetime of the plant. These X-rays create brief temperature spikes in the first few microns of the wall, which cause an associated highly compressive stress response on the surface of the material. The periodicity of this stress pulse is a concern due to the possibility of fatigue cracking of the wall. We have used finite element analyses to simulate the conditions present on the first wall in order to evaluate the driving force of crack propagation on fusion-facing surface cracks. Analysis results indicate that the X-ray induced plastic compressive stress creates a region of residual tension on the surface between pulses. This tension film will likely result in surface cracking upon repeated cycling. Additionally, the compressive pulse may induce plasticity ahead of the crack tip, leaving residual tension in its wake. However, the stress amplitude decreases dramatically for depths greater than 80–100 μm into the fusion-facing surface. Crack propagation models as well as stress-life estimates agree that even though small cracks may form on the surface of the wall, they are unlikely to propagate further than 100 μm without assistance from creep or grain erosion phenomena.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

    2005-07-01

    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

  5. 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)

  6. High-gain inertial confinement fusion by volume ignition, avoiding the complexities of fusion detonation fronts of spark ignition

    Energy Technology Data Exchange (ETDEWEB)

    Hora, H. [Univ. of New South Wales, Sydney, New South Wales (Australia). Dept. of Theoretical Physics; Eliezer, S. [Soreq Nuclear Research Center, Yavne (Israel); Honrubia, J.J.; Martinez-Val, J.M.; Valarde, G. [Univ. Politecnica de Madrid (Spain). Inst. de Fusion Nuclear; Miley, G.H. [Univ. of Illinois, Urbana, IL (United States). Fusion Studies Lab.; Hoepfl, R.

    1995-12-31

    The main approach to Inertial Confinement Fusion (ICF) uses a high-temperature, low-density core and a high-density, low-temperature outer region of the laser- (or ion beam-)compressed deuterium-tritium (D-T) fuel, in order to ignite a fusion detonation wave at the interface. This is an extremely delicate, unstable configuration which is very difficult to achieve, even with a carefully programmed time dependence of the deposition of the driver energy. This approach was devised in order to reach the high gains needed for low-efficiency lasers. Since 1978, several teams have developed an alternative scheme using volume ignition, where a natural and simple adiabatic compression, starting from a low initial temperature of 3 keV or less, is used. The high gains are obtained by self-heating due to the fusion reaction products plus self-absorption of Bremsstrahlung. Fortunately, a strong deviation from LTE occurs at ion temperatures above 100 keV, with much lower electron and even lower radiation temperatures. The authors report here how the gains calculated by different groups are relatively large, and despite detailed differences in the stopping power models, do not differ greatly. The high gain can be explained by introducing an effective value for the density-radius ({rho}R) product, where the volume ignition process increases the usual value of about 3 g-cm{sup {minus}2} to an effective value of 12 g-cm{sup {minus}2} or more, due to the self-generated additional heating that occurs for beam input energies > MJ and compression over 1,000 times solid state. This result is valid for direct drive as well as for indirect drive.

  7. 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.

  8. Thin Shell, High Velocity Inertial Confinement Fusion Implosions on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ma, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hurricane, O. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Callahan, D. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Barrios, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Casey, D. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dewald, E. L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dittrich, T. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Doppner, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Haan, S. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hinkel, D. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Berzak Hopkins, L. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Le Pape, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); MacPhee, A. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Pak, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Park, H. S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Patel, P. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Remington, B. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robey, H. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Salmonson, J. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Springer, P. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tommasini, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Benedetti, L. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bionta, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bond, E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bradley, D. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Caggiano, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Celliers, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Cerjan, C. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Church, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dixit, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dylla-Spears, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Edgell, D. [Univ. of Rochester, NY (United States); Edwards, M. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Field, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fittinghoff, D. N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frenje, J. A. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Plasma Science and Fusion Center; Gatu Johnson, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Plasma Science and Fusion Center; Grim, G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Guler, N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hatarik, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Herrmann, H. W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hsing, W. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Izumi, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jones, O. S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Khan, S. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kilkenny, J. D. [General Atomics, San Diego, CA (United States); Knauer, J. [Univ. of Rochester, NY (United States); Kohut, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kozioziemski, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kritcher, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kyrala, G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Landen, O. L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); MacGowan, B. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mackinnon, A. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meezan, N. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Merrill, F. E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moody, J. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nagel, S. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nikroo, A. [General Atomics, San Diego, CA (United States); Parham, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ralph, J. E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rosen, M. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rygg, J. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sater, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sayre, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schneider, M. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shaughnessy, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Spears, B. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Town, R.P. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Volegov, P. L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wan, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Widmann, K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wilde, C. H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yeamans, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-04-06

    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. Earlier results have shown good repeatability, with up to 1/2 the neutron yield coming from α-particle self-heating.

  9. Controlled Nuclear Fusion by Magnetic Confinement and ITER

    CERN Multimedia

    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.

  10. Novel Fusion Reactor For Space Power And Propulsion

    Science.gov (United States)

    Kammash, T.; Galbraith, D. L.

    1988-04-01

    Space exploration and other exotic applications of space have recently drawn attention to the need for multimegawatt power producing systems and other systems that could deliver gigawatts of power for short periods of time. Several nuclear fission schemes have been proposed but power to weight ratios and other considerations may open the door to other innovative approaches that may develop in the not too distant future. One such scheme is the "Magnetically Insulated Inertial Confinement Fusion" (MICF) reactor which combines the favorable aspects of both inertial and magnetic fusions in that physical containment of the burning plasma is provided by a metallic shell while thermal insulation is provided by a strong, self-generated magnetic field. Because of these unique properties the lifetime of the plasma is sufficiently longer than conventional, implosion type inertial fusion that very attractive energy gains can be achieved. In this paper we utilize a quasi one-dimensional, time-dependent set of appropriate equations to investigate the dynamic and reactor properties of this system and apply the results to a space-based power reactor, and to an advanced space propulsion device. In both instances we find that MICF can meet the space needs of the next century.

  11. ORNL fusion power demonstration study: interim report

    Energy Technology Data Exchange (ETDEWEB)

    STeiner, D.; Bettis, E. S.; Huxford, T. J.

    1977-03-01

    The purpose of the ORNL Fusion Power Demonstration Study (Demo study) is to develop a plan for demonstrating, in this century, the commercial feasibility of fusion power based on the tokamak concept. The two-year study was initiated in FY 1976, and this interim report summarizes the results for FY 1976. Major results include: (1) the outline of a three-phase plan for demonstrating the commercial feasibility of tokamak fusion power in this century; (2) a parametric analysis of tokamak costs which provides the economic basis for the demonstration plan; and (3) a critical evaluation of the technological directions, design approaches, and plasma characteristics which serve as the technical basis for the demonstration plan.

  12. 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.

  13. Inertial Confinement Fusion: Quarterly report, April-June 1996

    Energy Technology Data Exchange (ETDEWEB)

    Correll, D.

    1996-06-01

    The lead article, `Ion-beam propagation in a low-density reactor chamber for heavy-ion inertial fusion` (p. 89), explores the ability of heavy-ion beams to be adequately transported and focused in an IFE reactor. The next article, `Efficient production and applications of 2- to 10-keV x rays by laser-heated underdense radiators` (p. 96), explores the ability of the NIF to produce sufficient high-energy x rays for diagnostic backlighting, target preheating, or uniform irradiation of large test objects for Nuclear Weapons Effects Testing. For capsule implosion experiments, the increasing energies and distances involved in the NIF compared to Nova require the development of new diagnostics methods. The article `Fusion reaction-rate measurements--Nova and NIF` (p. 115) first reviews the use of time-resolved neutron measurements on Nova to monitor fusion burn histories and then explores the limitations of that technique, principally Doppler broadening, for the proposed NIF. It also explores the use of gamma rays on Nova, thereby providing a proof-of-principle for using gamma rays for monitoring fusion burn histories on the NIF. The articles `The energetics of gas-filled hohlraums` (p. 110) and `Measurements of laser- speckle-induced perturbations in laser-driven foils` (p. 123) report measurements on Nova of two important aspects of implosion experiments. The first characterizes the amount of energy lost from a hohlraum by stimulated Brillouin and Raman scattering as a function of gas fill and laser-beam uniformity. The second of these articles shows that the growth of density nonuniformities implanted on smooth capsule surfaces by laser speckle can be correlated with the effects of physical surface roughness. The article `Laser-tissue interaction modeling with the LATIS computer program` (p. 103) explores the use of modeling to enhance the effectiveness--maximize desired effects and minimize collateral damage--of lasers for medical purposes.

  14. Deflagration-to-detonation transition in inertial-confinement-fusion baseline targets

    Science.gov (United States)

    Gauthier, P.; Chaland, F.; Masse, L.

    2004-11-01

    By means of highly resolved one-dimensional hydrodynamics simulations, we provide an understanding of the burn process in inertial-confinement-fusion baseline targets. The cornerstone of the phenomenology of propagating burn in such laser-driven capsules is shown to be the transition from a slow unsteady reaction-diffusion regime of thermonuclear combustion (some sort of deflagration) to a fast detonative one. Remarkably, detonation initiation follows the slowing down of a shockless supersonic reaction wave driven by energy redeposition from the fusion products themselves. Such a route to detonation is specific to fusion plasmas.

  15. GAMMA-RAY DIAGNOSTICS OF ALPHA-SLOWING IN INERTIAL CONFINEMENT FUSION-TARGETS

    NARCIS (Netherlands)

    DENDOOVEN, PG; DRAKE, RP; CABLE, MD; Dendooven, Peter

    1993-01-01

    For large inertial confinement fusion deuterium-tritium targets, a way to diagnose alpha slowing might be via capture reaction gamma rays. Calculations are presented for two such methods: one uses the alpha+T direct capture gamma rays, the other is based on a series of resonant alpha-capture

  16. Inertial confinement fusion quarterly report, July--September 1994. Volume 4, Number 4

    Energy Technology Data Exchange (ETDEWEB)

    Honea, E. [ed.

    1994-09-01

    The ICF Quarterly continues with six articles in this issue describing recent developments in the Inertial Confinement Fusion (ICF) Program at Lawrence Livermore National Laboratory. The topics include plasma characterization, production of millimeter scale-length plasmas for studying laser-plasma instabilities, hohlraum physics, three-dimensional hydrodynamic modeling, crystal growth, and laser-beam smoothing.

  17. Engineering design of the Nova Laser Facility for inertial-confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, W W; Godwin, R O; Hurley, C A; Wallerstein, E. P.; Whitham, K.; Murray, J. E.; Bliss, E. S.; Ozarski, R. G.; Summers, M. A.; Rienecker, F.; Gritton, D. G.; Holloway, F. W.; Suski, G. J.; Severyn, J. R.

    1982-01-25

    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.

  18. 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.

  19. An Inertial-Fusion Z-Pinch Power Plant Concept

    Energy Technology Data Exchange (ETDEWEB)

    DERZON,MARK S.; ROCHAU,GARY E.; DEGROOT,J.; OLSON,CRAIG L.; PETERSON,P.; PETERSON,R.R.; SLUTZ,STEPHEN A.; ZAMORA,ANTONIO J.

    2000-12-15

    With the promising new results of fast z-pinch technology developed at Sandia National Laboratories, we are investigating using z-pinch driven high-yield Inertial Confinement Fusion (ICF) as a fusion power plant energy source. These investigations have led to a novel fusion system concept based on an attempt to separate many of the difficult fusion engineering issues and a strict reliance on existing technology, or a reasonable extrapolation of existing technology, wherever possible. In this paper, we describe the main components of such a system with a focus on the fusion chamber dynamics. The concept works with all of the electrically-coupled ICF proposed fusion designs. It is proposed that a z-pinch driven ICF power system can be feasibly operated at high yields (1 to 30 GJ) with a relatively low pulse rate (0.01-0.1 Hz). To deliver the required current from the rep-rated pulse power driver to the z-pinch diode, a Recyclable Transmission Line (RTL) and the integrated target hardware are fabricated, vacuum pumped, and aligned prior to loading for each power pulse. In this z-pinch driven system, no laser or ion beams propagate in the chamber such that the portion of the chamber outside the RTL does not need to be under vacuum. Additionally, by utilizing a graded-density solid lithium or fluorine/lithium/beryllium eutectic (FLiBe) blanket between the source and the first-wall the system can breed its own fuel absorb a large majority of the fusion energy released from each capsule and shield the first-wall from a damaging neutron flux. This neutron shielding significantly reduces the neutron energy fluence at the first-wall such that radiation damage should be minimal and will not limit the first-wall lifetime. Assuming a 4 m radius, 8 m tall cylindrical chamber design with an 80 cm thick spherical FLiBe blanket, our calculations suggest that a 20 cm thick 6061-T6 Al chamber wall will reach the equivalent uranium ore radioactivity level within 100 years after a 30

  20. 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.

  1. 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

  2. Using gamma-ray emission to measure areal density of inertial confinement fusion capsules.

    Science.gov (United States)

    Hoffman, N M; Wilson, D C; Herrmann, H W; Young, C S

    2010-10-01

    Fusion neutrons streaming from a burning inertial confinement fusion capsule generate gamma rays via inelastic nuclear scattering in the ablator of the capsule. The intensity of gamma-ray emission is proportional to the product of the ablator areal density (ρR) and the yield of fusion neutrons, so by detecting the gamma rays we can infer the ablator areal density, provided we also have a measurement of the capsule's total neutron yield. In plastic-shell capsules, for example, (12)C nuclei emit gamma rays at 4.44 MeV after excitation by 14.1 MeV neutrons from D+T fusion. These gamma rays can be measured by a new gamma-ray detector under development. Analysis of predicted signals is in progress, with results to date indicating that the method promises to be useful for diagnosing imploded capsules.

  3. 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.

  4. Pulsed power accelerators for particle beam fusion

    Science.gov (United States)

    Martin, T. H.; Barr, G. W.; Vandevender, J. P.; White, R. A.; Johnson, D. L.

    1980-05-01

    Sandia National Laboratories is completing the construction phase of the Particle Beam Fusion Accelerator-1 (PBFA-1). Testing of the 36 module, 30 TW, 1 MJ output accelerator is in the initial stages. The 4 MJ, PBFA Marx generator provided 3.6 MA into water-copper sulfate load resistors with a spread from first to last Marx firing between 15 to 25 ns and an output power of 5.7 TW. This accelerator is a modular, lower voltage, pulsed power device that is capable of scaling to power levels exceeding 100 TW. The elements of the PBFA technology and their integration into an accelerator system for particle beam fusion is discussed.

  5. Intense ion beam applications to magnetic confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Sudan, R N

    1980-08-18

    The ion ring project objective is to trap a ring of high energy, axis-encircling ions in a magnetic mirror. The number of ring ions should be such as to produce deltaB/B on the ring axis of order 10%. The second experiment, LONGSHOT, is directed to producing a long pulse ion beam source so that the total number of protons required for an ion ring can be provided a lower diode power and, hence, at much less cost than that of 100 nsec pulsed power generators like the NRL GAMBLE II. A detailed report of the progress on IREX and LONGSHOT is given. (MOW)

  6. A new ignition hohlraum design for indirect-drive inertial confinement fusion

    CERN Document Server

    Xin, Li; Zhensheng, Dai; Wudi, Zheng; Jianfa, Gu; Peijun, Gu; Shiyang, Zou; Jie, Liu; Shaoping, Zhu

    2016-01-01

    In this paper, a six-cylinder-port hohlraum is proposed to provide high symmetry flux on capsule. It is designed to ignite a capsule with 1.2 mm radius in indirect-drive inertial confinement fusion (ICF) . Flux symmetry and laser energy are calculated by using three dimensional view factor method and laser energy balance in hohlraums. Plasma conditions are analyzed based on the two dimensional radiation-hydrodynamic simulations. There is no Ylm (l<=4) asymmetry in the six-cylinder-port hohlraum when the influences of laser entrance holes (LEHs) and laser spots cancel each other out with suitable target parameters. A radiation drive with 300 eV and good flux symmetry can be achieved with use of laser energy of 2.3 MJ and 500 TW peak power. According to the simulations, the electron temperature and the electron density on the wall of laser cone are high and low, respectively, which are similar to those of outer cones in the hohlraums on National Ignition Facility (NIF). And the laser intensity is also as low...

  7. 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.

  8. 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.

  9. Self-Similar Structure and Experimental Signatures of Suprathermal Ion Distribution in Inertial Confinement Fusion Implosions.

    Science.gov (United States)

    Kagan, Grigory; Svyatskiy, D; Rinderknecht, H G; Rosenberg, M J; Zylstra, A B; Huang, C-K; McDevitt, C J

    2015-09-04

    The distribution function of suprathermal ions is found to be self-similar under conditions relevant to inertial confinement fusion hot spots. By utilizing this feature, interference between the hydrodynamic instabilities and kinetic effects is for the first time assessed quantitatively to find that the instabilities substantially aggravate the fusion reactivity reduction. The ion tail depletion is also shown to lower the experimentally inferred ion temperature, a novel kinetic effect that may explain the discrepancy between the exploding pusher experiments and rad-hydro simulations and contribute to the observation that temperature inferred from DD reaction products is lower than from DT at the National Ignition Facility.

  10. Fusion in a magnetically-shielded-grid inertial electrostatic confinement device

    CERN Document Server

    Hedditch, John; Khachan, Joe

    2015-01-01

    Theory for a gridded inertial electrostatic confinement (IEC) fusion system is presented that 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.

  11. Self-similar structure and experimental signatures of suprathermal ion distribution in inertial confinement fusion implosions

    CERN Document Server

    Kagan, Grigory; Rinderknecht, H G; Rosenberg, M J; Zylstra, A B; Huang, C -K

    2015-01-01

    The distribution function of suprathermal ions is found to be self-similar under conditions relevant to inertial confinement fusion hot-spots. By utilizing this feature, interference between the hydro-instabilities and kinetic effects is for the first time assessed quantitatively to find that the instabilities substantially aggravate the fusion reactivity reduction. The ion tail depletion is also shown to lower the experimentally inferred ion temperature, a novel kinetic effect that may explain the discrepancy between the exploding pusher experiments and rad-hydro simulations and contribute to the observation that temperature inferred from DD reaction products is lower than from DT at National Ignition Facility.

  12. Aperture tolerances for neutron-imaging systems in inertial confinement fusion.

    Science.gov (United States)

    Ghilea, M C; Sangster, T C; Meyerhofer, D D; Lerche, R A; Disdier, L

    2008-02-01

    Neutron-imaging systems are being considered as an ignition diagnostic for the National Ignition Facility (NIF) [Hogan et al., Nucl. Fusion 41, 567 (2001)]. Given the importance of these systems, a neutron-imaging design tool is being used to quantify the effects of aperture fabrication and alignment tolerances on reconstructed neutron images for inertial confinement fusion. The simulations indicate that alignment tolerances of more than 1 mrad would introduce measurable features in a reconstructed image for both pinholes and penumbral aperture systems. These simulations further show that penumbral apertures are several times less sensitive to fabrication errors than pinhole apertures.

  13. 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.

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

    Directory of Open Access Journals (Sweden)

    Yanez C.

    2013-11-01

    Full Text Available 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.

  15. Fusion Power Program. Quarterly progress report, January-March 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    This quarterly report summarizes the Argonne National Laboratory work performed for the Office of Fusion Energy during the January-March 1979 quarter in the following research and development areas: materials; energy storage and transfer; tritium containment, recovery and control; advanced reactor design; atomic data; reactor safety; fusion-fission hybrid systems; alternate applications of fusion energy; and other work related to fusion power.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. 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.

  18. 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

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

    Science.gov (United States)

    Shaughnessy, D A; Moody, K J; Gharibyan, N; Grant, P M; Gostic, J M; Torretto, P C; Wooddy, P T; Bandong, B B; Despotopulos, J D; 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

    2014-06-01

    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 (198m+g)Au and (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.

  20. 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.

  1. Down-scattered neutron imaging detector for areal density measurement of inertial confinement fusion.

    Science.gov (United States)

    Arikawa, Y; Yamanoi, K; Nakazato, T; Estacio, E S; Shimizu, T; Sarukura, N; Nakai, M; Hosoda, H; Norimatsu, T; Hironaka, Y; Azechi, H; Izumi, N; Murata, T; Fujino, S; Yoshida, H; Kamada, K; Usuki, Y; Suyama, T; Yoshikawa, A; Satoh, N; Kan, H

    2010-10-01

    A custom developed (6)Li glass scintillator (APLF80+3Pr) for down-scattered neutron diagnostics in inertial confinement fusion experiments is presented. (6)Li provides an enhanced sensitivity for down-scattered neutrons in DD fusion and its experimentally observed 5-6 ns response time fulfills the requirement for down-scattered neutron detectors. A time-of-flight detector operating in the current mode using the APLF80+3Pr was designed and its feasibility observing down-scattered neutrons was demonstrated. Furthermore, a prototype design for a down-scattered neutron imaging detector was also demonstrated. This material promises viability as a future down-scattered neutron detector for the National Ignition Facility.

  2. Optimization of laser illumination configuration for directly driven inertial confinement fusion

    Directory of Open Access Journals (Sweden)

    Masakatsu Murakami

    2017-03-01

    Full Text Available Optimum laser configurations are presented to achieve high illumination uniformity with directly driven inertial confinement fusion targets. Assuming axisymmetric absorption pattern of individual laser beams, theoretical models are reviewed in terms of the number of laser beams, system imperfection, and laser beam patterns. Utilizing a self-organizing system of charged particles on a sphere, a simple numerical model is provided to give an optimal configuration for an arbitrary number of laser beams. As a result, such new configurations as “M48” and “M60” are found to show substantially higher illumination uniformity than any other existing direct drive systems. A new polar direct-drive scheme is proposed with the laser axes keeping off the target center, which can be applied to laser configurations designed for indirectly driven inertial fusion.

  3. Down-scattered neutron imaging detector for areal density measurement of inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Arikawa, Y.; Yamanoi, K.; Nakazato, T.; Estacio, E. S.; Shimizu, T.; Sarukura, N.; Nakai, M.; Hosoda, H.; Norimatsu, T.; Hironaka, Y.; Azechi, H. [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Izumi, N. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); Murata, T. [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kumamoto University, 2-40-1 Kurokami, Kumamoto 860-8555 (Japan); Fujino, S. [Kyushu University, 744 Nishiku, Motooka, Fukuoka 819-0395 (Japan); Yoshida, H. [Ceramic Research Center of Nagasaki, Hiekoba, Hisami, Higashisonogi 859-3726 (Japan); Kamada, K.; Usuki, Y. [Furukawa Co. Ltd., 1-25-13 Kannondai, Tsukuba, Ibaraki 305-0856 (Japan); Suyama, T. [Tokuyama Co. Ltd., 3-3-1 Shibuyaku, Shibuya, Tokyo 150-8383 (Japan); Yoshikawa, A. [Tohoku University, 2-1-1 Katahira, Aoyou, Sendai, Miyagi 980-8577 (Japan); Satoh, N. [Hamamatsu Photonics K.K., 5000 Hiraguchi, Hamakitaku, Hamamatsu, Shizuoka 434-8601 (Japan); and others

    2010-10-15

    A custom developed {sup 6}Li glass scintillator (APLF80+3Pr) for down-scattered neutron diagnostics in inertial confinement fusion experiments is presented. {sup 6}Li provides an enhanced sensitivity for down-scattered neutrons in DD fusion and its experimentally observed 5-6 ns response time fulfills the requirement for down-scattered neutron detectors. A time-of-flight detector operating in the current mode using the APLF80+3Pr was designed and its feasibility observing down-scattered neutrons was demonstrated. Furthermore, a prototype design for a down-scattered neutron imaging detector was also demonstrated. This material promises viability as a future down-scattered neutron detector for the National Ignition Facility.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. 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.

  6. Magnetic field generation in Rayleigh-Taylor unstable inertial confinement fusion plasmas.

    Science.gov (United States)

    Srinivasan, Bhuvana; Dimonte, Guy; Tang, Xian-Zhu

    2012-04-20

    Rayleigh-Taylor instabilities (RTI) in inertial confinement fusion implosions are expected to generate magnetic fields. A Hall-MHD model is used to study the field generation by 2D single-mode and multimode RTI in a stratified two-fluid plasma. Self-generated magnetic fields are predicted and these fields grow as the RTI progresses via the ∇n(e)×∇T(e) term in the generalized Ohm's law. Scaling studies are performed to determine the growth of the self-generated magnetic field as a function of density, acceleration, Atwood number, and perturbation wavelength.

  7. Time-dependent nuclear measurements of mix in inertial confinement fusion.

    Science.gov (United States)

    Rygg, J R; Frenje, J A; Li, C K; Séguin, F H; Petrasso, R D; Glebov, V Yu; Meyerhofer, D D; Sangster, T C; Stoeckl, C

    2007-05-25

    The first time-dependent nuclear measurements of turbulent mix in inertial confinement fusion have been obtained. Implosions of spherical deuterated-plastic shells filled with pure 3He gas require atomic-scale mixing of the shell and gas for the D-3He nuclear reaction to proceed. The time necessary for Rayleigh-Taylor (RT) growth to induce mix delays peak nuclear production time, compared to equivalent capsules filled with a D2-3He mixture, by 75+/-30 ps, equal to half the nuclear burn duration. These observations indicate the likelihood of atomic mix at the tips of core-penetrating RT spikes.

  8. Prolate-spheroid ("rugby-shaped") hohlraum for inertial confinement fusion.

    Science.gov (United States)

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

    2007-08-10

    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.

  9. Effects of ionization gradients on inertial-confinement-fusion capsule hydrodynamic stability.

    Science.gov (United States)

    Amendt, Peter

    2008-09-12

    A linear perturbation analysis based on velocity potentials is adapted to include the regional, average-ion charge states (Z) in an imploding, inertial-confinement-fusion capsule and shown to lead to superclassical Rayleigh-Taylor growth following deceleration onset. The added instability is ascribed to an inverted ion-entropy gradient driven by the stepwise ionization mismatch DeltaZ across the fuel-pusher interface and is predicted to principally occur in low Atwood-number (<0.5) implosions associated with low-Z pushers. Similar instability enhancement may pertain to supernovae phenomena and ionization fronts in H II protostellar regions.

  10. Modeling Stimulated Raman Scattering in Direct-Drive Inertial Confinement Fusion Plasmas for National Ignition Facility Conditions

    Science.gov (United States)

    Maximov, A. V.; Shaw, J. G.; Myatt, J. F.; Short, R. W.

    2017-10-01

    In the plasmas of direct-drive inertial confinement fusion (ICF), the coupling of laser power to the target plasma is strongly influenced by the laser-plasma interaction (LPI) processes driven by multiple crossing laser beams. For the plasma parameters relevant to the conditions of the experiments at the National Ignition Facility (NIF), the threshold of the stimulated Raman scattering (SRS) is usually well exceeded because of the large scale length of the plasma density, making the study of SRS vital for the NIF ICF program. The SRS evolution starts as a convective or absolute instability, and the nonlinear saturation is determined by the ion-acoustic perturbations and kinetic effects. The LPI processes of cross-beam energy transfer and two-plasmon decay also drive the ion-acoustic modes and their interplay with SRS is analyzed. This work was supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  11. Fusion

    CERN Document Server

    Mahaffey, James A

    2012-01-01

    As energy problems of the world grow, work toward fusion power continues at a greater pace than ever before. The topic of fusion is one that is often met with the most recognition and interest in the nuclear power arena. Written in clear and jargon-free prose, Fusion explores the big bang of creation to the blackout death of worn-out stars. A brief history of fusion research, beginning with the first tentative theories in the early 20th century, is also discussed, as well as the race for fusion power. This brand-new, full-color resource examines the various programs currently being funded or p

  12. Fusion Power Program. Quarterly progress report, October--December 1978

    Energy Technology Data Exchange (ETDEWEB)

    1979-04-01

    This quarterly report summarizes the Argonne National Laboratory work performed for the Office of Fusion Energy during the October--December 1978 quarter in the following research and development areas: materials; energy storage and transfer; tritium containment, recovery and control; advanced reactor design; atomic data; reactor safety; fusion-fission hybrid systems; alternate applications of fusion energy; and other work related to fusion power. Three separate abstracts were prepared for the included sections. (MOW)

  13. 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.

  14. 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.

  15. On the importance of minimizing "coast-time" in x-ray driven inertially confined fusion implosions

    Science.gov (United States)

    Hurricane, O. A.; Kritcher, A.; Callahan, D. A.; Landen, O.; Patel, P. K.; Springer, P. T.; Casey, D. T.; Dewald, E. L.; Dittrich, T. R.; Döppner, T.; Hinkel, D. E.; Berzak Hopkins, L. F.; Kline, J.; Le Pape, S.; Ma, T.; MacPhee, A. G.; Moore, A.; Pak, A.; Park, H.-S.; Ralph, J.; Salmonson, J. D.; Widmann, K.

    2017-09-01

    By the time an inertially confined fusion (ICF) implosion has converged a factor of 20, its surface area has shrunk 400 × , making it an inefficient x-ray energy absorber. So, ICF implosions are traditionally designed to have the laser drive shut off at a time, toff, well before bang-time, tBT, for a coast-time of t coast = t B T - t o f f > 1 ns. High-foot implosions on NIF showed a strong dependence of many key ICF performance quantities on reduced coast-time (by extending the duration of laser power after the peak power is first reached), most notably stagnation pressure and fusion yield. Herein we show that the ablation pressure, pabl, which drives high-foot implosions, is essentially triangular in temporal shape, and that reducing tcoast boosts pabl by as much as ˜ 2 × prior to stagnation thus increasing fuel and hot-spot compression and implosion speed. One-dimensional simulations are used to track hydrodynamic characteristics for implosions with various coast-times and various assumed rates of hohlraum cooling after toff to illustrate how the late-time conditions exterior to the implosion can impact the fusion performance. A simple rocket model-like analytic theory demonstrates that reducing coast-time can lead to a ˜ 15 % higher implosion velocity because the reduction in x-ray absorption efficiency at late-time is somewhat compensated by small ( ˜ 5 % - 10 %) ablator mass remaining. Together with the increased ablation pressure, the additional implosion speed for short coast-time implosions can boost the stagnation pressure by ˜ 2 × as compared to a longer coast-time version of the same implosion. Four key dimensionless parameters are identified and we find that reducing coast-time to as little as 500 ps still provides some benefit. Finally, we show how the high-foot implosion data is consistent with the above mentioned picture.

  16. Impact of first-principles properties of deuterium-tritium on inertial confinement fusion target designsa)

    Science.gov (United States)

    Hu, S. X.; Goncharov, V. N.; Boehly, T. R.; McCrory, R. L.; Skupsky, S.; Collins, L. A.; Kress, J. D.; Militzer, B.

    2015-05-01

    A comprehensive knowledge of the properties of high-energy-density plasmas is crucial to understanding and designing low-adiabat, inertial confinement fusion (ICF) implosions through hydrodynamic simulations. Warm-dense-matter (WDM) conditions are routinely accessed by low-adiabat ICF implosions, in which strong coupling and electron degeneracy often play an important role in determining the properties of warm dense plasmas. The WDM properties of deuterium-tritium (DT) mixtures and ablator materials, such as the equation of state, thermal conductivity, opacity, and stopping power, were usually estimated by models in hydro-codes used for ICF simulations. In these models, many-body and quantum effects were only approximately taken into account in the WMD regime. Moreover, the self-consistency among these models was often missing. To examine the accuracy of these models, we have systematically calculated the static, transport, and optical properties of warm dense DT plasmas, using first-principles (FP) methods over a wide range of densities and temperatures that cover the ICF "path" to ignition. These FP methods include the path-integral Monte Carlo (PIMC) and quantum-molecular dynamics (QMD) simulations, which treat electrons with many-body quantum theory. The first-principles equation-of-state table, thermal conductivities (κQMD), and first principles opacity table of DT have been self-consistently derived from the combined PIMC and QMD calculations. They have been compared with the typical models, and their effects to ICF simulations have been separately examined in previous publications. In this paper, we focus on their combined effects to ICF implosions through hydro-simulations using these FP-based properties of DT in comparison with the usual model simulations. We found that the predictions of ICF neutron yield could change by up to a factor of ˜2.5; the lower the adiabat of DT capsules, the more variations in hydro-simulations. The FP-based properties of DT

  17. Impact of first-principles properties of deuterium–tritium on inertial confinement fusion target designs

    Energy Technology Data Exchange (ETDEWEB)

    Hu, S. X., E-mail: shu@lle.rochester.edu; Goncharov, V. N.; Boehly, T. R.; McCrory, R. L.; Skupsky, S. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Collins, L. A.; Kress, J. D. [Theoretical Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Militzer, B. [Department of Earth and Planetary Science and Department of Astronomy, University of California, Berkeley, California 94720 (United States)

    2015-05-15

    A comprehensive knowledge of the properties of high-energy-density plasmas is crucial to understanding and designing low-adiabat, inertial confinement fusion (ICF) implosions through hydrodynamic simulations. Warm-dense-matter (WDM) conditions are routinely accessed by low-adiabat ICF implosions, in which strong coupling and electron degeneracy often play an important role in determining the properties of warm dense plasmas. The WDM properties of deuterium–tritium (DT) mixtures and ablator materials, such as the equation of state, thermal conductivity, opacity, and stopping power, were usually estimated by models in hydro-codes used for ICF simulations. In these models, many-body and quantum effects were only approximately taken into account in the WMD regime. Moreover, the self-consistency among these models was often missing. To examine the accuracy of these models, we have systematically calculated the static, transport, and optical properties of warm dense DT plasmas, using first-principles (FP) methods over a wide range of densities and temperatures that cover the ICF “path” to ignition. These FP methods include the path-integral Monte Carlo (PIMC) and quantum-molecular dynamics (QMD) simulations, which treat electrons with many-body quantum theory. The first-principles equation-of-state table, thermal conductivities (κ{sub QMD}), and first principles opacity table of DT have been self-consistently derived from the combined PIMC and QMD calculations. They have been compared with the typical models, and their effects to ICF simulations have been separately examined in previous publications. In this paper, we focus on their combined effects to ICF implosions through hydro-simulations using these FP-based properties of DT in comparison with the usual model simulations. We found that the predictions of ICF neutron yield could change by up to a factor of ∼2.5; the lower the adiabat of DT capsules, the more variations in hydro-simulations. The FP

  18. 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.

  19. Fusion-power demonstration. [Next step beyond MFTF-B

    Energy Technology Data Exchange (ETDEWEB)

    Henning, C.D.; Logan, B.G.; Carlson, G.A.; Neef, W.S.; Moir, R.W.; Campbell, R.B.; Botwin, R.; Clarkson, I.R.; Carpenter, T.J.

    1983-03-29

    As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment.

  20. (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.

  1. 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.

  2. Study on discharge plasma in a cylindrical inertial electrostatic confinement fusion device

    Science.gov (United States)

    Buzarbaruah, N.; Dutta, N. J.; Borgohain, D.; Mohanty, S. R.; Bailung, H.

    2017-08-01

    Deuterium plasma has been produced in a cylindrical inertial electrostatic confinement fusion device using hot and cold cathode discharges and the plasma parameters are determined by employing an electrostatic probe. The plasma temperature and density are estimated at optimum experimental conditions and it is noted that the plasma temperature is 3 eV in the case of hot cathode discharge whereas 10 eV in the case of the cold cathode discharge. The plasma density as determined is two orders more in the case of the hot cathode discharge than the other. The probe is also used to observe the ion oscillation in the negative potential well that is formed in between the cathode grid and chamber (anode). The observation of spontaneous oscillation along with the harmonics has been reported.

  3. A strategy for reducing stagnation phase hydrodynamic instability growth in inertial confinement fusion implosions

    Energy Technology Data Exchange (ETDEWEB)

    Clark, D. S.; Robey, H. F.; Smalyuk, V. A. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

    2015-05-15

    Encouraging progress is being made in demonstrating control of ablation front hydrodynamic instability growth in inertial confinement fusion implosion experiments on the National Ignition Facility [E. I. Moses, R. N. Boyd, B. A. Remington, C. J. Keane, and R. Al-Ayat, Phys. Plasmas 16, 041006 (2009)]. Even once ablation front stabilities are controlled, however, instability during the stagnation phase of the implosion can still quench ignition. A scheme is proposed to reduce the growth of stagnation phase instabilities through the reverse of the “adiabat shaping” mechanism proposed to control ablation front growth. Two-dimensional radiation hydrodynamics simulations confirm that improved stagnation phase stability should be possible without compromising fuel compression.

  4. 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

    2016-10-05

    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.

  5. A new ignition scheme using hybrid indirect-direct drive for inertial confinement fusion

    CERN Document Server

    Fan, Zhengfeng; Dai, Zhensheng; Cai, Hong-bo; Zhu, Shao-ping; Zhang, W Y; He, X T

    2013-01-01

    A new hybrid indirect-direct-drive ignition scheme is proposed for inertial confinement fusion: a cryogenic capsule encased in a hohlraum is first compressed symmetrically by indirect-drive x-rays, and then accelerated and ignited by both direct-drive lasers and x-rays. A steady high-density plateau newly formed between the radiation and electron ablation fronts suppresses the rarefaction at the radiation ablation front and greatly enhances the drive pressure. Meanwhile, multiple shock reflections at the fuel/hot-spot interface are prevented during capsule deceleration. Thus rapid ignition and burn are realized. In comparison with the conventional indirect drive, the hybrid drive implodes the capsule with a higher velocity ($\\sim4.3\\times10^7$ cm/s) and a much lower convergence ratio ($\\sim$25), and the growth of hydrodynamic instabilities is significantly reduced, especially at the fuel/hot-spot interface.

  6. 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.

  7. 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.

  8. A Freon-filled bubble chamber for neutron detection in inertial confinement fusion experiments.

    Science.gov (United States)

    Ghilea, M C; Meyerhofer, D D; Sangster, T C

    2011-03-01

    Neutron imaging is one of the main methods used in inertial confinement fusion experiments to measure the core symmetry of target implosions. Previous studies have shown that bubble chambers have the potential to obtain higher resolution images of the targets for a shorter source-to-target distance than typical scintillator arrays. A bubble chamber for neutron imaging with Freon 115 as the active medium was designed and built for the OMEGA laser system. Bubbles resulting from spontaneous nucleation were recorded. Bubbles resulting from neutron-Freon interactions were observed at neutron yields of 10(13) emitted from deuterium-tritium target implosions on OMEGA. The measured column bubble density was too low for neutron imaging on OMEGA but agreed with the model of bubble formation. The recorded data suggest that neutron bubble detectors are a promising technology for the higher neutron yields expected at National Ignition Facility.

  9. Change in inertial confinement fusion implosions upon using an ab initio multiphase DT equation of state.

    Science.gov (United States)

    Caillabet, L; Canaud, B; Salin, G; Mazevet, S; Loubeyre, P

    2011-09-09

    Improving the description of the equation of state (EOS) of deuterium-tritium (DT) has recently been shown to change significantly the gain of an inertial confinement fusion target [S. X. Hu et al., Phys. Rev. Lett. 104, 235003 (2010)]. Here we use an advanced multiphase EOS, based on ab initio calculations, to perform a full optimization of the laser pulse shape with hydrodynamic simulations starting from 19 K in DT ice. The thermonuclear gain is shown to be a robust estimate over possible uncertainties of the EOS. Two different target designs are discussed, for shock ignition and self-ignition. In the first case, the areal density and thermonuclear energy can be recovered by slightly increasing the laser energy. In the second case, a lower in-flight adiabat is needed, leading to a significant delay (3 ns) in the shock timing of the implosion.

  10. Monoenergetic-proton-radiography measurements of implosion dynamics in direct-drive inertial-confinement fusion.

    Science.gov (United States)

    Li, C K; Séguin, F H; Rygg, J R; Frenje, J A; Manuel, M; Petrasso, R D; Betti, R; Delettrez, J; Knauer, J P; Marshall, F; Meyerhofer, D D; Shvarts, D; Smalyuk, V A; Stoeckl, C; Landen, O L; Town, R P J; Back, C A; Kilkenny, J D

    2008-06-06

    Time-gated, monoenergetic radiography with 15-MeV protons provides unique measurements of implosion dynamics in direct-drive inertial-confinement fusion. Images obtained during acceleration, coasting, deceleration, and stagnation display a comprehensive picture of spherical implosions. Critical information inferred from such images, hitherto unavailable, characterizes the spatial structure and temporal evolution of self-generated fields and plasma areal density. Results include the first observation of a radial electric field inside the imploding capsule. It is initially directed inward (at approximately 10(9) V/m), eventually reverses direction ( approximately 10(8) V/m), and is the probable consequence of the evolution of the electron pressure gradient.

  11. Evidence for stratification of deuterium-tritium fuel in inertial confinement fusion implosions.

    Science.gov (United States)

    Casey, D T; Frenje, J A; Johnson, M Gatu; Manuel, M J-E; Rinderknecht, H G; Sinenian, N; Séguin, F H; Li, C K; Petrasso, R D; Radha, P B; Delettrez, J A; Glebov, V Yu; Meyerhofer, D D; Sangster, T C; McNabb, D P; Amendt, P A; Boyd, R N; Rygg, J R; Herrmann, H W; Kim, Y H; Bacher, A D

    2012-02-17

    Measurements of the D(d,p)T (dd) and T(t,2n)(4)He (tt) reaction yields have been compared with those of the D(t,n)(4)He (dt) reaction yield, using deuterium-tritium gas-filled inertial confinement fusion capsule implosions. In these experiments, carried out on the OMEGA laser, absolute spectral measurements of dd protons and tt neutrons were obtained. From these measurements, it was concluded that the dd yield is anomalously low and the tt yield is anomalously high relative to the dt yield, an observation that we conjecture to be caused by a stratification of the fuel in the implosion core. This effect may be present in ignition experiments planned on the National Ignition Facility.

  12. 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.

  13. 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

    2016-12-09

    The first cryogenic deuterium and deuterium-tritium liquid layer implosions at the National Ignition Facility (NIF) demonstrate D_{2} and DT layer inertial confinement fusion (ICF) implosions that can access a low-to-moderate hot-spot convergence ratio (1230) 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.

  14. A technique for thick polymer coating of inertial-confinement-fusion targets

    Science.gov (United States)

    Lee, M. C.; Feng, I.-A.; Wang, T. G.; Kim, H.-G.

    1983-01-01

    A technique to coat a stalk-mounted inertial-confinement fusion (ICF) target with a thick polymer layer has been successfully demonstrated. The polymer solution is first atomized, allowed to coalesce into a droplet, and positioned in a stable acoustic levitating field. The stalk-mounted ICF target is then moved into the acoustic field by manipulating a 3-D positioner to penetrate the surface membrane of the droplet, thus immersing the target in the levitated coating solution. The target inside the droplet is maintained at the center of the levitated liquid using the 3-D positional information provided by two orthogonally placed TV cameras until the drying process is completed. The basic components of the experimental apparatus, including an acoustic levitator, liquid sample deployment device, image acquisition instrumentation, and 3-D positioner, are briefly described.

  15. Limitation on prepulse level for cone-guided fast-ignition inertial confinement fusion.

    Science.gov (United States)

    Macphee, A G; Divol, L; Kemp, A J; Akli, K U; Beg, F N; Chen, C D; Chen, H; Hey, D S; Fedosejevs, R J; Freeman, R R; Henesian, M; Key, M H; Le Pape, S; Link, A; Ma, T; Mackinnon, A J; Ovchinnikov, V M; Patel, P K; Phillips, T W; Stephens, R B; Tabak, M; Town, R; Tsui, Y Y; Van Woerkom, L D; Wei, M S; Wilks, S C

    2010-02-05

    The viability of fast-ignition (FI) inertial confinement fusion hinges on the efficient transfer of laser energy to the compressed fuel via multi-MeV electrons. Preformed plasma due to the laser prepulse strongly influences ultraintense laser plasma interactions and hot electron generation in the hollow cone of an FI target. We induced a prepulse and consequent preplasma in copper cone targets and measured the energy deposition zone of the main pulse by imaging the emitted K_{alpha} radiation. Simulation of the radiation hydrodynamics of the preplasma and particle in cell modeling of the main pulse interaction agree well with the measured deposition zones and provide an insight into the energy deposition mechanism and electron distribution. It was demonstrated that a under these conditions a 100 mJ prepulse eliminates the forward going component of approximately 2-4 MeV electrons.

  16. Effect of Laser-Plasma Interactions on Inertial Confinement Fusion Hohlraum Dynamics

    CERN Document Server

    Strozzi, D J; Michel, P; Divol, L; Sepke, S M; Kerbel, G D; Thomas, C A; Ralph, J E; Moody, J D; Schneider, M B

    2016-01-01

    The effects of laser-plasma interactions (LPI) on the dynamics of inertial confinement fusion hohlraums is investigated via a new approach that self-consistently couples reduced LPI models into radiation-hydrodynamics numerical codes. The interplay between hydrodynamics and LPI - specifically stimulated Raman scattering (SRS) and crossed-beam energy transfer (CBET) - mostly occurs via momentum and energy deposition into Langmuir and ion acoustic waves. This spatially redistributes energy coupling to the target, which affects the background plasma conditions and thus modifies the laser propagation. This model shows a reduction of CBET, and significant laser energy depletion by Langmuir waves, which reduce the discrepancy between modeling and data from hohlraum experiments on wall x-ray emission and capsule implosion shape.

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

    CERN Document Server

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

    2016-01-01

    A novel ignition hohlraum for indirect-drive inertial confinement fusion is proposed, which is named as three-axis cylindrical hohlraum (TACH). TACH is a kind of 6 laser entrance holes (LEHs) hohlraum, which is made of three cylindrical hohlraums orthogonally jointed. Laser beams are injected through every entrance hole with the same incident angle of 55{\\deg}. The view-factor simulation result shows that the time-varying drive asymmetry of TACH is no more than 1.0% in the whole drive pulse period without any supplementary technology such as beam phasing etc. Its coupling efficiency of TACH is close to that of 6 LEHs spherical hohlraum with corresponding size. Its plasma-filling time is close to typical cylindrical ignition hohlraum. Its laser plasma interaction has as low backscattering as the outer cone of the cylindrical ignition hohlraum. Therefore, the proposed hohlraum provides a competitive candidate for ignition hohlraum.

  18. Studies of plastic-ablator compressibility for direct-drive inertial confinement fusion on OMEGA.

    Science.gov (United States)

    Hu, S X; Smalyuk, V A; Goncharov, V N; Knauer, J P; Radha, P B; Igumenshchev, I V; Marozas, J A; Stoeckl, C; Yaakobi, B; Shvarts, D; Sangster, T C; McKenty, P W; Meyerhofer, D D; Skupsky, S; McCrory, R L

    2008-05-09

    The compression of planar plastic targets was studied with x-ray radiography in the range of laser intensities of I approximately 0.5 to 1.5x10(15) W/cm2 using square (low-compression) and shaped (high-compression) pulses. Two-dimensional simulations with the radiative hydrocode DRACO show good agreement with measurements at laser intensities up to I approximately 10(15) W/cm2. These results provide the first experimental evidence for low-entropy, adiabatic compression of plastic shells in the laser intensity regime relevant to direct-drive inertial confinement fusion. A density reduction near the end of the drive at a high intensity of I approximately 1.5x10(15) W/cm2 has been correlated with the hard x-ray signal caused by hot electrons from two-plasmon-decay instability.

  19. A highly efficient neutron time-of-flight detector for inertial confinement fusion experiments

    Science.gov (United States)

    Izumi, N.; Yamaguchi, K.; Yamagajo, T.; Nakano, T.; Kasai, T.; Urano, T.; Azechi, H.; Nakai, S.; Iida, T.

    1999-01-01

    We have developed the highly efficient neutron detector system MANDALA for the inertial-confinement-fusion experiment. The MANDALA system consists of 842 elements plastic scintillation detectors and data acquisition electronics. The detection level is the yield of 1.2×105 for 2.5 MeV and 1×105 for 14.1 MeV neutrons (with 100 detected hits). We have calibrated the intrinsic detection efficiencies of the detector elements using a neutron generator facility. Timing calibration and integrity test of the system were also carried out with a 60Co γ ray source. MANDALA system was applied to the implosion experiments at the GEKKO XII laser facility. The integrity test was carried out by implosion experiments.

  20. 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.

  1. KULL: LLNL's ASCI Inertial Confinement Fusion Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J. A.; Miller, D. S.; Owen, J. M.; Zike, M. R.; Eltgroth, P. G.; Madsen, N. K.; McCandless, K. P.; Nowak, P. F.; Nemanic, M. K.; Gentile, N. A.; Stuart, L. M.; Keen, N. D.; Palmer, T. S.

    2000-01-10

    KULL is a three dimensional, time dependent radiation hydrodynamics simulation code under development at Lawrence Livermore National Laboratory. A part of the U.S. Department of Energy's Accelerated Strategic Computing Initiative (ASCI), KULL's purpose is to simulate the physical processes in Inertial Confinement Fusion (ICF) targets. The National Ignition Facility, where ICF experiments will be conducted, and ASCI are part of the experimental and computational components of DOE's Stockpile Stewardship Program. This paper provides an overview of ASCI and describes KULL, its hydrodynamic simulation capability and its three methods of simulating radiative transfer. Particular emphasis is given to the parallelization techniques essential to obtain the performance required of the Stockpile Stewardship Program and to exploit the massively parallel processor machines that ASCI is procuring.

  2. 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.

  3. 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

  4. Modular control of fusion power heating applications

    Energy Technology Data Exchange (ETDEWEB)

    Demers, D. R.

    2012-08-24

    This work is motivated by the growing demand for auxiliary heating on small and large machines worldwide. Numerous present and planned RF experiments (EBW, Lower Hybrid, ICRF, and ECH) are increasingly complex systems. The operational challenges are indicative of a need for components of real-time control that can be implemented with a moderate amount of effort in a time- and cost-effective fashion. Such a system will improve experimental efficiency, enhance experimental quality, and expedite technological advancements. The modular architecture of this control-suite serves multiple purposes. It facilitates construction on various scales from single to multiple controller systems. It enables expandability of control from basic to complex via the addition of modules with varying functionalities. It simplifies the control implementation process by reducing layers of software and electronic development. While conceived with fusion applications in mind, this suite has the potential to serve a broad range of scientific and industrial applications. During the Phase-I research effort we established the overall feasibility of this modular control-suite concept. We developed the fundamental modules needed to implement open-loop active-control and demonstrated their use on a microwave power deposition experiment.

  5. Advances in Tandem Mirror fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  6. Fusion Power Demonstrations I and II

    Energy Technology Data Exchange (ETDEWEB)

    Doggett, J.N. (ed.)

    1985-01-01

    In this report we present a summary of the first phase of the Fusion Power Demonstration (FPD) design study. During this first phase, we investigated two configurations, performed detailed studies of major components, and identified and examined critical issues. In addition to these design specific studies, we also assembled a mirror-systems computer code to help optimize future device designs. The two configurations that we have studied are based on the MARS magnet configuration and are labeled FPD-I and FPD-II. The FPD-I configuration employs the same magnet set used in the FY83 FPD study, whereas the FPD-II magnets are a new, much smaller set chosen to help reduce the capital cost of the system. As part of the FPD study, we also identified and explored issues critical to the construction of an Engineering Test Reactor (ETR). These issues involve subsystems or components, which because of their cost or state of technology can have a significant impact on our ability to meet FPD's mission requirements on the assumed schedule. General Dynamics and Grumman Aerospace studied two of these systems, the high-field choke coil and the halo pump/direct converter, in great detail and their findings are presented in this report.

  7. Inertial Fusion Power Plant Concept of Operations and Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Anklam, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Knutson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunne, A. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kasper, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheehan, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Lang, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Roberts, V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mau, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-01-15

    Parsons and LLNL scientists and engineers performed design and engineering work for power plant pre-conceptual designs based on the anticipated laser fusion demonstrations at the National Ignition Facility (NIF). Work included identifying concepts of operations and maintenance (O&M) and associated requirements relevant to fusion power plant systems analysis. A laser fusion power plant would incorporate a large process and power conversion facility with a laser system and fusion engine serving as the heat source, based in part on some of the systems and technologies advanced at NIF. Process operations would be similar in scope to those used in chemical, oil refinery, and nuclear waste processing facilities, while power conversion operations would be similar to those used in commercial thermal power plants. While some aspects of the tritium fuel cycle can be based on existing technologies, many aspects of a laser fusion power plant presents several important and unique O&M requirements that demand new solutions. For example, onsite recovery of tritium; unique remote material handling systems for use in areas with high radiation, radioactive materials, or high temperatures; a five-year fusion engine target chamber replacement cycle with other annual and multi-year cycles anticipated for major maintenance of other systems, structures, and components (SSC); and unique SSC for fusion target waste recycling streams. This paper describes fusion power plant O&M concepts and requirements, how O&M requirements could be met in design, and how basic organizational and planning issues can be addressed for a safe, reliable, economic, and feasible fusion power plant.

  8. Inertial fusion power plant concept of operations and maintenance

    Science.gov (United States)

    Knutson, Brad; Dunne, Mike; Kasper, Jack; Sheehan, Timothy; Lang, Dwight; Anklam, Tom; Roberts, Valerie; Mau, Derek

    2015-02-01

    Parsons and LLNL scientists and engineers performed design and engineering work for power plant pre-conceptual designs based on the anticipated laser fusion demonstrations at the National Ignition Facility (NIF). Work included identifying concepts of operations and maintenance (O&M) and associated requirements relevant to fusion power plant systems analysis. A laser fusion power plant would incorporate a large process and power conversion facility with a laser system and fusion engine serving as the heat source, based in part on some of the systems and technologies advanced at NIF. Process operations would be similar in scope to those used in chemical, oil refinery, and nuclear waste processing facilities, while power conversion operations would be similar to those used in commercial thermal power plants. While some aspects of the tritium fuel cycle can be based on existing technologies, many aspects of a laser fusion power plant presents several important and unique O&M requirements that demand new solutions. For example, onsite recovery of tritium; unique remote material handling systems for use in areas with high radiation, radioactive materials, or high temperatures; a five-year fusion engine target chamber replacement cycle with other annual and multi-year cycles anticipated for major maintenance of other systems, structures, and components (SSC); and unique SSC for fusion target waste recycling streams. This paper describes fusion power plant O&M concepts and requirements, how O&M requirements could be met in design, and how basic organizational and planning issues can be addressed for a safe, reliable, economic, and feasible fusion power plant.

  9. Health physics aspects of fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Easterly, C.E.; Shank, K.E.; Shoup, R.L.

    1976-10-01

    Fusion reactor technology is presently in conceptual and early developmental stages. Concomitant with hardware development, potential health and environmental impacts must be evaluated. This evaluation is necessary to ensure that technologists have pertinent information available such that adequate consideration is given to health and environmental problems. Problem areas attendant to tritium, activation products, and magnetic fields associated with fusion reactor systems are discussed.

  10. High-resolution modeling of indirectly driven high-convergence layered inertial confinement fusion capsule implosions

    Science.gov (United States)

    Haines, Brian M.; Aldrich, C. H.; Campbell, J. M.; Rauenzahn, R. M.; Wingate, C. A.

    2017-05-01

    In this paper, we present the results of high-resolution simulations of the implosion of high-convergence layered indirect-drive inertial confinement fusion capsules of the type fielded on the National Ignition Facility using the xRAGE radiation-hydrodynamics code. In order to evaluate the suitability of xRAGE to model such experiments, we benchmark simulation results against available experimental data, including shock-timing, shock-velocity, and shell trajectory data, as well as hydrodynamic instability growth rates. We discuss the code improvements that were necessary in order to achieve favorable comparisons with these data. Due to its use of adaptive mesh refinement and Eulerian hydrodynamics, xRAGE is particularly well suited for high-resolution study of multi-scale engineering features such as the capsule support tent and fill tube, which are known to impact the performance of high-convergence capsule implosions. High-resolution two-dimensional (2D) simulations including accurate and well-resolved models for the capsule fill tube, support tent, drive asymmetry, and capsule surface roughness are presented. These asymmetry seeds are isolated in order to study their relative importance and the resolution of the simulations enables the observation of details that have not been previously reported. We analyze simulation results to determine how the different asymmetries affect hotspot reactivity, confinement, and confinement time and how these combine to degrade yield. Yield degradation associated with the tent occurs largely through decreased reactivity due to the escape of hot fuel mass from the hotspot. Drive asymmetries and the fill tube, however, degrade yield primarily via burn truncation, as associated instability growth accelerates the disassembly of the hotspot. Modeling all of these asymmetries together in 2D leads to improved agreement with experiment but falls short of explaining the experimentally observed yield degradation, consistent with previous

  11. Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Werner, R.W.; Carlson, G.A.; Cornish, D.N.

    1976-04-01

    Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements. Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated.

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

    Energy Technology Data Exchange (ETDEWEB)

    Moses, E

    2009-10-15

    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.

  13. 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

  14. The ITER project - the road to fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Hay, Jennifer; Spears, Bill

    2007-07-01

    Commercial fusion power promises safe, CO2-free, environmentally benign, large-scale energy production, with abundant and widely available fuels. Fusion researchers are ready to take the next step towards this goal, with the construction of ITER, which should demonstrate the technical and scientific feasibility of fusion power production. The signing, by seven Parties representing more than half of humanity, of the ITER Joint Implementation Agreement on 21 November 2006, established the ITER Organization (pending ratification by some of the ITER Parties), and opens the way to the construction phase of the project. This paper describes the history of the project, its objectives, schedule and organization, and its place on the road to the realisation of commercial fusion power. (auth)

  15. Fusion Breeding for Sustainable, Mid Century, Carbon Free Power

    Science.gov (United States)

    Manheimer, Wallace

    2015-11-01

    If ITER achieves Q ~10, it is still very far from useful fusion. The fusion power, and the driver power will allow only a small amount of power to be delivered, power producers. Considering the status of other magnetic fusion concepts, it is also very unlikely that any alternate concept will either. Laser fusion does not seem to be constrained by any conservative design rules, but considering the failure of NIF to achhieve ignition, at this point it has many more obstacles to overcome than magnetic fusion. One way out of this dilemma is to use an ITER size tokamak, or a NIF size laser, as a fuel breeder for searate nuclear reactors. Hence ITER and NIF become ends in themselves, instead of steps to who knows what DEMO decades later. Such a tokamak can easily live within the consrtaints of conservative design rules. This has led the author to propose ``The Energy Park'' a sustainable, carbon free, economical, and environmently viable power source without prolifertion risk. It is one fusion breeder fuels 5 conventional nuclear reactors, and one fast neutron reactor burns the actinide wastes.

  16. A new time and space resolved transmission spectrometer for research in inertial confinement fusion and radiation source development.

    Science.gov (United States)

    Knapp, P F; Ball, C; Austin, K; Hansen, S B; Kernaghan, M D; Lake, P W; Ampleford, D J; McPherson, L A; Sandoval, D; Gard, P; Wu, M; Bourdon, C; Rochau, G A; McBride, R D; Sinars, D B

    2017-01-01

    We describe the design and function of a new time and space resolved x-ray spectrometer for use in Z-pinch inertial confinement fusion and radiation source development experiments. The spectrometer is designed to measure x-rays in the range of 0.5-1.5 Å (8-25 keV) with a spectral resolution λ/Δλ ∼ 400. The purpose of this spectrometer is to measure the time- and one-dimensional space-dependent electron temperature and density during stagnation. These relatively high photon energies are required to escape the dense plasma created at stagnation and to obtain sensitivity to electron temperatures ≳3 keV. The spectrometer is of the Cauchois type, employing a large 30 × 36 mm(2), transmissive quartz optic for which a novel solid beryllium holder was designed. The performance of the crystal was verified using offline tests, and the integrated system was tested using experiments on the Z pulsed power accelerator.

  17. Review of the safety concept for fusion reactor concepts and transferability of the nuclear fission regulation to potential fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Raeder, Juergen; Weller, Arthur; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik (IPP), Garching (Germany); Jin, Xue Zhou; Boccaccini, Lorenzo V.; Stieglitz, Robert; Carloni, Dario [Karlsruher Institute fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany); Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany); Herb, Joachim [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

    2016-01-15

    This paper summarizes the current state of the art in science and technology of the safety concept for future fusion power plants (FPPs) and examines the transferability of the current nuclear fission regulation to the concepts of future fusion power plants. At the moment there exist only conceptual designs of future fusion power plants. The most detailed concepts with regards to safety aspects were found in the European Power Plant Conceptual Study (PPCS). The plant concepts discussed in the PPCS are based on magnetic confinement of the plasma. The safety concept of fusion power plants, which has been developed during the last decades, is based on the safety concepts of installations with radioactive inventories, especially nuclear fission power plants. It applies the concept of defence in depth. However, there are specific differences between the implementations of the safety concepts due to the physical and technological characteristics of fusion and fission. It is analysed whether for fusion a safety concept is required comparable to the one of fission. For this the consequences of a purely hypothetical release of large amounts of the radioactive inventory of a fusion power plant and a fission power plant are compared. In such an event the evacuation criterion outside the plant is exceeded by several orders of magnitude for a fission power plant. For a fusion power plant the expected radiological consequences are of the order of the evacuation criterion. Therefore, a safety concept is also necessary for fusion to guarantee the confinement of the radioactive inventory. The comparison between the safety concepts for fusion and fission shows that the fundamental safety function ''confinement of the radioactive materials'' can be transferred directly in a methodical way. For a fusion power plant this fundamental safety function is based on both, physical barriers as well as on active retention functions. After the termination of the fusion

  18. Liquid lithium loop system to solve challenging technology issues for fusion power plant

    Science.gov (United States)

    Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

    2017-11-01

    Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle

  19. A Model for the Growth of Localized Shell Features in Inertial Confinement Fusion Implosions

    Science.gov (United States)

    Goncharov, V. N.

    2017-10-01

    Engineering features and target debris on inertial confinement fusion capsules play detrimental role in target performance. The contact points of such features with target surface as well as shadowing effects produce localized shell nonuniformities that grow in time because of the Rayleigh-Taylor instability developed during shell acceleration. Such growth leads to significant mass modulation in the shell and injection of ablator and cold fuel material into the target vapor region. These effects are commonly modeled using 2-D and 3-D hydrodynamic codes that take into account multiple physics effects. Such simulations, however, are very challenging since in many cases they are inherently three dimensional (as in the case of fill tube or stalk shadowing) and require very high grid resolution to accurately model short-scale features. To gain physics insight, an analytic model describing the growth of these features has been developed. The model is based on the Layzer-type approach. The talk will discuss the results of the model used to study perturbation growth seeded by localized target debris, glue spots, fill tubes, and stalks. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  20. 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.

  1. JRTF: A Flexible Software Framework for Real-Time Control in Magnetic Confinement Nuclear Fusion Experiments

    Science.gov (United States)

    Zhang, M.; Zheng, G. Z.; Zheng, W.; Chen, Z.; Yuan, T.; Yang, C.

    2016-04-01

    The magnetic confinement nuclear fusion experiments require various real-time control applications like plasma control. ITER has designed the Fast Plant System Controller (FPSC) for this job. ITER provided hardware and software standards and guidelines for building a FPSC. In order to develop various real-time FPSC applications efficiently, a flexible real-time software framework called J-TEXT real-time framework (JRTF) is developed by J-TEXT tokamak team. JRTF allowed developers to implement different functions as independent and reusable modules called Application Blocks (AB). The AB developers only need to focus on implementing the control tasks or the algorithms. The timing, scheduling, data sharing and eventing are handled by the JRTF pipelines. JRTF provides great flexibility on developing ABs. Unit test against ABs can be developed easily and ABs can even be used in non-JRTF applications. JRTF also provides interfaces allowing JRTF applications to be configured and monitored at runtime. JRTF is compatible with ITER standard FPSC hardware and ITER (Control, Data Access and Communication) CODAC Core software. It can be configured and monitored using (Experimental Physics and Industrial Control System) EPICS. Moreover the JRTF can be ported to different platforms and be integrated with supervisory control software other than EPICS. The paper presents the design and implementation of JRTF as well as brief test results.

  2. Inertial confinement fusion. ICF quarterly report, October 1993--December 1993, Volume 4, Number 1

    Energy Technology Data Exchange (ETDEWEB)

    Powell, H.T.; Schleich, D.P.; Murphy, P.W. [eds.

    1994-05-01

    In the 1990 National Academy of Sciences (NAS) report of its review of the U.S. Inertial Confinement Fusion (ICF) Program, it was recommended that a high priority be placed on completing the Precision Nova Project and its associated experimental campaign. Since fiscal year 1990, the lab has therefore campaigned vigorously on Nova and in its supporting laboratories to develop the Precision Nova capabilities needed to perform the stressful target experiments recommended in the 1990 NAS report. The activities to enable these experiments have been directed at improvements in three areas - the Nova laser, target fabrication capabilities, and target diagnostics. As summarized in the five articles in this report, the Precision Nova improvements have been successfully completed. These improvements have had a positive impact on target performance and on the ability to diagnose the results, as evidenced by the HEP-1 experimental results. The five articles generally concentrate on improvements to the capabilities rather than on the associated target physics experiments. Separate abstracts are included for each paper.

  3. 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.

  4. Development of the large neutron imaging system for inertial confinement fusion experiments.

    Science.gov (United States)

    Caillaud, T; Landoas, O; Briat, M; Kime, S; Rossé, B; Thfoin, I; Bourgade, J L; Disdier, L; Glebov, V Yu; Marshall, F J; Sangster, T C

    2012-03-01

    Inertial confinement fusion (ICF) requires a high resolution (~10 μm) neutron imaging system to observe deuterium and tritium (DT) core implosion asymmetries. A new large (150 mm entrance diameter: scaled for Laser MégaJoule [P. A. Holstein, F. Chaland, C. Charpin, J. M. Dufour, H. Dumont, J. Giorla, L. Hallo, S. Laffite, G. Malinie, Y. Saillard, G. Schurtz, M. Vandenboomgaerde, and F. Wagon, Laser and Particle Beams 17, 403 (1999)]) neutron imaging detector has been developed for such ICF experiments. The detector has been fully characterized using a linear accelerator and a (60)Co γ-ray source. A penumbral aperture was used to observe DT-gas-filled target implosions performed on the OMEGA laser facility. [T. R. Boehly, D. L. Brown, R. S. Craxton, R. L. Keck, J. P. Knauer, J. H. Kelly, T. J. Kessler, S. A. Kumpan, S. J. Loucks, S. A. Letzring, F. J. Marshall, R. L. McCrory, S. F. B. Morse, W. Seka, J. M. Soures, and C. P. Verdon, Opt. Commun. 133, 495 (1997)] Neutron core images of 14 MeV with a resolution of 15 μm were obtained and are compared to x-ray images of comparable resolution.

  5. Ion separation effects in mixed-species ablators for inertial-confinement-fusion implosions.

    Science.gov (United States)

    Amendt, Peter; Bellei, Claudio; Ross, J Steven; Salmonson, Jay

    2015-02-01

    Recent efforts to demonstrate significant self-heating of the fuel and eventual ignition at the National Ignition Facility make use of plastic (CH) ablators [O. A. Hurricane et al., Phys. Plasmas 21, 056314 (2014)]. Mainline simulation techniques for modeling CH capsule implosions treat the ablator as an average-atom fluid and neglect potential species separation phenomena. The mass-ablation process for a mixture is shown to lead to the potential for species separation, parasitic energy loss according to thermodynamic arguments, and reduced rocket efficiency. A generalized plasma barometric formula for a multispecies concentration gradient that includes collisionality and steady flows in spherical geometry is presented. A model based on plasma expansion into a vacuum is used to interpret reported experimental evidence for ablator species separation in an inertial-confinement-fusion target [J. S. Ross et al., Rev. Sci. Instrum. 83, 10E323 (2012)]. The possibility of "runaway" hydrogen ions in the thermoelectric field of the ablation front is conjectured.

  6. Development of backlighting sources for a Compton radiography diagnostic of inertial confinement fusion targets (invited).

    Science.gov (United States)

    Tommasini, R; MacPhee, A; Hey, D; Ma, T; Chen, C; Izumi, N; Unites, W; MacKinnon, A; Hatchett, S P; Remington, B A; Park, H S; Springer, P; Koch, J A; Landen, O L; Seely, John; Holland, Glenn; Hudson, Larry

    2008-10-01

    We present scaled demonstrations of backlighter sources, emitting bremsstrahlung x rays with photon energies above 75 keV, that we will use to record x-ray Compton radiographic snapshots of cold dense DT fuel in inertial confinement fusion implosions at the National Ignition Facility (NIF). In experiments performed at the Titan laser facility at Lawrence Livermore National Laboratory, we measured the source size and the bremsstrahlung spectrum as a function of laser intensity and pulse length from solid targets irradiated at 2x10(17)-5x10(18) W/cm(2) using 2-40 ps pulses. Using Au planar foils we achieved source sizes down to 5.5 microm and conversion efficiencies of about 1x10(-13) J/J into x-ray photons with energies in the 75-100 keV spectral range. We can now use these results to design NIF backlighter targets and shielding and to predict Compton radiography performance as a function of the NIF implosion yield and associated background.

  7. 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.

  8. Collection of solid and gaseous samples to diagnose inertial confinement fusion implosions.

    Science.gov (United States)

    Stoyer, M A; Velsko, C A; Spears, B K; Hicks, D G; Hudson, G B; Sangster, T C; Freeman, C G

    2012-02-01

    Collection of representative samples of debris following inertial confinement fusion implosions in order to diagnose implosion conditions and efficacy is a challenging endeavor because of the unique conditions within the target chamber such as unconverted laser light, intense pulse of x-rays, physical chunks of debris, and other ablative effects. We present collection of gas samples following an implosion for the first time. High collection fractions for noble gases were achieved. We also present collection of solid debris samples on flat plate collectors. Geometrical collection efficiencies for Au hohlraum material were achieved and collection of capsule debris (Be and Cu) was also observed. Asymmetric debris distributions were observed for Au and Be samples. Collection of Be capsule debris was higher for solid collectors viewing the capsule through the laser entrance hole in the hohlraum than for solid collectors viewing the capsule around the waist of the hohlraum. Collection of Au hohlraum material showed the opposite pattern: more Au debris was collected around the waist than through the laser entrance hole. The solid debris collectors were not optimized for minimal Cu backgrounds, which limited the conclusions about the symmetry of the Cu debris. The quality of the data limited conclusions on chemical fractionation effects within the burning, expanding, and then cooling plasma.

  9. 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.

  10. Mitigating laser imprint in direct-drive inertial confinement fusion implosions with high-Z dopants.

    Science.gov (United States)

    Hu, S X; Fiksel, G; Goncharov, V N; Skupsky, S; Meyerhofer, D D; Smalyuk, V A

    2012-05-11

    Nonuniformities seeded by both long- and short-wavelength laser perturbations can grow via Rayleigh-Taylor (RT) instability in direct-drive inertial confinement fusion, leading to performance reduction in low-adiabat implosions. To mitigate the effect of laser imprinting on target performance, spherical RT experiments have been performed on OMEGA using Si- or Ge-doped plastic targets in a cone-in-shell configuration. Compared to a pure plastic target, radiation preheating from these high-Z dopants (Si/Ge) increases the ablation velocity and the standoff distance between the ablation front and laser-deposition region, thereby reducing both the imprinting efficiency and the RT growth rate. Experiments showed a factor of 2-3 reduction in the laser-imprinting efficiency and a reduced RT growth rate, leading to significant (3-5 times) reduction in the σ(rms) of shell ρR modulation for Si- or Ge-doped targets. These features are reproduced by radiation-hydrodynamics simulations using the two-dimensional hydrocode DRACO.

  11. Indirect-direct hybrid-drive work-dominated hotspot ignition for inertial confinement fusion

    CERN Document Server

    He, X T; Li, J W; Liu, J; Lan, K; Wu, J F; Wang, L F; Ye, W H

    2015-01-01

    An indirect-direct hybrid-drive work-dominated hotspot ignition scheme for inertial confinement fusion is proposed: a layered fuel capsule inside a spherical hohlraum with an octahedral symmetry is compressed first by indirect-drive soft-x rays (radiation) and then by direct-drive lasers in last pulse duration. In this scheme, an enhanced shock and a follow-up compression wave for ignition with pressure far greater than the radiation ablation pressure are driven by the direct-drive lasers, and provide large pdV work to the hotspot to perform the work-dominated ignition. The numerical simulations show that the enhanced shock stops the reflections of indirect-drive shock at the main fuel-hotspot interface, and therefore significantly suppresses the hydrodynamic instabilities and asymmetry. Based on the indirect-drive implosion dynamics the hotspot is further compressed and heated by the enhanced shock and follow-up compression wave, resulting in the work-dominated hotspot ignition and burn with a maximal implos...

  12. Neutron flux assessment of a neutron irradiation facility based on inertial electrostatic confinement fusion.

    Science.gov (United States)

    Sztejnberg Gonçalves-Carralves, M L; Miller, M E

    2015-12-01

    Neutron generators based on inertial electrostatic confinement fusion were considered for the design of a neutron irradiation facility for explanted organ Boron Neutron Capture Therapy (BNCT) that could be installed in a health care center as well as in research areas. The chosen facility configuration is "irradiation chamber", a ~20×20×40 cm(3) cavity near or in the center of the facility geometry where samples to be irradiated can be placed. Neutron flux calculations were performed to study different manners for improving scattering processes and, consequently, optimize neutron flux in the irradiation position. Flux distributions were assessed through numerical simulations of several models implemented in MCNP5 particle transport code. Simulation results provided a wide spectrum of combinations of net fluxes and energy spectrum distributions. Among them one can find a group that can provide thermal neutron fluxes per unit of production rate in a range from 4.1·10(-4) cm(-2) to 1.6·10(-3) cm(-2) with epithermal-to-thermal ratios between 0.3% and 13% and fast-to-thermal ratios between 0.01% to 8%. Neutron generators could be built to provide more than 10(10) n s(-1) and, consequently, with an arrangement of several generators appropriate enough neutron fluxes could be obtained that would be useful for several BNCT-related irradiations and, eventually, for clinical practice. Copyright © 2015 Elsevier Ltd. All rights reserved.

  13. 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.

  14. Controlling stimulated Raman scattering by two-color light in inertial confinement fusion

    Science.gov (United States)

    Liu, Z. J.; Chen, Y. H.; Zheng, C. Y.; Cao, L. H.; Li, B.; Xiang, J.; Hao, L.; Lan, K.

    2017-08-01

    A method is proposed to control the stimulated Raman scattering in the inertial confinement fusion by using auxiliary 2ω light to suppress the stimulated Raman scattering of the 3ω light. In this scheme, inverse bremsstrahlung absorption and parametric instabilities in the 2ω light increase the electron temperature and the plasma-density fluctuation, thus preventing the development of Raman scattering of the 3ω light. This scheme is successfully demonstrated by both one-dimensional kinetic simulations and two-dimensional radiative hydrodynamic simulations. The one-dimensional Vlasov results show that the time-averaged transmissivity of the 3ω light increases from 0.75 to 0.95 under certain conditions. Results obtained using the particle-in-cell method with Monte Carlo collisions show that the electron temperature is greatly increased with the increasing intensity of the 2ω light. The two-dimensional radiative hydrodynamic simulation results show that the electron temperature increases from 3.2 keV to 3.5 keV, and the time-averaged backscattering level decreases from 0.28 to 0.1 in the presence of the auxiliary 2ω light.

  15. Implosion experiments using glass ablators for direct-drive inertial confinement fusion.

    Science.gov (United States)

    Smalyuk, V A; Betti, R; Delettrez, J A; Glebov, V Yu; Meyerhofer, D D; Radha, P B; Regan, S P; Sangster, T C; Sanz, J; Seka, W; Stoeckl, C; Yaakobi, B; Frenje, J A; Li, C K; Petrasso, R D; Séguin, F H

    2010-04-23

    Direct-drive implosions with 20-microm-thick glass shells were conducted on the Omega Laser Facility to test the performance of high-Z glass ablators for direct-drive, inertial confinement fusion. The x-ray signal caused by hot electrons generated by two-plasmon-decay instability was reduced by more than approximately 40x and hot-electron temperature by approximately 2x in the glass compared to plastic ablators at ignition-relevant drive intensities of approximately 1x10(15) W/cm2, suggesting reduced target preheat. The measured absorption and compression were close to 1D predictions. The measured soft x-ray production in the spectral range of approximately 2 to 4 keV was approximately 2x to 3x lower than 1D predictions, indicating that the shell preheat caused by soft x-rays is less than predicted. A direct-drive-ignition design based on glass ablators is introduced.

  16. A method for fine positioning of diagnostic packages in inertial confinement fusion experiments.

    Science.gov (United States)

    Wang, Wei; Shi, Tielin; Lee, Kok-Meng

    2011-12-01

    A method based on binocular vision servoing for positioning of a diagnostic package in inertial confinement fusion (ICF) experiments is presented. The general diagnostic instrument manipulator will provide precision three dimension positioning and alignment-to-target capability in ICF experiments. In this work, we focus on the final precise automatic positioning with a binocular vision system. A three dimension image projection vector (IPV), which has an almost linear relationship with the target position in 3D space under the condition of weak perspective, is introduced to extract target position information from binocular image. The difference of the IPV between the current image and the desired image will be used as the input of servo controller. A differential motion model was found for the hybrid manipulator with three degrees of freedom. With this model and the said IPV, the servo strategy will be dramatically simplified compare with general image based visual servo in which the image Jacobian matrix needs estimated online. The experiment result implies that the locating accuracy of the manipulator is less than two pixels. This method can also be used in micromanipulation visual servo field.

  17. 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.

  18. Inertial Electrostatic Confinement Fusion: The Laser Elevator Solar System Survey for Propellants Abstract

    Science.gov (United States)

    Pryor, Wayne

    1999-01-01

    Dr. Wayne Pryor worked on three projects this summer. These were: 1) Inertial Electrostatic Confinement; 2) The Laser Elevator; and 3) Solar System Survey for Propellants Abstract. We Assisted Jon Nadler from Richland Community College in assembling and operating a table-top nuclear fusion reactor. We successfully demonstrated neutron production in a deuterium plasma. Pryor also obtained basic spectroscopic information on the atomic and molecular emissions in the plasma. The second project consisted of the completion of a paper on a novel propulsion concept (due to Tom Meyer of Colorado, the first author): a laser sail that bounces light back to the laser source. Recycling the photons from source to sail perhaps 100-1000 times dramatically improves the energy efficiency of this system, which may become very important for high-velocity missions in the future. Lastly, we compiled a very basic inventory of solar system propellant resources, their locations, and their accessibility. This initial inventory concentrates on sunlight availability, water availability, and the difficulty (delta-velocity requirement and radiation environment) in getting there.

  19. First-principles equation of state of polystyrene and its effect on inertial confinement fusion implosions.

    Science.gov (United States)

    Hu, S X; Collins, L A; Goncharov, V N; Kress, J D; McCrory, R L; Skupsky, S

    2015-10-01

    Obtaining an accurate equation of state (EOS) of polystyrene (CH) is crucial to reliably design inertial confinement fusion (ICF) capsules using CH/CH-based ablators. With first-principles calculations, we have investigated the extended EOS of CH over a wide range of plasma conditions (ρ=0.1to100g/cm(3) and T=1000 to 4,000,000 K). When compared with the widely used SESAME-EOS table, the first-principles equation of state (FPEOS) of CH has shown significant differences in the low-temperature regime, in which strong coupling and electron degeneracy play an essential role in determining plasma properties. Hydrodynamic simulations of cryogenic target implosions on OMEGA using the FPEOS table of CH have predicted ∼30% decrease in neutron yield in comparison with the usual SESAME simulations. This is attributed to the ∼5% reduction in implosion velocity that is caused by the ∼10% lower mass ablation rate of CH predicted by FPEOS. Simulations using CH-FPEOS show better agreement with measurements of Hugoniot temperature and scattered light from ICF implosions.

  20. Development of the large neutron imaging system for inertial confinement fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Caillaud, T.; Landoas, O.; Briat, M.; Kime, S.; Rosse, B.; Thfoin, I.; Bourgade, J. L.; Disdier, L. [CEA, DAM, DIF, F-91297 Arpajon (France); Glebov, V. Yu.; Marshall, F. J.; Sangster, T. C. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States)

    2012-03-15

    Inertial confinement fusion (ICF) requires a high resolution ({approx}10 {mu}m) neutron imaging system to observe deuterium and tritium (DT) core implosion asymmetries. A new large (150 mm entrance diameter: scaled for Laser MegaJoule [P. A. Holstein, F. Chaland, C. Charpin, J. M. Dufour, H. Dumont, J. Giorla, L. Hallo, S. Laffite, G. Malinie, Y. Saillard, G. Schurtz, M. Vandenboomgaerde, and F. Wagon, Laser and Particle Beams 17, 403 (1999)]) neutron imaging detector has been developed for such ICF experiments. The detector has been fully characterized using a linear accelerator and a {sup 60}Co {gamma}-ray source. A penumbral aperture was used to observe DT-gas-filled target implosions performed on the OMEGA laser facility. [T. R. Boehly, D. L. Brown, R. S. Craxton, R. L. Keck, J. P. Knauer, J. H. Kelly, T. J. Kessler, S. A. Kumpan, S. J. Loucks, S. A. Letzring, F. J. Marshall, R. L. McCrory, S. F. B. Morse, W. Seka, J. M. Soures, and C. P. Verdon, Opt. Commun. 133, 495 (1997)] Neutron core images of 14 MeV with a resolution of 15 {mu}m were obtained and are compared to x-ray images of comparable resolution.

  1. X-ray ablation rates in inertial confinement fusion capsule materials

    Science.gov (United States)

    Olson, R. E.; Rochau, G. A.; Landen, O. L.; Leeper, R. J.

    2011-03-01

    X-ray ablation rates have been measured in beryllium, copper-doped beryllium, germanium-doped plastic (Ge-doped CH), and diamondlike high density carbon (HDC) for radiation temperatures T in the range of 160-260 eV. In beryllium, the measured ablation rates range from 3 to 12 mg/cm2/ns; in Ge-doped CH, the ablation rates range from 2 to 6 mg/cm2/ns; and for HDC, the rates range from 2 to 9 mg/cm2/ns. The ablation rates follow an approximate T3 dependence and, for T below 230 eV, the beryllium ablation rates are significantly higher than HDC and Ge-doped CH. The corresponding implied ablation pressures are in the range of 20-160 Mbar, scaling as T3.5. The results are found to be well predicted by computational simulations using the physics packages and computational techniques employed in the design of indirect-drive inertial confinement fusion capsules. An iterative rocket model has been developed and used to compare the ablation rate data set to spherical indirect-drive capsule implosion experiments and to confirm the validity of some aspects of proposed full-scale National Ignition Facility ignition capsule designs.

  2. 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.

  3. 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 NIF utilized high convergence (CR >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.

  4. Thermonuclear reactions probed at stellar-core conditions with laser-based inertial-confinement fusion

    Science.gov (United States)

    Casey, D. T.; Sayre, D. B.; Brune, C. R.; Smalyuk, V. A.; Weber, C. R.; Tipton, R. E.; Pino, J. E.; Grim, G. P.; Remington, B. A.; Dearborn, D.; Benedetti, L. R.; Frenje, J. A.; Gatu-Johnson, M.; Hatarik, R.; Izumi, N.; McNaney, J. M.; Ma, T.; Kyrala, G. A.; MacLaren, S.; Salmonson, J.; Khan, S. F.; Pak, A.; Hopkins, L. Berzak; Lepape, S.; Spears, B. K.; Meezan, N. B.; Divol, L.; Yeamans, C. B.; Caggiano, J. A.; McNabb, D. P.; Holunga, D. M.; Chiarappa-Zucca, M.; Kohut, T. R.; Parham, T. G.

    2017-12-01

    Stars are giant thermonuclear plasma furnaces that slowly fuse the lighter elements in the universe into heavier elements, releasing energy, and generating the pressure required to prevent collapse. To understand stars, we must rely on nuclear reaction rate data obtained, up to now, under conditions very different from those of stellar cores. Here we show thermonuclear measurements of the 2H(d, n)3He and 3H(t,2n)4He S-factors at a range of densities (1.2-16 g cm-3) and temperatures (2.1-5.4 keV) that allow us to test the conditions of the hydrogen-burning phase of main-sequence stars. The relevant conditions are created using inertial-confinement fusion implosions at the National Ignition Facility. Our data agree within uncertainty with previous accelerator-based measurements and establish this approach for future experiments to measure other reactions and to test plasma-nuclear effects present in stellar interiors, such as plasma electron screening, directly in the environments where they occur.

  5. First-principles investigations on ionization and thermal conductivity of polystyrene for inertial confinement fusion applications

    Science.gov (United States)

    Hu, S. X.; Collins, L. A.; Goncharov, V. N.; Kress, J. D.; McCrory, R. L.; Skupsky, S.

    2016-04-01

    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/cm3 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 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.

  6. First-principles equation of state of polystyrene and its effect on inertial confinement fusion implosions

    Science.gov (United States)

    Hu, S. X.; Collins, L. A.; Goncharov, V. N.; Kress, J. D.; McCrory, R. L.; Skupsky, S.

    2015-10-01

    Obtaining an accurate equation of state (EOS) of polystyrene (CH) is crucial to reliably design inertial confinement fusion (ICF) capsules using CH/CH-based ablators. With first-principles calculations, we have investigated the extended EOS of CH over a wide range of plasma conditions (ρ =0.1 to 100 g /cm3 and T =1000 to 4 000 000 K ). When compared with the widely used SESAME-EOS table, the first-principles equation of state (FPEOS) of CH has shown significant differences in the low-temperature regime, in which strong coupling and electron degeneracy play an essential role in determining plasma properties. Hydrodynamic simulations of cryogenic target implosions on OMEGA using the FPEOS table of CH have predicted ˜30% decrease in neutron yield in comparison with the usual SESAME simulations. This is attributed to the ˜5% reduction in implosion velocity that is caused by the ˜10% lower mass ablation rate of CH predicted by FPEOS. Simulations using CH-FPEOS show better agreement with measurements of Hugoniot temperature and scattered light from ICF implosions.

  7. Criteria for practical fusion power systems: Report from the EPRI fusion panel

    Science.gov (United States)

    Kaslow, J.; Brown, M.; Hirsch, R.; izzo, R.; McCann, J.; McCloud, D.; Muston, B.; Peterson, A.; Rosen, S.; Schneider, T.; Skrgic, P.; Snow, B.

    1994-09-01

    Electric utilities are keenly interested in the promise of fusion: large-scale electricity production anywhere, with virtually no natural resource depletion or environmental pollution. To expedite development of commercially viable fusion systems, the Electric Power Research Institute (EPRI)—the R&D wing of the U.S. electric utility industry—recently convened a panel of top utility R&D managers and executive officers to identify the key criteria that must be met by fusion plants in order to be acceptable to utilities. The panel's findings, summarized in this report, emphasize competitive economics, positive public perception, and regulatory simplicity.

  8. Diagnostic components in harsh radiation environments: possible overlap in R&D requirements of inertial confinement and magnetic fusion systems.

    Science.gov (United States)

    Bourgade, J L; Costley, A E; Reichle, R; Hodgson, E R; Hsing, W; Glebov, V; Decreton, M; Leeper, R; Leray, J L; Dentan, M; Hutter, T; Moroño, A; Eder, D; Shmayda, W; Brichard, B; Baggio, J; Bertalot, L; Vayakis, G; Moran, M; Sangster, T C; Vermeeren, L; Stoeckl, C; Girard, S; Pien, G

    2008-10-01

    The next generation of large scale fusion devices--ITER/LMJ/NIF--will require diagnostic components to operate in environments far more severe than those encountered in present facilities. This harsh environment is the result of high fluxes of neutrons, gamma rays, energetic ions, electromagnetic radiation, and in some cases, debris and shrapnel, at levels several orders of magnitude higher than those experienced in today's devices. The similarities and dissimilarities between environmental effects on diagnostic components for the inertial confinement and magnetic confinement fusion fields have been assessed. Areas in which considerable overlap have been identified are optical transmission materials and optical fibers in particular, neutron detection systems and electronics needs. Although both fields extensively use cables in the hostile environment, there is little overlap because the environments and requirements are very different.

  9. Growth and saturation of convective modes of the two-plasmon decay instability in inertial confinement fusion.

    Science.gov (United States)

    Yan, R; Maximov, A V; Ren, C; Tsung, F S

    2009-10-23

    Particle-in-cell (PIC) and fluid simulations of two-plasmon decay (TPD) instability under conditions relevant to inertial confinement fusion show the importance of convective modes. Growing at the lower density region, the convective modes can cause pump depletion and are energetically dominant in the nonlinear stage. The PIC simulations show that TPD saturates due to ion density fluctuations, which can turn off TPD by raising the instability threshold through mode coupling.

  10. 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

  11. Detailed high-resolution three-dimensional simulations of OMEGA separated reactants inertial confinement fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Haines, Brian M.; Grim, Gary P.; Fincke, James R.; Shah, Rahul C.; Forrest, Chad J.; Silverstein, Kevin; Marshall, Frederic J.; 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.

    2016-07-01

    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

  12. 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.

  13. Symmetric inertial confinement fusion implosions at ultra-high laser energies.

    Science.gov (United States)

    Glenzer, S H; MacGowan, B J; Michel, P; Meezan, N B; Suter, L J; Dixit, S N; Kline, J L; Kyrala, G A; Bradley, D K; Callahan, D A; Dewald, E L; Divol, L; Dzenitis, E; Edwards, M J; Hamza, A V; Haynam, C A; Hinkel, D E; Kalantar, D H; Kilkenny, J D; Landen, O L; Lindl, J D; LePape, S; Moody, J D; Nikroo, A; Parham, T; Schneider, M B; Town, R P J; Wegner, P; Widmann, K; Whitman, P; Young, B K F; Van Wonterghem, B; Atherton, L J; Moses, E I

    2010-03-05

    Indirect-drive hohlraum experiments at the National Ignition Facility have demonstrated symmetric capsule implosions at unprecedented laser drive energies of 0.7 megajoule. One hundred and ninety-two simultaneously fired laser beams heat ignition-emulate hohlraums to radiation temperatures of 3.3 million kelvin, compressing 1.8-millimeter-diameter capsules by the soft x-rays produced by the hohlraum. Self-generated plasma optics gratings on either end of the hohlraum tune the laser power distribution in the hohlraum, which produces a symmetric x-ray drive as inferred from the shape of the capsule self-emission. These experiments indicate that the conditions are suitable for compressing deuterium-tritium-filled capsules, with the goal of achieving burning fusion plasmas and energy gain in the laboratory.

  14. High-Energy-Density-Physics Studies for Inertial Confinement Fusion Applications

    Science.gov (United States)

    Hu, S. X.

    2017-10-01

    Accurate knowledge of the static, transport, and optical properties of high-energy-density (HED) plasmas is essential for reliably designing and understanding inertial confinement fusion (ICF) implosions. In the warm-dense-matter regime routinely accessed by low-adiabat ICF implosions, many-body strong-coupling and quantum electron degeneracy effects play an important role in determining plasma properties. The past several years have witnessed intense efforts to assess the importance of the microphysics of ICF targets, both theoretically and experimentally. On the theory side, first-principles methods based on quantum mechanics have been applied to investigate the properties of warm, dense plasmas. Specifically, self-consistent investigations have recently been performed on the equation of state, thermal conductivity, and opacity of a variety of ICF ablators such as polystyrene (CH), beryllium, carbon, and silicon over a wide range of densities and temperatures. In this talk, we will focus on the most-recent progress on these ab initio HED physics studies, which generally result in favorable comparisons with experiments. Upon incorporation into hydrocodes for ICF simulations, these first-principles ablator-plasma properties have produced significant differences over traditional models in predicting 1-D target performance of ICF implosions on OMEGA and direct-drive-ignition designs for the National Ignition Facility. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944. *In collaboration with L. A. Collins, T. R. Boehly, G. W. Collins, J. D. Kress, and V. N. Goncharov.

  15. 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).

  16. Fusion of ranging data from robot teams operating in confined areas

    Science.gov (United States)

    Lyons, Damian M.; Shrestha, Karma; Liu, Tsung-Ming

    2013-05-01

    We address the problem of fusing laser ranging data from multiple mobile robots that are surveying an area as part of a robot search and rescue or area surveillance mission. We are specifically interested in the case where members of the robot team are working in close proximity to each other. The advantage of this teamwork is that it greatly speeds up the surveying process; the area can be quickly covered even when the robots use a random motion exploration approach. However, the disadvantage of the close proximity is that it is possible, and even likely, that the laser ranging data from one robot include many depth readings caused by another robot. We refer to this as mutual interference. Using a team of two Pioneer 3-AT robots with tilted SICK LMS-200 laser sensors, we evaluate several techniques for fusing the laser ranging information so as to eliminate the mutual interference. There is an extensive literature on the mapping and localization aspect of this problem. Recent work on mapping has begun to address dynamic or transient objects. Our problem differs from the dynamic map problem in that we look at one kind of transient map feature, other robots, and we know that we wish to completely eliminate the feature. We present and evaluate three different approaches to the map fusion problem: a robot-centric approach, based on estimating team member locations; a map-centric approach, based on inspecting local regions of the map, and a combination of both approaches. We show results for these approaches for several experiments for a two robot team operating in a confined indoor environment .

  17. 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.

  18. First-principles equation-of-state table of deuterium for inertial confinement fusion applications

    Science.gov (United States)

    Hu, S. X.; Militzer, B.; Goncharov, V. N.; Skupsky, S.

    2011-12-01

    Understanding and designing inertial confinement fusion (ICF) implosions through radiation-hydrodynamics simulations relies on the accurate knowledge of the equation of state (EOS) of the deuterium and tritium fuels. To minimize the drive energy for ignition, the imploding shell of DT fuel must be kept as cold as possible. Such low-adiabat ICF implosions can access to coupled and degenerate plasma conditions, in which the analytical EOS models become inaccurate due to many-body effects. Using the path-integral Monte Carlo (PIMC) simulations we have derived a first-principles EOS (FPEOS) table of deuterium that covers typical ICF fuel conditions at densities ranging from 0.002 to 1596 g/cm3 and temperatures of 1.35 eV to 5.5 keV. We report the internal energy and the pressure and discuss the structure of the plasma in terms of pair-correlation functions. When compared with the widely used SESAME table and the revised Kerley03 table, discrepancies in the internal energy and in the pressure are identified for moderately coupled and degenerate plasma conditions. In contrast to the SESAME table, the revised Kerley03 table is in better agreement with our FPEOS results over a wide range of densities and temperatures. Although subtle differences still exist for lower temperatures (T hydrodynamics simulations of cryogenic ICF implosions using the FPEOS table and the Kerley03 table have resulted in similar results for the peak density, areal density (ρR), and neutron yield, which differ significantly from the SESAME simulations.

  19. First-principles opacity table of warm dense deuterium for inertial-confinement-fusion applications

    Science.gov (United States)

    Hu, S. X.; Collins, L. A.; Goncharov, V. N.; Boehly, T. R.; Epstein, R.; McCrory, R. L.; Skupsky, S.

    2014-09-01

    Accurate knowledge of the optical properties of a warm dense deuterium-tritium (DT) mixture is important for reliable design of inertial confinement fusion (ICF) implosions using radiation-hydrodynamics simulations. The opacity of a warm dense DT shell essentially determines how much radiation from hot coronal plasmas can be deposited in the DT fuel of an imploding capsule. Even for the simplest species of hydrogen, the accurate calculation of their opacities remains a challenge in the warm-dense matter regime because strong-coupling and quantum effects play an important role in such plasmas. With quantum-molecular-dynamics (QMD) simulations, we have derived a first-principles opacity table (FPOT) of deuterium (and the DT mixture by mass scaling) for a wide range of densities from ρD=0.5 to 673.518g/cm3 and temperatures from T=5000K up to the Fermi temperature TF for each density. Compared with results from the astrophysics opacity table (AOT) currently used in our hydrocodes, the FPOT of deuterium from our QMD calculations has shown a significant increase in opacity for strongly coupled and degenerate plasma conditions by a factor of 3-100 in the ICF-relevant photon-energy range. As conditions approach those of classical plasma, the opacity from the FPOT converges to the corresponding values of the AOT. By implementing the FPOT of deuterium and the DT mixture into our hydrocodes, we have performed radiation-hydrodynamics simulations for low-adiabat cryogenic DT implosions on the OMEGA laser and for direct-drive-ignition designs for the National Ignition Facility. The simulation results using the FPOT show that the target performance (in terms of neutron yield and energy gain) could vary from ˜10% up to a factor of ˜2 depending on the adiabat of the imploding DT capsule; the lower the adiabat, the more variation is seen in the prediction of target performance when compared to the AOT modeling.

  20. Preliminary study on a tetrahedral hohlraum with four half-cylindrical cavities for indirectly driven inertial confinement fusion

    Science.gov (United States)

    Jing, Longfei; Jiang, Shaoen; Kuang, Longyu; Zhang, Lu; Li, Liling; Lin, Zhiwei; Li, Hang; Zheng, Jianhua; Hu, Feng; Huang, Yunbao; Huang, Tianxuan; Ding, Yongkun

    2017-04-01

    A tetrahedral hohlraum with four half-cylindrical cavities (FHCH) is proposed to balance tradeoffs among the drive symmetry, coupling efficiency, and plasma filling of the hohlraum performance for indirectly driven inertial confinement fusion. The peak drive symmetry in the FHCH with a cavity-to-capsule ratio (CCR) of 2.2 is comparable to those in the spherical hohlraum of CCR  =  4.5 with six laser entrance holes (6LEHs-Sph.) ((Lan et al 2014 Phys. Plasmas 21 010704) and three-axis cylindrical hohlraum (6LEHs-Cyls.) of CCR  =  2.0 (Kuang et al 2016 Sci. Rep. 6 34636), and the filling time of plasma is close to the ones in the 6LEHs-Cyls. and the ignition target Rev5-CH of the national ignition campaign, and about half of that in the 6LEHs-Sph. In particular, the coupling efficiency is about 19% and 16% higher than those of the 6LEHs-Sph. and 6LEHs-Cyls., respectively. Besides, preliminary study indicates that the FHCH has a robust symmetry to uncertainties of power imbalance and pointing errors of laser beams. Furthermore, utilizing the FHCH, the feasibility of a tetrahedral indirect drive approach on the national ignition facility and hybrid indirect-direct drive approach with the laser arrangement designed specially for 6LEHs-Sph. or 6LEHs-Cyls., is also envisioned. Therefore, the proposed hohlraum configuration merits consideration as an alternative route to indirect-drive ignition.

  1. Bump evolution driven by the x-ray ablation Richtmyer-Meshkov effect in plastic inertial confinement fusion Ablators

    Directory of Open Access Journals (Sweden)

    Loomis Eric

    2013-11-01

    Full Text Available Growth of hydrodynamic instabilities at the interfaces of inertial confinement fusion capsules (ICF due to ablator and fuel non-uniformities are a primary concern for the ICF program. Recently, observed jetting and parasitic mix into the fuel were attributed to isolated defects on the outer surface of the capsule. Strategies for mitigation of these defects exist, however, they require reduced uncertainties in Equation of State (EOS models prior to invoking them. In light of this, we have begun a campaign to measure the growth of isolated defects (bumps due to x-ray ablation Richtmyer-Meshkov in plastic ablators to validate these models. Experiments used hohlraums with radiation temperatures near 70 eV driven by 15 beams from the Omega laser (Laboratory for Laser Energetics, University of Rochester, NY, which sent a ∼1.25Mbar shock into a planar CH target placed over one laser entrance hole. Targets consisted of 2-D arrays of quasi-gaussian bumps (10 microns tall, 34 microns FWHM deposited on the surface facing into the hohlraum. On-axis radiography with a saran (Cl Heα − 2.76keV backlighter was used to measure bump evolution prior to shock breakout. Shock speed measurements were also performed to determine target conditions. Simulations using the LEOS 5310 and SESAME 7592 models required the simulated laser power be turned down to 80 and 88%, respectively to match observed shock speeds. Both LEOS 5310 and SESAME 7592 simulations agreed with measured bump areal densities out to 6 ns where ablative RM oscillations were observed in previous laser-driven experiments, but did not occur in the x-ray driven case. The QEOS model, conversely, over predicted shock speeds and under predicted areal density in the bump.

  2. The Pulsed Fission-Fusion (PUFF) Concept for Deep Space Exploration and Terrestrial Power Generation

    Science.gov (United States)

    Adams, Robert; Cassibry, Jason; Schillo, Kevin

    2017-01-01

    This team is exploring a modified Z-pinch geometry as a propulsion system, imploding a liner of liquid lithium onto a pellet containing both fission and fusion fuel. The plasma resulting from the fission and fusion burn expands against a magnetic nozzle, for propulsion, or a magnetic confinement system, for terrestrial power generation. There is considerable synergy in the concept; the lithium acts as a temporary virtual cathode, and adds reaction mass for propulsion. Further, the lithium acts as a radiation shield against generated neutrons and gamma rays. Finally, the density profile of the column can be tailored using the lithium sheath. Recent theoretical and experimental developments (e.g. tailored density profile in the fuel injection, shear stabilization, and magnetic shear stabilization) have had great success in mitigating instabilities that have plagued previous fusion efforts. This paper will review the work in evaluating the pellet sizes and z-pinch conditions for optimal PuFF propulsion. Trades of pellet size and composition with z-pinch power levels and conditions for the tamper and lithium implosion are evaluated. Current models, both theoretical and computational, show that a z-pinch can ignite a small (1 cm radius) fission-fusion target with significant yield. Comparison is made between pure fission and boosted fission targets. Performance is shown for crewed spacecraft for high speed Mars round trip missions and near interstellar robotic missions. The PuFF concept also offers a solution for terrestrial power production. PuFF can, with recycling of the effluent, achieve near 100% burnup of fission fuel, providing a very attractive power source with minimal waste. The small size of PuFF relative to today's plants enables a more distributed power network and less exposure to natural or man-made disruptions.

  3. A high power, tunable free electron maser for fusion

    NARCIS (Netherlands)

    Urbanus, W. H.; Bratman, V. L.; Bongers, W. A.; Caplan, M.; Denisov, G. G.; van der Geer, C. A. J.; Manintveld, P.; Militsyn, B.; Oomens, A. A. M.; Poelman, A. J.; Plomp, J.; Pluygers, J.; Savilov, A. V.; Smeets, P. H. M.; Sterk, A. B.; Verhoeven, A. G. A.

    2001-01-01

    The Fusion-FEM experiment, a high-power, electrostatic free-electron maser being built at the FOM-Institute for Plasma Physics 'Rijnhuizen', is operated at various frequencies. So far, experiments were done without a depressed collector, and the pulse length was limited to 12 mus.

  4. Advanced Fusion Reactors for Space Propulsion and Power Systems

    Science.gov (United States)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  5. 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

  6. 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

  7. Inertial Confinement Fusion quarterly report April-June 1999, volume 9, number 3

    Energy Technology Data Exchange (ETDEWEB)

    MacGowan, B

    1999-06-01

    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.

  8. First-principles opacity table of warm dense deuterium for inertial-confinement-fusion applications.

    Science.gov (United States)

    Hu, S X; Collins, L A; Goncharov, V N; Boehly, T R; Epstein, R; McCrory, R L; Skupsky, S

    2014-09-01

    Accurate knowledge of the optical properties of a warm dense deuterium-tritium (DT) mixture is important for reliable design of inertial confinement fusion (ICF) implosions using radiation-hydrodynamics simulations. The opacity of a warm dense DT shell essentially determines how much radiation from hot coronal plasmas can be deposited in the DT fuel of an imploding capsule. Even for the simplest species of hydrogen, the accurate calculation of their opacities remains a challenge in the warm-dense matter regime because strong-coupling and quantum effects play an important role in such plasmas. With quantum-molecular-dynamics (QMD) simulations, we have derived a first-principles opacity table (FPOT) of deuterium (and the DT mixture by mass scaling) for a wide range of densities from ρ(D)=0.5 to 673.518g/cm(3) and temperatures from T=5000K up to the Fermi temperature T(F) for each density. Compared with results from the astrophysics opacity table (AOT) currently used in our hydrocodes, the FPOT of deuterium from our QMD calculations has shown a significant increase in opacity for strongly coupled and degenerate plasma conditions by a factor of 3-100 in the ICF-relevant photon-energy range. As conditions approach those of classical plasma, the opacity from the FPOT converges to the corresponding values of the AOT. By implementing the FPOT of deuterium and the DT mixture into our hydrocodes, we have performed radiation-hydrodynamics simulations for low-adiabat cryogenic DT implosions on the OMEGA laser and for direct-drive-ignition designs for the National Ignition Facility. The simulation results using the FPOT show that the target performance (in terms of neutron yield and energy gain) could vary from ∼10% up to a factor of ∼2 depending on the adiabat of the imploding DT capsule; the lower the adiabat, the more variation is seen in the prediction of target performance when compared to the AOT modeling.

  9. Fusion plasmas

    Science.gov (United States)

    Engelmann, F.

    1995-09-01

    In the following, a synthetic review of the information reported at the Conference will be given. No attempt is made to summarize specific contributions; rather the material contributed will be looked at from a few different angles. All areas of fusion plasma physics were represented: there were experimental results on magnetic confinement (tokamaks; stellarators; mirrors; reversed field pinches; field reversed configurations; Z-pinches, with emphasis on the dense Z-pinch; plasma focus, ect.) and on inertial confinement; related modelling and diagnolstics development; theory, as well as some technological activities (power generators; RF sources, etc.) and component (e.g. antennae) development for smaller fusion devices. In particular, fusion-related research in Latin America was exhaustively covered. In addition, large future projects in fusion research were summarized. (AIP)

  10. High-adiabat high-foot inertial confinement fusion implosion experiments on the national ignition facility.

    Science.gov (United States)

    Park, H-S; Hurricane, O A; Callahan, D A; Casey, D T; Dewald, E L; Dittrich, T R; Döppner, T; Hinkel, D E; Berzak Hopkins, L F; Le Pape, S; Ma, T; Patel, P K; Remington, B A; Robey, H F; Salmonson, J D; Kline, J L

    2014-02-07

    This Letter reports on a series of high-adiabat implosions of cryogenic layered deuterium-tritium (DT) capsules indirectly driven by a "high-foot" laser drive pulse at the National Ignition Facility. High-foot implosions have high ablation velocities and large density gradient scale lengths and are more resistant to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot. Indeed, the observed hot spot mix in these implosions was low and the measured neutron yields were typically 50% (or higher) of the yields predicted by simulation. On one high performing shot (N130812), 1.7 MJ of laser energy at a peak power of 350 TW was used to obtain a peak hohlraum radiation temperature of ∼300  eV. The resulting experimental neutron yield was (2.4±0.05)×10(15) DT, the fuel ρR was (0.86±0.063)  g/cm2, and the measured Tion was (4.2±0.16)  keV, corresponding to 8 kJ of fusion yield, with ∼1/3 of the yield caused by self-heating of the fuel by α particles emitted in the initial reactions. The generalized Lawson criteria, an ignition metric, was 0.43 and the neutron yield was ∼70% of the value predicted by simulations that include α-particle self-heating.

  11. A hardened gated x-ray imaging diagnostic for inertial confinement fusion experiments at the National Ignition Facility.

    Science.gov (United States)

    Glenn, S; Koch, J; Bradley, D K; Izumi, N; Bell, P; Holder, J; Stone, G; Prasad, R; MacKinnon, A; Springer, P; Landen, O L; Kyrala, G

    2010-10-01

    A gated x-ray detector is under development for use at the National Ignition Facility that is intended to provide plasma emission images in the presence of neutron yields up to 10(15) expected during inertial confinement fusion experiments with layered cryogenic targets. These images are expected to provide valuable time-resolved measurements of core and fuel symmetries. Additional capabilities of this instrument will include the ability to make spatially resolved electron temperature measurements. A description of this instrument and its operation is given with emphasis on features that differentiate it from previous designs.

  12. Direct asymmetry measurement of temperature and density spatial distributions in inertial confinement fusion plasmas from pinhole space-resolved spectra

    CERN Document Server

    Nagayama, T; Florido, R; Mayes, D; Tommasini, R; Koch, J A; Delettrez, J A; Regan, S P; Smalyuk, V A

    2014-01-01

    Two-dimensional space-resolved temperature and density images of an inertial confinement fusion (ICF) implosion core have been diagnosed for the first time. Argon-doped, direct-drive ICF experiments were performed at the Omega Laser Facility and a collection of two-dimensional space-resolved spectra were obtained from an array of gated, spectrally resolved pinhole images recorded by a multi-monochromatic x-ray imager. Detailed spectral analysis revealed asymmetries of the core not just in shape and size but in the temperature and density spatial distributions, thus characterizing the core with an unprecedented level of detail.

  13. An Overview of the Los Alamos Inertial Confinement Fusion and High-Energy-Density Physics Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Batha, Steven H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Physics Division

    2016-07-15

    The Los Alamos Inertial Confinement Fusion and Science Programs engage in a vigorous array of experiments, theory, and modeling. We use the three major High Energy Density facilities, NIF, Omega, and Z to perform experiments. These include opacity, radiation transport, hydrodynamics, ignition science, and burn experiments to aid the ICF and Science campaigns in reaching their stewardship goals. The ICF program operates two nuclear diagnostics at NIF, the neutron imaging system and the gamma reaction history instruments. Both systems are being expanded with significant capability enhancements.

  14. Propagation of nuclear data uncertainties for fusion power measurements

    Directory of Open Access Journals (Sweden)

    Sjöstrand Henrik

    2017-01-01

    Full Text Available Neutron measurements using neutron activation systems are an essential part of the diagnostic system at large fusion machines such as JET and ITER. Nuclear data is used to infer the neutron yield. Consequently, high-quality nuclear data is essential for the proper determination of the neutron yield and fusion power. However, uncertainties due to nuclear data are not fully taken into account in uncertainty analysis for neutron yield calibrations using activation foils. This paper investigates the neutron yield uncertainty due to nuclear data using the so-called Total Monte Carlo Method. The work is performed using a detailed MCNP model of the JET fusion machine; the uncertainties due to the cross-sections and angular distributions in JET structural materials, as well as the activation cross-sections in the activation foils, are analysed. It is found that a significant contribution to the neutron yield uncertainty can come from uncertainties in the nuclear data.

  15. Propagation of nuclear data uncertainties for fusion power measurements

    Science.gov (United States)

    Sjöstrand, Henrik; Conroy, Sean; Helgesson, Petter; Hernandez, Solis Augusto; Koning, Arjan; Pomp, Stephan; Rochman, Dimitri

    2017-09-01

    Neutron measurements using neutron activation systems are an essential part of the diagnostic system at large fusion machines such as JET and ITER. Nuclear data is used to infer the neutron yield. Consequently, high-quality nuclear data is essential for the proper determination of the neutron yield and fusion power. However, uncertainties due to nuclear data are not fully taken into account in uncertainty analysis for neutron yield calibrations using activation foils. This paper investigates the neutron yield uncertainty due to nuclear data using the so-called Total Monte Carlo Method. The work is performed using a detailed MCNP model of the JET fusion machine; the uncertainties due to the cross-sections and angular distributions in JET structural materials, as well as the activation cross-sections in the activation foils, are analysed. It is found that a significant contribution to the neutron yield uncertainty can come from uncertainties in the nuclear data.

  16. Neutronics and pumping power analyses on the Tokamak reactor for the fusion-biomass hybrid concept

    Energy Technology Data Exchange (ETDEWEB)

    Ibano, Kenzo, E-mail: kibano@gmail.com [Graduate School of Energy Science, Kyoto University, Uji (Japan); Kasada, Ryuta [Graduate School of Energy Science, Kyoto University, Uji (Japan); Yamamoto, Yasushi [Faculty of Engineering Science, Kansai University, Suita, Osaka (Japan); Konishi, Satoshi [Graduate School of Energy Science, Kyoto University, Uji (Japan)

    2013-11-15

    Highlights: • MCNP analyses on a Tokamak with LiPb-cooled components shows concentrations of nuclear heating at the in-board region in addition to the out-board region. • Required pumping power of LiPb coolants for the nuclear heating exponentially increases as fusion power increases. • Pumping power analysis for the divertor also indicates the increasing pumping power as the fusion power increases. -- Abstract: The authors aim to develop a fusion-biomass combined plant concept with a small power fusion reactor. A concern for the small power reactor is the coolant pumping power which may significantly decreases the apparent energy outcome. Thus pressure loss and corresponding pumping power were studied for a designed Tokamak reactor: GNOME. First, 3-D Monte-Carlo Neutron transport analysis for the reactor model with dual-coolant blankets was taken in order to simulate the tritium breeding ability and the distribution of nuclear heat. Considering calculated concentration of nuclear heat on the in-board blankets, pressure loss of the liquid LiPb at coolant pipes due to MHD and friction forces was analyzed as a function of fusion power. It was found that as the fusion power increases, the pressure loss and corresponding pumping power exponentially increase. Consequently, the proportion of the pumping power to the fusion power increases as the fusion power increases. In case of ∼360 MW fusion power operation, pumping power required for in-board cooling pipes was estimated as ∼1% of the fusion power.

  17. Radioactive waste management and disposal scenario for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

    1997-10-01

    The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

  18. Personnel Safety for Future Magnetic Fusion Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee Cadwallader

    2009-07-01

    The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

  19. Kinetic transport in a magnetically confined and flux-constrained fusion plasma; Transport cinetique dans un plasma de fusion magnetique a flux force

    Energy Technology Data Exchange (ETDEWEB)

    Darmet, G

    2007-11-15

    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.)

  20. LIFE: a sustainable solution for developing safe, clean fusion power.

    Science.gov (United States)

    Reyes, Susana; Dunne, Mike; Kramer, Kevin; Anklam, Tom; Havstad, Mark; Mazuecos, Antonio Lafuente; Miles, Robin; Martinez-Frias, Joel; Deri, Bob

    2013-06-01

    The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in California is currently in operation with the goal to demonstrate fusion energy gain for the first time in the laboratory-also referred to as "ignition." Based on these demonstration experiments, the Laser Inertial Fusion Energy (LIFE) power plant is being designed at LLNL in partnership with other institutions with the goal to deliver baseload electricity from safe, secure, sustainable fusion power in a time scale that is consistent with the energy market needs. For this purpose, the LIFE design takes advantage of recent advances in diode-pumped, solid-state laser technology and adopts the paradigm of Line Replaceable Units used on the NIF to provide high levels of availability and maintainability and mitigate the need for advanced materials development. The LIFE market entry plant will demonstrate the feasibility of a closed fusion fuel cycle, including tritium breeding, extraction, processing, refueling, accountability, and safety, in a steady-state power-producing device. While many fusion plant designs require large quantities of tritium for startup and operations, a range of design choices made for the LIFE fuel cycle act to reduce the in-process tritium inventory. This paper presents an overview of the delivery plan and the preconceptual design of the LIFE facility with emphasis on the key safety design principles being adopted. In order to illustrate the favorable safety characteristics of the LIFE design, some initial accident analysis results are presented that indicate potential for a more attractive licensing regime than that of current fission reactors.

  1. The potential of imposed magnetic fields for enhancing ignition probability and fusion energy yield in indirect-drive inertial confinement fusion

    Science.gov (United States)

    Perkins, L. J.; Ho, D. D.-M.; Logan, B. G.; Zimmerman, G. B.; Rhodes, M. A.; Strozzi, D. J.; Blackfield, D. T.; Hawkins, S. A.

    2017-06-01

    We examine the potential that imposed magnetic fields of tens of Tesla that increase to greater than 10 kT (100 MGauss) under implosion compression may relax the conditions required for ignition and propagating burn in indirect-drive inertial confinement fusion (ICF) targets. This may allow the attainment of ignition, or at least significant fusion energy yields, in presently performing ICF targets on the National Ignition Facility (NIF) that today are sub-marginal for thermonuclear burn through adverse hydrodynamic conditions at stagnation [Doeppner et al., Phys. Rev. Lett. 115, 055001 (2015)]. Results of detailed two-dimensional radiation-hydrodynamic-burn simulations applied to NIF capsule implosions with low-mode shape perturbations and residual kinetic energy loss indicate that such compressed fields may increase the probability for ignition through range reduction of fusion alpha particles, suppression of electron heat conduction, and potential stabilization of higher-mode Rayleigh-Taylor instabilities. Optimum initial applied fields are found to be around 50 T. Given that the full plasma structure at capsule stagnation may be governed by three-dimensional resistive magneto-hydrodynamics, the formation of closed magnetic field lines might further augment ignition prospects. Experiments are now required to further assess the potential of applied magnetic fields to ICF ignition and burn on NIF.

  2. Conceptual design of the Fast-Liner Reactor (FLR) for fusion power

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-02-01

    The generation of fusion power from the Fast-Liner Reactor (FLR) concept envisages the implosion of a thin (3-mm) metallic cylinder (0.2-m radius by 0.2-m length) onto a preinjected plasma. This plasma would be heated to thermonuclear temperatures by adiabatic compression, pressure confinement would be provided by the liner inertia, and thermal insulation of the wall-confined plasma would be established by an embedded azimuthal magnetic field. A 2- to 3-mu s burn would follow the approx. 10/sup 4/ m/s radial implosion and would result in a thermonuclear yield equal to 10 to 15 times the energy initially invested into the liner kinetic energy. For implosions occurring once every 10 s a gross thermal power of 430 MWt would be generated. The results of a comprehensive systems study of both physics and technology (economics) optima are presented. Despite unresolved problems associated with both the physics and technology of the FLR, a conceptual power plant design is presented.

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

    Science.gov (United States)

    Glowacki, B. A.; Nuttall, W. J.

    2008-02-01

    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 H2 (rather than electricity) is presented. Exploitation of produced hydrogen as a coolant and as a fuel is proposed in conjunction with MgB2 conductors for the tokomak magnets windings, and electrotechnical devices for Fusion Island's infrastructure. The benefits of using MgB2 over the Nb-based conductors during construction, operation and decommissioning of the Fusion Island Reactor are presented. The comparison of Nb3Sn strands for ITER fusion magnet with newly developed high field composite MgB2 PIT conductors has shown that at 14 Tesla MgB2 possesses better properties than any of the Nb3Sn conductors produced. In this paper the potential of MgB2 conductors is examined for tokamaks of both the conventional ITER type and a Spherical Tokamak geometry. In each case MgB2 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 MgB2 conductors for Fusion Island are also considered.

  4. A new simpler way to obtain high fusion power gain in tandem mirrors

    Science.gov (United States)

    Fowler, T. K.; Moir, R. W.; Simonen, T. C.

    2017-05-01

    From the earliest days of fusion research, Richard F. Post and other advocates of magnetic mirror confinement recognized that mirrors favor high ion temperatures where nuclear reaction rates begin to peak for all fusion fuels. In this paper we review why high ion temperatures are favored, using Post’s axisymmetric Kinetically Stabilized Tandem Mirror as the example; and we offer a new idea that appears to greatly improve reactor prospects at high ion temperatures. The idea is, first, to take advantage of recent advances in superconducting magnet technology to minimize the size and cost of End Plugs; and secondly, to utilize parallel advances in gyrotrons that would enable intense electron cyclotron heating (ECH) in these high field End Plugs. The yin-yang magnets and thermal barriers that complicated earlier tandem mirror designs are not required. We find that, concerning end losses, intense ECH in symmetric End Plugs could increase the fusion power gain Q, for both DT and Catalyzed DD fuel cycles, to levels competitive with steady-state tokamaks burning DT fuel. Radial losses remain an issue that will ultimately determine reactor viability.

  5. Development of an inertial confinement fusion platform to study charged-particle-producing nuclear reactions relevant to nuclear astrophysics

    Science.gov (United States)

    Gatu Johnson, M.; Zylstra, A. B.; Bacher, A.; Brune, C. R.; Casey, D. T.; Forrest, C.; Herrmann, H. W.; Hohenberger, M.; Sayre, D. B.; Bionta, R. M.; Bourgade, J.-L.; Caggiano, J. A.; Cerjan, C.; Craxton, R. S.; Dearborn, D.; Farrell, M.; Frenje, J. A.; Garcia, E. M.; Glebov, V. Yu.; Hale, G.; Hartouni, E. P.; Hatarik, R.; Hohensee, M.; Holunga, D. M.; Hoppe, M.; Janezic, R.; Khan, S. F.; Kilkenny, J. D.; Kim, Y. H.; Knauer, J. P.; Kohut, T. R.; Lahmann, B.; Landoas, O.; Li, C. K.; Marshall, F. J.; Masse, L.; McEvoy, A.; McKenty, P.; McNabb, D. P.; Nikroo, A.; Parham, T. G.; Paris, M.; Petrasso, R. D.; Pino, J.; Radha, P. B.; Remington, B.; Rinderknecht, H. G.; Robey, H.; Rosenberg, M. J.; Rosse, B.; Rubery, M.; Sangster, T. C.; Sanchez, J.; Schmitt, M.; Schoff, M.; Séguin, F. H.; Seka, W.; Sio, H.; Stoeckl, C.; Tipton, R. E.

    2017-04-01

    This paper describes the development of a platform to study astrophysically relevant nuclear reactions using inertial-confinement fusion implosions on the OMEGA and National Ignition Facility laser facilities, with a particular focus on optimizing the implosions to study charged-particle-producing reactions. Primary requirements on the platform are high yield, for high statistics in the fusion product measurements, combined with low areal density, to allow the charged fusion products to escape. This is optimally achieved with direct-drive exploding pusher implosions using thin-glass-shell capsules. Mitigation strategies to eliminate a possible target sheath potential which would accelerate the emitted ions are discussed. The potential impact of kinetic effects on the implosions is also considered. The platform is initially employed to study the complementary T(t,2n)α, T(3He,np)α and 3He(3He,2p)α reactions. Proof-of-principle results from the first experiments demonstrating the ability to accurately measure the energy and yields of charged particles are presented. Lessons learned from these experiments will be used in studies of other reactions. The goals are to explore thermonuclear reaction rates and fundamental nuclear physics in stellar-like plasma environments, and to push this new frontier of nuclear astrophysics into unique regimes not reachable through existing platforms, with thermal ion velocity distributions, plasma screening, and low reactant energies.

  6. Plasma confinement

    CERN Document Server

    Hazeltine, R D

    2003-01-01

    Detailed and authoritative, this volume examines the essential physics underlying international research in magnetic confinement fusion. It offers readable, thorough accounts of the fundamental concepts behind methods of confining plasma at or near thermonuclear conditions. Designed for a one- or two-semester graduate-level course in plasma physics, it also represents a valuable reference for professional physicists in controlled fusion and related disciplines.

  7. Experimental investigation of the confinement of d(3He,p)α and d(d,p)t fusion reaction products in JET

    DEFF Research Database (Denmark)

    Bonheure, Georges; Hult, M.; Gonzalez de Orduna, R.

    2012-01-01

    In ITER, magnetic fusion will explore the burning plasma regime. Because such burning plasma is sustained by its own fusion reactions, alpha particles need to be confined (Hazeltine 2010 Fusion Eng. Des. 7–9 85). New experiments using d(3He,p)α and d(d,p)t fusion reaction products were performed...... in JET. Fusion product loss was measured from MHD-quiescent plasmas with a charged particle activation probe installed at a position opposite to the magnetic field ion gradient drift (see figure 1)—1.77 m above mid-plane—in the ceiling of JET tokamak. This new kind of escaping ion detector (Bonheure et...... al 2008 Fusion Sci. Technol. 53 806) provides for absolutely calibrated measurements. Both the mechanism and the magnitude of the loss are dealt with by this research. Careful analysis shows measured loss is in quantitative agreement with predictions from the classical orbit loss model. However...

  8. Nuclear Fusion

    National Research Council Canada - National Science Library

    Ghoranneviss, Mahmood; Parashar, S. K. S; Aslan, Necdet; Aslaninejad, Morteza; Salar Elahi, A

    2014-01-01

    ... in both inertial and magnetic confinement fusion, with attendees from major fusion energy research centers worldwide. It is one of the most important issues in this field. Nuclear fusion continues t...

  9. Neutron penumbral imaging simulation and reconstruction for Inertial Confinement Fusion Experiments

    CERN Document Server

    Wang, Xian-You; Tang, Yun-qing; Tang, Zhi-Cheng; Xiao, Hong; Xu, Ming

    2012-01-01

    Neutron penumbral imaging technique has been successfully used as the diagnosis method in Inertial Con?ned Fusion. To help the design of the imaging systems in the future in CHINA. We construct the Monte carlo imaging system by Geant4. Use the point spread function from the simulation and decode algorithm (Lucy-Rechardson algorithm) we got the recovery image.

  10. Generating energetic electrons through staged acceleration in the two-plasmon-decay instability in inertial confinement fusion.

    Science.gov (United States)

    Yan, R; Ren, C; Li, J; Maximov, A V; Mori, W B; Sheng, Z-M; Tsung, F S

    2012-04-27

    A new hot-electron generation mechanism in two-plasmon-decay instabilities is described based on a series of 2D, long-term (~10 ps) particle-in-cell and fluid simulations under parameters relevant to inertial confinement fusion. The simulations show that significant laser absorption and hot-electron generation occur in the nonlinear stage. The hot electrons are stage accelerated from the low-density region to the high-density region. New modes with small phase velocities develop in the low-density region in the nonlinear stage and form the first stage for electron acceleration. Electron-ion collisions are shown to significantly reduce the efficiency of this acceleration mechanism.

  11. First Measurements of Fuel-Ablator Interface Instability Growth in Inertial Confinement Fusion Implosions on the National Ignition Facility.

    Science.gov (United States)

    Weber, C R; Döppner, T; Casey, D T; Bunn, T L; Carlson, L C; Dylla-Spears, R J; Kozioziemski, B J; MacPhee, A G; Nikroo, A; Robey, H F; Sater, J D; Smalyuk, V A

    2016-08-12

    Direct measurements of hydrodynamic instability growth at the fuel-ablator interface in inertial confinement fusion (ICF) implosions are reported for the first time. These experiments investigate one of the degradation mechanisms behind the lower-than-expected performance of early ICF implosions on the National Ignition Facility. Face-on x-ray radiography is used to measure instability growth occurring between the deuterium-tritium fuel and the plastic ablator from well-characterized perturbations. This growth starts in two ways through separate experiments-either from a preimposed interface modulation or from ablation front feedthrough. These experiments are consistent with analytic modeling and radiation-hydrodynamic simulations, which say that a moderately unstable Atwood number and convergence effects are causing in-flight perturbation growth at the interface. The analysis suggests that feedthrough from outersurface perturbations dominates the interface perturbation growth at mode 60.

  12. Two-Plasmon Decay Mitigation in Direct-Drive Inertial-Confinement-Fusion Experiments Using Multilayer Targets.

    Science.gov (United States)

    Follett, R K; Delettrez, J A; Edgell, D H; Goncharov, V N; Henchen, R J; Katz, J; Michel, D T; Myatt, J F; Shaw, J; Solodov, A A; Stoeckl, C; Yaakobi, B; Froula, D H

    2016-04-15

    Multilayer direct-drive inertial-confinement-fusion targets are shown to significantly reduce two-plasmon decay (TPD) driven hot-electron production while maintaining high hydrodynamic efficiency. Implosion experiments on the OMEGA laser used targets with silicon layered between an inner beryllium and outer silicon-doped plastic ablator. A factor-of-5 reduction in hot-electron generation (>50  keV) was observed in the multilayer targets relative to pure CH targets. Three-dimensional simulations of the TPD-driven hot-electron production using a laser-plasma interaction code (lpse) that includes nonlinear and kinetic effects show good agreement with the measurements. The simulations suggest that the reduction in hot-electron production observed in the multilayer targets is primarily caused by increased electron-ion collisional damping.

  13. Observation of Interspecies Ion Separation in Inertial-Confinement-Fusion Implosions via Imaging X-Ray Spectroscopy

    CERN Document Server

    Hsu, S C; Hakel, P; Vold, E L; Schmitt, M J; Hoffman, N M; Rauenzahn, R M; Kagan, G; Tang, X -Z; Mancini, R C; Kim, Y; Herrmann, H W

    2016-01-01

    We report direct experimental evidence of interspecies ion separation in direct-drive, inertial-confinement-fusion experiments on the OMEGA laser facility. These experiments, which used plastic capsules with D$_2$/Ar gas fill (1% Ar by atom), were designed specifically to reveal interspecies ion separation by exploiting the predicted, strong ion thermo-diffusion between ion species of large mass and charge difference. Via detailed analyses of imaging x-ray-spectroscopy data, we extract Ar-atom-fraction radial profiles at different times, and observe both enhancement and depletion compared to the initial 1%-Ar gas fill. The experimental results are interpreted with radiation-hydrodynamic simulations that include recently implemented, first-principles models of interspecies ion diffusion. The experimentally inferred Ar-atom-fraction profiles agree reasonably, but not exactly, with calculated profiles associated with the incoming and rebounding first shock.

  14. On the change in Inertial Confinement Fusion Implosions upon using an ab initio multiphase DT equation of state

    CERN Document Server

    Caillabet, Laurent; Salin, Gwenaël; Mazevet, Stéphane; Loubeyre, Paul

    2011-01-01

    Improving the description of the equation of state (EoS) of deuterium-tritium (DT) has recently been shown to change significantly the gain of an Inertial Confinement Fusion (ICF) target (Hu et al., PRL 104, 235003 (2010)). We use here an advanced multi-phase equation of state (EoS), based on ab initio calculations, to perform a full optimization of the laser pulse shape with hydrodynamic simulations starting from 19 K in DT ice. The thermonuclear gain is shown to be a robust estimate over possible uncertainties of the EoS. Two different target designs are discussed, for shock ignition and self-ignition. In the first case, the areal density and thermonuclear energy can be recovered by slightly increasing the laser energy. In the second case, a lower in-flight adiabat is needed, leading to a significant delay (3ns) in the shock timing of the implosion.

  15. Uniformity of spherical shock wave dynamically stabilized by two successive laser profiles in direct-drive inertial confinement fusion implosions

    Energy Technology Data Exchange (ETDEWEB)

    Temporal, M., E-mail: mauro.temporal@hotmail.com [Centre de Mathématiques et de Leurs Applications, ENS Cachan and CNRS, 61 Av. du President Wilson, F-94235 Cachan Cedex (France); Canaud, B. [CEA, DIF, F-91297 Arpajon Cedex (France); Garbett, W. J. [AWE plc, Aldermaston, Reading, Berkshire RG7 4PR (United Kingdom); Ramis, R. [ETSI Aeronáutica y del Espacio, Universidad Politécnica de Madrid, 28040 Madrid (Spain)

    2015-10-15

    The implosion uniformity of a directly driven spherical inertial confinement fusion capsule is considered within the context of the Laser Mégajoule configuration. Two-dimensional (2D) hydrodynamic simulations have been performed assuming irradiation with two laser beam cones located at 49° and 131° with respect to the axis of symmetry. The laser energy deposition causes an inward shock wave whose surface is tracked in time, providing the time evolution of its non-uniformity. The illumination model has been used to optimize the laser intensity profiles used as input in the 2D hydro-calculations. It is found that a single stationary laser profile does not maintain a uniform shock front over time. To overcome this drawback, it is proposed to use two laser profiles acting successively in time, in order to dynamically stabilize the non-uniformity of the shock front.

  16. Comment on 'Evidence for Stratification of Deuterium-Tritium Fuel in Inertial Confinement Fusion Implosions'

    CERN Document Server

    Zheng, Hua

    2013-01-01

    Recent implosion experiments performed at the OMEGA laser facility reported by Casey et al.[1], displayed an anomalously low dd proton yield and a high tt neutron yield as compared to dt fusion reactions, explained as a stratification of the fuel in the implosion core. We suggest that in the com- pression stage the fuel is out of equilibrium. Ions are inward accelerated to a velocity v0 independent on the particle type. Yield ratios are simply given by the ratios of fusion cross-sections obtained at the same velocity. A 'Hubble' type model gives also a reasonable description of the data. These considerations might be relevant for implosion experiments at the National Ignition Facility as well.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hoppe, M. [ed.

    1997-02-01

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

  18. Inertial Confinement Fusion Target Component Fabrication and Technology Development report. Annual report, October 1, 1992--September 30, 1993

    Energy Technology Data Exchange (ETDEWEB)

    Steinman, D. [ed.

    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.

  19. Inertial confinement fusion quarterly report, October--December 1992. Volume 3, No. 1

    Energy Technology Data Exchange (ETDEWEB)

    Dixit, S.N. [ed.

    1992-12-31

    This report contains papers on the following topics: The Beamlet Front End: Prototype of a new pulse generation system;imaging biological objects with x-ray lasers; coherent XUV generation via high-order harmonic generation in rare gases; theory of high-order harmonic generation; two-dimensional computer simulations of ultra- intense, short-pulse laser-plasma interactions; neutron detectors for measuring the fusion burn history of ICF targets; the recirculator; and lasnex evolves to exploit computer industry advances.

  20. Proceedings of the US-Japan workshop and the satellite meeting of ITC-9 on physics of high beta plasma confinement in innovative fusion system

    Energy Technology Data Exchange (ETDEWEB)

    Goto, Seiichi; Yoshimura, Satoru [eds.

    1999-04-01

    The US-Japan Workshop on Physics of High Beta Plasma Confinement in Innovative Fusion System was held jointly with the Satellite Meeting of ITC-9 at National Institute for Fusion Science (NIFS), Toki-city during December 14-15, 1998. This proceedings book includes the papers of the talks given at the workshop. These include: Theoretical analysis on the stability of field reversed configuration (FRC) plasmas; Theory and Modeling of high {beta} plasmas; Recent progressive experiments in high {beta} systems; Formation of high {beta} plasmas using merging phenomenon; Theory and Modeling of a FRC Fusion Reactor. The 15 papers are indexed individually. (J.P.N.)

  1. Considerations of the social impact of fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Gastil, R.D.; Markus, H.S.

    1976-09-01

    It is concluded that the direct effects of an ideal form of fusion technologies would be socially more desirable than those of the alternatives. This is particularly true of the second generation fusion power plant. However, given our technological inputs, this was a trivial result. Less trivial was consideration of the negative effects that might accrue through the availability of potentially unlimited supplies of low cost energy. It is concluded that while there may be reasonable humanist argument both for and against such abundance, in a democratic society control of energy development for its own sake is likely to be unacceptable. However, if the indirect effects of pollution, despoilment, and resource depletion through ever expanding energy use become sufficiently disturbing to the well-being of the majority, unlimited energy may come to be seen as undesirable by the society. To this extent successful research and development for unlimited sources such as the fusion or mixed solar alternatives might be judged from some point far in the future to have been a mistake. This could occur even though advances in the technology of pollution control and resource use greatly reduce the pollution and hazard accompanying a much higher rate of energy utilization.

  2. Demonstration of Ion Kinetic Effects in Inertial Confinement Fusion Implosions and Investigation of Magnetic Reconnection Using Laser-Produced Plasmas

    Science.gov (United States)

    Rosenberg, M. J.

    2016-10-01

    Shock-driven laser inertial confinement fusion (ICF) implosions have demonstrated the presence of ion kinetic effects in ICF implosions and also have been used as a proton source to probe the strongly driven reconnection of MG magnetic fields in laser-generated plasmas. Ion kinetic effects arise during the shock-convergence phase of ICF implosions when the mean free path for ion-ion collisions (λii) approaches the size of the hot-fuel region (Rfuel) and may impact hot-spot formation and the possibility of ignition. To isolate and study ion kinetic effects, the ratio of N - K =λii /Rfuel was varied in D3He-filled, shock-driven implosions at the Omega Laser Facility and the National Ignition Facility, from hydrodynamic-like conditions (NK 0.01) to strongly kinetic conditions (NK 10). A strong trend of decreasing fusion yields relative to the predictions of hydrodynamic models is observed as NK increases from 0.1 to 10. Hydrodynamics simulations that include basic models of the kinetic effects that are likely to be present in these experiments-namely, ion diffusion and Knudsen-layer reduction of the fusion reactivity-are better able to capture the experimental results. This type of implosion has also been used as a source of monoenergetic 15-MeV protons to image magnetic fields driven to reconnect in laser-produced plasmas at conditions similar to those encountered at the Earth's magnetopause. These experiments demonstrate that for both symmetric and asymmetric magnetic-reconnection configurations, when plasma flows are much stronger than the nominal Alfvén speed, the rate of magnetic-flux annihilation is determined by the flow velocity and is largely insensitive to initial plasma conditions. This work was supported by the Department of Energy Grant Number DENA0001857.

  3. Direct measurement of energetic electrons coupling to an imploding low-adiabat inertial confinement fusion capsule.

    Science.gov (United States)

    Döppner, T; Thomas, C A; Divol, L; Dewald, E L; Celliers, P M; Bradley, D K; Callahan, D A; Dixit, S N; Harte, J A; Glenn, S M; Haan, S W; Izumi, N; Kyrala, G A; LaCaille, G; Kline, J K; Kruer, W L; Ma, T; MacKinnon, A J; McNaney, J M; Meezan, N B; Robey, H F; Salmonson, J D; Suter, L J; Zimmerman, G B; Edwards, M J; MacGowan, B J; Kilkenny, J D; Lindl, J D; Van Wonterghem, B M; Atherton, L J; Moses, E I; Glenzer, S H; Landen, O L

    2012-03-30

    We have imaged hard x-ray (>100 keV) bremsstrahlung emission from energetic electrons slowing in a plastic ablator shell during indirectly driven implosions at the National Ignition Facility. We measure 570 J in electrons with E>100 keV impinging on the fusion capsule under ignition drive conditions. This translates into an acceptable increase in the adiabat α, defined as the ratio of total deuterium-tritium fuel pressure to Fermi pressure, of 3.5%. The hard x-ray observables are consistent with detailed radiative-hydrodynamics simulations, including the sourcing and transport of these high energy electrons.

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

    Energy Technology Data Exchange (ETDEWEB)

    Basin, A.A. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Vatulin, V.V. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Vakhlamova, L.L. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Vinokurov, P.A. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Dement`ev, Yu.A. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Eliseev, G.M. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Ermolovich, V.F. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Morenko, L.Z. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Morenko, A.I. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Remizov, G.N. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Romanov, Yu.A. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Ryabikina, N.A. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Skrypnik, S.I. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Skidan, G.I. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Tikhomirov, B.P. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.; Shagaliev, R.M. [Russian Federal Nucl. Centre, Arzamas (Russian Federation). Inst. of Exp. Phys.

    1996-11-01

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

  5. Fusion power in a future low carbon global electricity system

    DEFF Research Database (Denmark)

    Cabal, H.; Lechón, Y.; Bustreo, C.

    2017-01-01

    of fusion technologies in the global electricity system in the long term. Results show that fusion technologies penetration is higher in scenarios with stricter CO2 emissions reduction targets. In addition, investment costs and discount rates of fusion technologies are key factors for fusion implementation...

  6. Core conditions for alpha heating attained in direct-drive inertial confinement fusion.

    Science.gov (United States)

    Bose, A; Woo, K M; Betti, R; Campbell, E M; Mangino, D; Christopherson, A R; McCrory, R L; Nora, R; Regan, S P; Goncharov, V N; Sangster, T C; Forrest, C J; Frenje, J; Gatu Johnson, M; Glebov, V Yu; Knauer, J P; Marshall, F J; Stoeckl, C; Theobald, W

    2016-07-01

    It is shown that direct-drive implosions on the OMEGA laser have achieved core conditions that would lead to significant alpha heating at incident energies available on the National Ignition Facility (NIF) scale. The extrapolation of the experimental results from OMEGA to NIF energy assumes only that the implosion hydrodynamic efficiency is unchanged at higher energies. This approach is independent of the uncertainties in the physical mechanism that degrade implosions on OMEGA, and relies solely on a volumetric scaling of the experimentally observed core conditions. It is estimated that the current best-performing OMEGA implosion [Regan et al., Phys. Rev. Lett. 117, 025001 (2016)10.1103/PhysRevLett.117.025001] extrapolated to a 1.9 MJ laser driver with the same illumination configuration and laser-target coupling would produce 125 kJ of fusion energy with similar levels of alpha heating observed in current highest performing indirect-drive NIF implosions.

  7. Strong screening by lattice confinement and resultant fusion reaction rates in fcc metals

    Science.gov (United States)

    Prados-Estévez, F. M.; Subashiev, A. V.; Nee, H. H.

    2017-09-01

    The effects of electronic screening on the cross sections and reactivities for the nuclear reactions between light nuclei in Pd and Ni is studied. We consider the applicability of the theory of thermonuclear burning in stars to the D-D nuclear reaction in metals. The screening model based on the mean field potential of the electron cloud in the metal plasma is used. We discuss the specifics of the screening for the H (D) atoms embedded in vacancies and divacancies. High concentration of hydrogen isotopes segregated to monovacancies and divacancies in face-centered cubic (fcc) metals such as Ni and Pd with densities of ∼ 6 ×1023atom /cm3 , makes the hydrogen cluster a favorable active site for the fusion reaction. Still the observation of a nuclear reaction requires an accumulation of energy in D nuclei of at least several eV, which is far above what can be achieved in the thermal heating experiments.

  8. Controlled thermonuclear fusion; La fusion thermonucleaire controlee

    Energy Technology Data Exchange (ETDEWEB)

    Bobin, Jean-Louis

    2011-10-27

    This book presents, first, the basic nuclear physics and plasma physics principles at the origin of the researches on nuclear fusion which started in the 1950's. Then the magnetic and inertial confinement principles are described with their corresponding facilities: the tokamaks and the laser-induced inertial confinement fusion devices. After a brief outline of some exotic processes, the book describes some projects of thermonuclear power plants capable to supply electricity by the end of the 21. century. (J.S.)

  9. A novel design for scintillator-based neutron and gamma imaging in inertial confinement fusion

    Science.gov (United States)

    Geppert-Kleinrath, Verena; Cutler, Theresa; Danly, Chris; Madden, Amanda; Merrill, Frank; Tybo, Josh; Volegov, Petr; Wilde, Carl

    2017-10-01

    The LANL Advanced Imaging team has been providing reliable 2D neutron imaging of the burning fusion fuel at NIF for years, revealing possible multi-dimensional asymmetries in the fuel shape, and therefore calling for additional views. Adding a passive imaging system using image plate techniques along a new polar line of sight has recently demonstrated the merit of 3D neutron image reconstruction. Now, the team is in the process of designing a new active neutron imaging system for an additional equatorial view. The design will include a gamma imaging system as well, to allow for the imaging of carbon in the ablator of the NIF fuel capsules, constraining the burning fuel shape even further. The selection of ideal scintillator materials for a position-sensitive detector system is the key component for the new design. A comprehensive study of advanced scintillators has been carried out at the Los Alamos Neutron Science Center and the OMEGA Laser Facility in Rochester, NY. Neutron radiography using a fast-gated CCD camera system delivers measurements of resolution, light output and noise characteristics. The measured performance parameters inform the novel design, for which we conclude the feasibility of monolithic scintillators over pixelated counterparts.

  10. Foam Au driven by 4ω-2ω ignition laser pulse for inertial confinement fusion

    Science.gov (United States)

    Lan, Ke; Song, Peng

    2017-05-01

    Green light (2ω) has the potential to drive ignition target for laser fusion with significantly more energy than blue light (3ω) and a relatively higher damage threshold for the optic components in the final optic assembly, but it has issues of a relatively low laser to x-ray conversion efficiency and a hard x-ray spectrum as compared to 3ω. In this paper, we propose to drive a foam hohlraum wall with an ignition laser pulse by taking a 4ω laser at the pre-pulse and a 2ω laser at the main-pulse, called as 4ω-2ω ignition pulse. This novel design has the following advantages: (1) benefiting from 2ω of its relatively high energy output and low damage threshold during main-pulse; (2) benefiting from foam in its relatively high laser to x-ray conversion efficiency and relatively low M-band fraction in re-emission; (3) benefiting from 4ω of its low LPI and low M-band fraction during pre-pulse. From our one-dimensional simulations with the Au material, the laser to x-ray conversion in a foam driven by 4ω-2ω pulse has an increase of 28% as compared to a solid target driven by 3ω with the same pulse shape. The relatively thin optical depth of foam is one of the main reasons for the increase of laser to x-ray conversion efficiency inside a foam target.

  11. Testing nonlocal models of electron thermal conduction for magnetic and inertial confinement fusion applications

    Science.gov (United States)

    Brodrick, J. P.; Kingham, R. J.; Marinak, M. M.; Patel, M. V.; Chankin, A. V.; Omotani, J. T.; Umansky, M. V.; Del Sorbo, D.; Dudson, B.; Parker, J. T.; Kerbel, G. D.; Sherlock, M.; Ridgers, C. P.

    2017-09-01

    Three models for nonlocal electron thermal transport are here compared against Vlasov-Fokker-Planck (VFP) codes to assess their accuracy in situations relevant to both inertial fusion hohlraums and tokamak scrape-off layers. The models tested are (i) a moment-based approach using an eigenvector integral closure (EIC) originally developed by Ji, Held, and Sovinec [Phys. Plasmas 16, 022312 (2009)]; (ii) the non-Fourier Landau-fluid (NFLF) model of Dimits, Joseph, and Umansky [Phys. Plasmas 21, 055907 (2014)]; and (iii) Schurtz, Nicolaï, and Busquet's [Phys. Plasmas 7, 4238 (2000)] multigroup diffusion model (SNB). We find that while the EIC and NFLF models accurately predict the damping rate of a small-amplitude temperature perturbation (within 10% at moderate collisionalities), they overestimate the peak heat flow by as much as 35% and do not predict preheat in the more relevant case where there is a large temperature difference. The SNB model, however, agrees better with VFP results for the latter problem if care is taken with the definition of the mean free path. Additionally, we present for the first time a comparison of the SNB model against a VFP code for a hohlraum-relevant problem with inhomogeneous ionisation and show that the model overestimates the heat flow in the helium gas-fill by a factor of ˜2 despite predicting the peak heat flux to within 16%.

  12. Damage of actively cooled plasma facing components of magnetic confinement controlled fusion machines

    Science.gov (United States)

    Chevet, G.; Schlosser, J.; Martin, E.; Herb, V.; Camus, G.

    2009-03-01

    Plasma facing components (PFCs) of magnetic fusion machines have high manufactured residual stresses and have to withstand important stress ranges during operation. These actively cooled PFCs have a carbon fibre composite (CFC) armour and a copper alloy heat sink. Cracks mainly appear in the CFC near the composite/copper interface. In order to analyse damage mechanisms, it is important to well simulate the damage mechanisms both of the CFC and the CFC/Cu interface. This study focuses on the mechanical behaviour of the N11 material for which the scalar ONERA damage model was used. The damage parameters of this model were identified by similarity to a neighbour material, which was extensively analysed, according to the few characterization test results available for the N11. The finite elements calculations predict a high level of damage of the CFC at the interface zone explaining the encountered difficulties in the PFCs fabrication. These results suggest that the damage state of the CFC cells is correlated with a conductivity decrease to explain the temperature increase of the armour surface under fatigue heat load.

  13. Developing maintainability for fusion power systems. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Zahn, H.S.; Mantz, H.C.; Curtis, C.T.; Buchheit, R.J.; Green, W.M.; Zuckerman, D.S.

    1979-11-01

    The overall purpose of the study is to identify design features of fusion power reactors which contribute to the achievement of high levels of maintainability. Previous phases evaluated several commercial tokamak reactor design concepts. This final phase compares the maintainability of a tandem mirror reactor (TMR) commercial conceptual design with the most maintainable tokamak concept selected from earlier work. A series of maintainability design guidelines and desirable TMR design features are defined. The effects of scheduled and unscheduled maintenance for most of the reactor subsystems are defined. The comparison of the TMR and tokamak reactor maintenance costs and availabilities show that both reactors have similar costs for scheduled maintenance at 19.4 and 20.8 million dollars annually and similar scheduled downtime availability impacts, achieving approximate availabilities of 79% at optimized maintenance intervals and cost of electricity.

  14. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  15. Neutronics requirements for a DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion Consortium , Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA UT-FUS C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2015-10-15

    Highlights: • Discussion and specification of neutronic requirements for a DEMO power plant. • TBR uncertainties are reviewed/discussed and design margins are elaborated. • Limits are given for radiation loads to super-conducting magnets and steel structural components. • Available DEMO results are compared to recommended limits and TBR design target. - Abstract: This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

  16. Tools for Predicting Optical Damage on Inertial Confinement Fusion-Class Laser Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nostrand, M C; Carr, C W; Liao, Z M; Honig, J; Spaeth, M L; Manes, K R; Johnson, M A; Adams, J J; Cross, D A; Negres, R A; Widmayer, C C; Williams, W H; Matthews, M J; Jancaitis, K S; Kegelmeyer, L M

    2010-12-20

    Operating a fusion-class laser to its full potential requires a balance of operating constraints. On the one hand, the total laser energy delivered must be high enough to give an acceptable probability for ignition success. On the other hand, the laser-induced optical damage levels must be low enough to be acceptably handled with the available infrastructure and budget for optics recycle. Our research goal was to develop the models, database structures, and algorithmic tools (which we collectively refer to as ''Loop Tools'') needed to successfully maintain this balance. Predictive models are needed to plan for and manage the impact of shot campaigns from proposal, to shot, and beyond, covering a time span of years. The cost of a proposed shot campaign must be determined from these models, and governance boards must decide, based on predictions, whether to incorporate a given campaign into the facility shot plan based upon available resources. Predictive models are often built on damage ''rules'' derived from small beam damage tests on small optics. These off-line studies vary the energy, pulse-shape and wavelength in order to understand how these variables influence the initiation of damage sites and how initiated damage sites can grow upon further exposure to UV light. It is essential to test these damage ''rules'' on full-scale optics exposed to the complex conditions of an integrated ICF-class laser system. Furthermore, monitoring damage of optics on an ICF-class laser system can help refine damage rules and aid in the development of new rules. Finally, we need to develop the algorithms and data base management tools for implementing these rules in the Loop Tools. The following highlights progress in the development of the loop tools and their implementation.

  17. MEASUREMENTS OF THE CONFINEMENT LEAKTIGHTNESS AT THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-08-01

    This is the final report on the INSP project entitled, ``Kola Confinement Leaktightness'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.1. This project was initiated in February 1993 to assist the Russians to reduce risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Units 1 and 2, through upgrades in the confinement performance to reduce the uncontrolled leakage rate. The major technical objective of this-project was to improve the leaktightness of the Kola NPP VVER confinement boundaries, through the application of a variety of sealants to penetrations, doors and hatches, seams and surfaces, to the extent that current technology permitted. A related objective was the transfer, through training of Russian staff, of the materials application procedures to the staff of the Kola NPP. This project was part of an overall approach to minimizing uncontrolled releases from the Kola NPP VVER440/230s in the event of a serious accident, and to thereby significantly mitigate the consequences of such an accident. The US provided materials, application technology, and applications equipment for application of sealant materials, surface coatings, potting materials and gaskets, to improve the confinement leaktightness of the Kola VVER-440/23Os. The US provided for training of Russian personnel in the applications technology.

  18. Simultaneous neutron and x-ray imaging of inertial confinement fusion experiments along a single line of sight at Omega.

    Science.gov (United States)

    Danly, C R; Day, T H; Fittinghoff, D N; Herrmann, H; Izumi, N; Kim, Y H; Martinez, J I; Merrill, F E; Schmidt, D W; Simpson, R A; Volegov, P L; Wilde, C H

    2015-04-01

    Neutron and x-ray imaging provide critical information about the geometry and hydrodynamics of inertial confinement fusion implosions. However, existing diagnostics at Omega and the National Ignition Facility (NIF) cannot produce images in both neutrons and x-rays along the same line of sight. This leads to difficulty comparing these images, which capture different parts of the plasma geometry, for the asymmetric implosions seen in present experiments. Further, even when opposing port neutron and x-ray images are available, they use different detectors and cannot provide positive information about the relative positions of the neutron and x-ray sources. A technique has been demonstrated on implosions at Omega that can capture x-ray images along the same line of sight as the neutron images. The technique is described, and data from a set of experiments are presented, along with a discussion of techniques for coregistration of the various images. It is concluded that the technique is viable and could provide valuable information if implemented on NIF in the near future.

  19. 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.

  20. Investigation of the geometrical efficiency for the two-dimensional space-resolving flux detector in inertial confinement fusion

    Science.gov (United States)

    Xie, X.; Ren, K.; Du, H.; Liu, S.; Chen, J.; Hou, L.; Li, Z.; Li, S.; Yang, D.; Guo, L.; Jiang, X.; Zhang, X.; Chen, Z.; Che, X.; Yang, G.; Yang, Y.; Jiang, S.; Ding, Y.

    2017-08-01

    Space-resolving flux detection is a novel technique for the precise diagnostic of the radiation field within the hohlraum in inertial confinement fusion. Detailed analysis is made on the geometrical efficiency of the space-resolving flux detector (SRFD), which is of essential importance for the application of the detector for the absolutely localized radiation flux measurement. With the two cascade collimation modules, namely the pinhole and the defining aperture, the field of view of the system could be classified as full-light area and half-light area, within which the solid angles of the point sources with respect to the sensitive detector are different. Therefore, many relevant factors could affect the systematical geometrical efficiency, such as the pattern of the pinhole and the spatial intensity distribution of the radiation source, and the influences of these factors are investigated explicitly. With this method, detailed analysis is made on the experimental radiation flux measured from the Au planar target, with the specific pattern of the pinhole.

  1. Measurement of Hydrodynamic Growth near Peak Velocity in an Inertial Confinement Fusion Capsule Implosion using a Self-Radiography Technique.

    Science.gov (United States)

    Pickworth, L A; Hammel, B A; Smalyuk, V A; MacPhee, A G; Scott, H A; Robey, H F; Landen, O L; Barrios, M A; Regan, S P; Schneider, M B; Hoppe, M; Kohut, T; Holunga, D; Walters, C; Haid, B; Dayton, M

    2016-07-15

    First measurements of hydrodynamic growth near peak implosion velocity in an inertial confinement fusion (ICF) implosion at the National Ignition Facility were obtained using a self-radiographing technique and a preimposed Legendre mode 40, λ=140  μm, sinusoidal perturbation. These are the first measurements of the total growth at the most unstable mode from acceleration Rayleigh-Taylor achieved in any ICF experiment to date, showing growth of the areal density perturbation of ∼7000×. Measurements were made at convergences of ∼5 to ∼10× at both the waist and pole of the capsule, demonstrating simultaneous measurements of the growth factors from both lines of sight. The areal density growth factors are an order of magnitude larger than prior experimental measurements and differed by ∼2× between the waist and the pole, showing asymmetry in the measured growth factors. These new measurements significantly advance our ability to diagnose perturbations detrimental to ICF implosions, uniquely intersecting the change from an accelerating to decelerating shell, with multiple simultaneous angular views.

  2. Basic hydrodynamics of Richtmyer-Meshkov-type growth and oscillations in the inertial confinement fusion-relevant conditions.

    Science.gov (United States)

    Aglitskiy, Y; Velikovich, A L; Karasik, M; Metzler, N; Zalesak, S T; Schmitt, A J; Phillips, L; Gardner, J H; Serlin, V; Weaver, J L; Obenschain, S P

    2010-04-13

    In inertial confinement fusion (ICF), the possibility of ignition or high energy gain is largely determined by our ability to control the Rayleigh-Taylor (RT) instability growth in the target. The exponentially amplified RT perturbation eigenmodes are formed from all sources of the target and radiation non-uniformity in a process called seeding. This process involves a variety of physical mechanisms that are somewhat similar to the classical Richtmyer-Meshkov (RM) instability (in particular, most of them are active in the absence of acceleration), but differ from it in many ways. In the last decade, radiographic diagnostic techniques have been developed that made direct observations of the RM-type effects in the ICF-relevant conditions possible. New experiments stimulated the advancement of the theory of the RM-type processes. The progress in the experimental and theoretical studies of such phenomena as ablative RM instability, re-shock of the RM-unstable interface, feedout and perturbation development associated with impulsive loading is reviewed.

  3. Observation and modeling of interspecies ion separation in inertial confinement fusion implosions via imaging x-ray spectroscopy

    Science.gov (United States)

    Joshi, T. R.; Hakel, P.; Hsu, S. C.; Vold, E. L.; Schmitt, M. J.; Hoffman, N. M.; Rauenzahn, R. M.; Kagan, G.; Tang, X.-Z.; Mancini, R. C.; Kim, Y.; Herrmann, H. W.

    2017-05-01

    We report the first direct experimental evidence of interspecies ion separation in direct-drive inertial confinement fusion experiments performed at the OMEGA laser facility via spectrally, temporally, and spatially resolved imaging x-ray-spectroscopy data [S. C. Hsu et al., Europhys. Lett. 115, 65001 (2016)]. These experiments were designed based on the expectation that interspecies ion thermo-diffusion would be the strongest for species with a large mass and charge difference. The targets were spherical plastic shells filled with D2 and a trace amount of Ar (0.1% or 1% by atom). Ar K-shell spectral features were observed primarily between the time of first-shock convergence and slightly before the neutron bang time, using a time- and space-integrated spectrometer, a streaked crystal spectrometer, and two gated multi-monochromatic x-ray imagers fielded along quasi-orthogonal lines of sight. Detailed spectroscopic analyses of spatially resolved Ar K-shell lines reveal the deviation from the initial 1% Ar gas fill and show both Ar-concentration enhancement and depletion at different times and radial positions of the implosion. The experimental results are interpreted using radiation-hydrodynamic simulations that include recently implemented, first-principles models of interspecies ion diffusion. The experimentally inferred Ar-atom fraction profiles agree reasonably with calculated profiles associated with the incoming and rebounding first shock.

  4. A wave-based model for cross-beam energy transfer in direct-drive inertial confinement fusion

    Science.gov (United States)

    Myatt, J. F.; Follett, R. K.; Shaw, J. G.; Edgell, D. H.; Froula, D. H.; Igumenshchev, I. V.; Goncharov, V. N.

    2017-05-01

    Cross-beam energy transfer (CBET) is thought to be responsible for a 30% reduction in hydrodynamic coupling efficiency on OMEGA and up to 50% at the ignition scale for direct-drive (DD) implosions. These numbers are determined by ray-based models that have been developed and integrated within the radiation-hydrodynamics codes LILAC (1-D) and DRACO (2-D). However, ray-based modeling of CBET in an inhomogeneous plasma assumes a steady-state plasma response, does not include the effects of beam speckle, and treats ray caustics in an ad hoc manner. The validity of the modeling for ignition-scale implosions has not yet been determined. To address the physics shortcomings, which have important implications for DD inertial confinement fusion, a new wave-based model has been developed. It solves the time-enveloped Maxwell equations in three dimensions, including polarization effects, plasma inhomogeneity, and open-boundary conditions with the ability to prescribe beams incident at arbitrary angles. Beams can be made realistic with respect to laser speckle, polarization smoothing, and laser bandwidth. This, coupled to a linearized low-frequency plasma response that does not assume a steady state, represents the most-complete model of CBET to date.

  5. Mitigation of cross-beam energy transfer in direct-drive inertial-confinement-fusion implosions with enhanced laser bandwidth

    Science.gov (United States)

    Bates, Jason; Myatt, Jason; Shaw, John; Follett, Russell; Weaver, James; Lehmberg, Robert; Obenschain, Stephen

    2017-10-01

    Cross-beam energy transfer (CBET) is a special category of stimulated Brillouin scattering in which two overlapping laser beams exchange energy by means of an ion acoustic wave in an under-dense expanding plasma. CBET can cause the incident laser energy to be misdirected in direct-drive inertial-confinement-fusion (ICF) implosions, thereby reducing both the maximum ablation pressure achieved and the overall symmetry of the implosion. One strategy for mitigating CBET may be to increase the bandwidth of the laser light, thereby disrupting the coherent wave-wave interactions underlying this resonant parametric process. In this presentation, we report on results of two-dimensional planar simulations performed with the code LPSE-CBET that demonstrate a significant reduction in CBET for bandwidths between 2 and 5 THz. Although large compared to OMEGA and NIF values (about 1 and 0.3 THz, respectively), it may be possible to reach such bandwidths with existing ICF lasers using a technique based on stimulated rotational Raman scattering, which is a subject that we also briefly discuss. Work supported by DOE/NNSA.

  6. A diamond detector for inertial confinement fusion X-ray bang-time measurements at the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    MacPhee, A G; Brown, C; Burns, S; Celeste, J; Glenzer, S H; Hey, D; Jones, O S; Landen, O; Mackinnon, A J; Meezan, N; Parker, J; Edgell, D; Glebov, V Y; Kilkenny, J; Kimbrough, J

    2010-11-09

    An instrument has been developed to measure X-ray bang-time for inertial confinement fusion capsules; the time interval between the start of the laser pulse and peak X-ray emission from the fuel core. The instrument comprises chemical vapor deposited polycrystalline diamond photoconductive X-ray detectors with highly ordered pyrolytic graphite X-ray monochromator crystals at the input. Capsule bang-time can be measured in the presence of relatively high thermal and hard X-ray background components due to the selective band pass of the crystals combined with direct and indirect X-ray shielding of the detector elements. A five channel system is being commissioned at the National Ignition Facility at Lawrence Livermore National Laboratory for implosion optimization measurements as part of the National Ignition Campaign. Characteristics of the instrument have been measured demonstrating that X-ray bang-time can be measured with {+-} 30ps precision, characterizing the soft X-ray drive to +/- 1eV or 1.5%.

  7. Note: Light output enhanced fast response and low afterglow 6Li glass scintillator as potential down-scattered neutron diagnostics for inertial confinement fusion.

    Science.gov (United States)

    Arikawa, Yasunobu; Yamanoi, Kohei; Nagai, Takahiro; Watanabe, Kozue; Kouno, Masahiro; Sakai, Kohei; Nakazato, Tomoharu; Shimizu, Toshihiko; Cadatal, Marilou Raduban; Estacio, Elmer Surat; Sarukura, Nobuhiko; Nakai, Mitsuo; Norimatsu, Takayoshi; Azechi, Hiroshi; Murata, Takahiro; Fujino, Shigeru; Yoshida, Hideki; Izumi, Nobuhiko; Satoh, Nakahiro; Kan, Hirofumi

    2010-10-01

    The characteristics of an APLF80+3Ce scintillator are presented. Its sufficiently fast decay profile, low afterglow, and an improved light output compared to the recently developed APLF80+3Pr, were experimentally demonstrated. This scintillator material holds promise for applications in neutron imaging diagnostics at the energy regions of 0.27 MeV of DD fusion down-scattered neutron peak at the world's largest inertial confinement fusion facilities such as the National Ignition Facility and the Laser Mégajoule.

  8. Solenoid transport of a heavy ion beam for warm dense matterstudies and inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Armijo, Julien

    2006-10-01

    From February to July 2006, I have been doing research as a guest at Lawrence Berkeley National Laboratory (LBNL), in the Heavy Ion Fusion group. This internship, which counts as one semester in my master's program in France, I was very pleased to do it in a field that I consider has the beauty of fundamental physics, and at the same time the special appeal of a quest for a long-term and environmentally-respectful energy source. During my stay at LBNL, I have been involved in three projects, all of them related to Neutralized Drift Compression Experiment (NDCX). The first one, experimental and analytical, has consisted in measuring the effects of the eddy currents induced by the pulsed magnets in the conducting plates of the source and diagnostic chambers of the Solenoid Transport Experiment (STX, which is a subset of NDCX). We have modeled the effect and run finite-element simulations that have reproduced the perturbation to the field. Then, we have modified WARP, the Particle-In-Cell code used to model the whole experiment, in order to import realistic fields including the eddy current effects and some details of each magnet. The second project has been to take part in a campaign of WARP simulations of the same experiment to understand the leakage of electrons that was observed in the experiment as a consequence to some diagnostics and the failure of the electrostatic electron trap. The simulations have shown qualitative agreement with the measured phenomena, but are still in progress. The third project, rather theoretical, has been related to the upcoming target experiment of a thin aluminum foil heated by a beam to the 1-eV range. At the beginning I helped by analyzing simulations of the hydrodynamic expansion and cooling of the heated material. But, progressively, my work turned into making estimates for the nature of the liquid/vapor two-phase flow. In particular, I have been working on criteria and models to predict the formation of droplets, their size

  9. X-ray shadow imprint of hydrodynamic instabilities on the surface of inertial confinement fusion capsules by the fuel fill tube

    Science.gov (United States)

    MacPhee, A. G.; Casey, D. T.; Clark, D. S.; Felker, S.; Field, J. E.; Haan, S. W.; Hammel, B. A.; Kroll, J.; Landen, O. L.; Martinez, D. A.; Michel, P.; Milovich, J.; Moore, A.; Nikroo, A.; Rice, N.; Robey, H. F.; Smalyuk, V. A.; Stadermann, M.; Weber, C. R.

    2017-03-01

    Measurements of hydrodynamic instability growth for a high-density carbon ablator for indirectly driven inertial confinement fusion implosions on the National Ignition Facility are reported. We observe significant unexpected features on the capsule surface created by shadows of the capsule fill tube, as illuminated by laser-irradiated x-ray spots on the hohlraum wall. These shadows increase the spatial size and shape of the fill tube perturbation in a way that can significantly degrade performance in layered implosions compared to previous expectations. The measurements were performed at a convergence ratio of ˜2 using in-flight x-ray radiography. The initial seed due to shadow imprint is estimated to be equivalent to ˜50-100 nm of solid ablator material. This discovery has prompted the need for a mitigation strategy for future inertial confinement fusion designs as proposed here.

  10. Noise factor of a high-speed cinematography system; Facteur de bruit d'une chaine de cinematographie ultrarapide: application a la fusion par confinement inertiel

    Energy Technology Data Exchange (ETDEWEB)

    Secroun, A

    2000-03-01

    Inertial confinement fusion simulates in a laboratory the thermodynamic state of the center of stars, thus leading to the determination of stellar parameters. In order to reach that aim, high-speed cinematography brings up instruments specifically adapted to picosecond measurement, for which it is necessary to know the final precision. A model of the noise factor of the instruments under study is introduced and confronted to the experimental results obtained. (authors)

  11. Experimental studies of materials migration in magnetic confinement fusion devices : Novel methods for measurement of macro particle migration, transport of atomic impurities and characterization of exposed surfaces

    OpenAIRE

    Bykov, Igor

    2014-01-01

    During several decades of research and development in the field of Magnetically Confined Fusion (MCF) the preferred selection of materials for Plasma Facing Components (PFC) has changed repeatedly. Without doubt, endurance of the first wall will decide research availability and lifespan of the first International Thermonuclear Research Reactor (ITER). Materials erosion, redeposition and mixing in the reactor are the critical processes responsible for modification of materials properties under...

  12. Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Broden, K. [Association EURATOM, Nykoeping (Sweden); Lindberg, M. [Association EURATOM, Nykoeping (Sweden); Nisan, S. [The NET Team, Garching (Germany); Rocco, P. [European Commission, Institute for Advanced Materials, Joint Research Centre, Ispra (Vatican City State, Holy See) (Italy); Zucchetti, M. [Energetics Department, Polytechnic of Turin, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy); Taylor, N. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Forty, C. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

    1997-04-01

    An important inherent advantage of fusion would be the total absence of high-level radioactive spent fuel as produced in fission reactors. Fusion will, however, produce activated material containing both activation products and tritium. Part of the material may also contain chemically toxic substances. This paper describes methods that could be used to manage these materials and also methods to reduce or entirely eliminate the waste quantities. The results are based on studies for the International Thermonuclear Experimental Reactor and also for future fusion power station designs currently under investigation within the European programme on the safety and environmental assessment of fusion power, long-term. (orig.)

  13. Controlled fusion; La fusion controlee

    Energy Technology Data Exchange (ETDEWEB)

    Bobin, J.L

    2005-07-01

    During the last fifty years the researches on controlled thermonuclear fusion reached great performance in the magnetic confinement (tokamaks) as in the inertial confinement (lasers). But the state of the art is not in favor of the apparition of the fusion in the energy market before the second half of the 21 century. To explain this opinion the author presents the fusion reactions of light nuclei and the problems bound to the magnetic confinement. (A.L.B.)

  14. Magnetic fusion reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A.

    1995-12-01

    An almost primordial trend in the conversion and use of energy is an increased complexity and cost of conversion systems designed to utilize cheaper and more-abundant fuels; this trend is exemplified by the progression fossil fission {yields} fusion. The present projections of the latter indicate that capital costs of the fusion ``burner`` far exceed any commensurate savings associated with the cheapest and most-abundant of fuels. These projections suggest competitive fusion power only if internal costs associate with the use of fossil or fission fuels emerge to make them either uneconomic, unacceptable, or both with respect to expensive fusion systems. This ``implementation-by-default`` plan for fusion is re-examined by identifying in general terms fusion power-plant embodiments that might compete favorably under conditions where internal costs (both economic and environmental) of fossil and/or fission are not as great as is needed to justify the contemporary vision for fusion power. Competitive fusion power in this context will require a significant broadening of an overly focused program to explore the physics and simbiotic technologies leading to more compact, simplified, and efficient plasma-confinement configurations that reside at the heart of an attractive fusion power plant.

  15. Inertial confinement fusion. Quarterly report, July--September 1995, Volume 5, Number 4

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-06-01

    The 1990 National Academy of Sciences (NAS) final report recommended proceeding with the construction of a 1- to 2-MJ Nd-doped glass laser designed to achieve ignition in the laboratory (a laser originally called the Nova Upgrade, but now called the National Ignition Facility, or NIF, and envisioned as a national user facility). As a prerequisite, the report recommended completion of a series of target physics objectives on the Nova laser in use at the Lawrence Livermore National Laboratory (LLNL). Meeting these objectives, which were called the Nova Technical Contract (NTC), would demonstrate (the Academy committee believed) that the physics of ignition targets was understood well enough that the laser requirements could be accurately specified. Completion of the NTC objectives was given the highest priority in the NAS report. The NAS committee also recommended a concentrated effort on advanced target design for ignition. As recommended in the report, completion of these objectives has been the joint responsibility of LLNL and the Los Alamos National Laboratory. Most of the articles in this issue of the ICF Quarterly were written jointly by scientists from both institutions. Several of the NTC objectives required the completion of improvements to Nova`s power balance and pointing accuracy and of new diagnostics and new target fabrication capabilities. These improvements were called {open_quotes}Precision Nova{close_quotes} and are documented. The original NTC objectives have been largely met. This Introduction summarizes those objectives and their motivation in the context of the requirements for ignition. The articles that follow describe the NIF ignition target designs and summarize the principal accomplishments in the various elements of the NTC.

  16. Environmental impacts of nonfusion power systems. [Data on environmental effects of all power sources that may be competitive with fusion reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Brouns, R.J.

    1976-09-01

    Data were collected on the environmental effects of power sources that may be competitive with future fusion reactor power plants. Data are included on nuclear power plants using HTGR, LMBR, GCFR, LMFBR, and molten salt reactors; fossil-fuel electric power plants; geothermal power plants; solar energy power plants, including satellite-based solar systems; wind energy power plants; ocean thermal gradient power plants; tidal energy power plants; and power plants using hydrogen and other synthetic fuels as energy sources.

  17. Effect of the laser wavelength: A long story of laser-plasma interaction physics for Inertial Confinement Fusion Teller Medal Lecture

    Directory of Open Access Journals (Sweden)

    Labaune Christine

    2013-11-01

    Full Text Available Laser-driven Inertial Confinement Fusion (ICF relies on the use of high-energy laser beams to compress and ignite a thermonuclear fuel with the ultimate goal of producing energy. Fusion is the holy grail of energy sources–combining abundant fuel with no greenhouse gas emissions, minimal waste products and a scale that can meet mankind's long-term energy demands. The quality and the efficiency of the coupling of the laser beams with the target are an essential step towards the success of laser fusion. A long-term program on laser-plasma interaction physics has been pursued to understand the propagation and the coupling of laser pulses in plasmas for a wide range of parameters.

  18. Neutronic Analysis of the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine Using Various Thorium Molten Salts

    Science.gov (United States)

    Acır, Adem

    2013-08-01

    In this study, a neutronic performance of the Laser Inertial Confinement Fusion Fission Energy (LIFE) molten salt blanket is investigated. Neutronic calculations are performed by using XSDRNPM/SCALE5 codes in S8-P3 approximation. The thorium molten salt composition considered in this calculation is 75 % LiF—25 % ThF4, 75 % LiF—24 % ThF4—1 % 233UF4, 75 % LiF—23 % ThF4—2 % 233UF4. Also, effects of the 6Li enrichment in molten salt are performed for all heavy metal salt. The radiation damage behaviors of SS-304 structural material with respect to higher fissionable fuel content and 6Li enrichment are computed. By higher fissionable fuel content in molten salt and with 6Li enrichment (20 and 50 %) in the coolant in form of 75 % LiF—23 % ThF4—2 % 233UF4, an initial TBR >1.05 can be realized. On the other hand, the 75 % LiF—25 % ThF4 or 75 % LiF—24 % ThF4—1 % 233UF4 molten salt fuel as regards maintained tritium self-sufficiency is not suitable as regards improving neutronic performance of LIFE engine. A high quality fissile fuel with a rate of ~2,850 kg/year of 233U can be produced with 75 % LiF—23 % ThF4—2 % 233UF4. The energy multiplication factor is increased with high rate fission reactions of 233U occurring in the molten salt zone. Major damage mechanisms in SS-304 first wall stell have been computed as DPA = 48 and He = 132 appm per year with 75 % LiF—23 % ThF4—2 % 233UF4. This implies a replacement of the SS-304 first wall stell of every between 3 and 4 years.

  19. Three-dimensional simulation strategy to determine the effects of turbulent mixing on inertial-confinement-fusion capsule performance.

    Science.gov (United States)

    Haines, Brian M; Grinstein, Fernando F; Fincke, James R

    2014-05-01

    In this paper, we present and justify an effective strategy for performing three-dimensional (3D) inertial-confinement-fusion (ICF) capsule simulations. We have evaluated a frequently used strategy in which two-dimensional (2D) simulations are rotated to 3D once sufficient relevant 2D flow physics has been captured and fine resolution requirements can be restricted to relatively small regions. This addresses situations typical of ICF capsules which are otherwise prohibitively intensive computationally. We tested this approach for our previously reported fully 3D simulations of laser-driven reshock experiments where we can use the available 3D data as reference. Our studies indicate that simulations that begin as purely 2D lead to significant underprediction of mixing and turbulent kinetic energy production at later time when compared to the fully 3D simulations. If, however, additional suitable nonuniform perturbations are applied at the time of rotation to 3D, we show that one can obtain good agreement with the purely 3D simulation data, as measured by vorticity distributions as well as integrated mixing and turbulent kinetic energy measurements. Next, we present results of simulations of a simple OMEGA-type ICF capsule using the developed strategy. These simulations are in good agreement with available experimental data and suggest that the dominant mechanism for yield degradation in ICF implosions is hydrodynamic instability growth seeded by long-wavelength surface defects. This effect is compounded by drive asymmetries and amplified by repeated shock interactions with an increasingly distorted shell, which results in further yield reduction. Our simulations are performed with and without drive asymmetries in order to compare the importance of these effects to those of surface defects; our simulations indicate that long-wavelength surface defects degrade yield by approximately 60% and short-wavelength drive asymmetry degrades yield by a further 30%.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J. [ed.

    1998-12-01

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

  1. Advanced smart tungsten alloys for a future fusion power plant

    Science.gov (United States)

    Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch; Rasinski, M.; Kreter, A.; Tan, X.; Schmitz, J.; Mao, Y.; Coenen, J. W.; Bram, M.; Gonzalez-Julian, J.

    2017-06-01

    The severe particle, radiation and neutron environment in a future fusion power plant requires the development of advanced plasma-facing materials. At the same time, the highest level of safety needs to be ensured. The so-called loss-of-coolant accident combined with air ingress in the vacuum vessel represents a severe safety challenge. In the absence of a coolant the temperature of the tungsten first wall may reach 1200 °C. At such a temperature, the neutron-activated radioactive tungsten forms volatile oxide which can be mobilized into atmosphere. Smart tungsten alloys are being developed to address this safety issue. Smart alloys should combine an acceptable plasma performance with the suppressed oxidation during an accident. New thin film tungsten-chromium-yttrium smart alloys feature an impressive 105 fold suppression of oxidation compared to that of pure tungsten at temperatures of up to 1000 °C. Oxidation behavior at temperatures up to 1200 °C, and reactivity of alloys in humid atmosphere along with a manufacturing of reactor-relevant bulk samples, impose an additional challenge in smart alloy development. First exposures of smart alloys in steady-state deuterium plasma were made. Smart tungsten-chroimium-titanium alloys demonstrated a sputtering resistance which is similar to that of pure tungsten. Expected preferential sputtering of alloying elements by plasma ions was confirmed experimentally. The subsequent isothermal oxidation of exposed samples did not reveal any influence of plasma exposure on the passivation of alloys.

  2. Issues in the commercialization of magnetic fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Rockwood, A.D.; Willke, T.L.

    1979-12-01

    This study identifies and outlines the issues that must be considered if fusion is to be put into commercial practice. The issues are put into perspective around a consistent framework and a program of study and research is recommended to anticipate and handle the issues for a successful fusion commercialization program. (MOW)

  3. Commercial objectives, technology transfer, and systems analysis for fusion power development

    Science.gov (United States)

    Dean, Stephen O.

    1988-03-01

    Fusion is an essentially inexhaustible source of energy that has the potential for economically attractive commercial applications with excellent safety and environmental characteristics. The primary focus for the fusion-energy development program is the generation of centralstation electricity. Fusion has the potential, however, for many other applications. The fact that a large fraction of the energy released in a DT fusion reaction is carried by high-energy neutrons suggests potentially unique applications. These include breeding of fissile fuels, production of hydrogen and other chemical products, transmutation or “burning” of various nuclear or chemical wastes, radiation processing of materials, production of radioisotopes, food preservation, medical diagnosis and medical treatment, and space power and space propulsion. In addition, fusion R&D will lead to new products and new markets. Each fusion application must meet certain standards of economic and safety and environmental attractiveness. For this reason, economics on the one hand, and safety and environment and licensing on the other hand, are the two primary criteria for setting long-range commercial fusion objectives. A major function of systems analysis is to evaluate the potential of fusion against these objectives and to help guide the fusion R&D program toward practical applications. The transfer of fusion technology and skills from the national laboratories and universities to industry is the key to achieving the long-range objective of commercial fusion applications.

  4. Comment on ‘On the fusion triple product and fusion power gain of tokamak pilot plants and reactors’, by A. Costley

    Science.gov (United States)

    Biel, W.; Lackner, K.; Sauter, O.; Wenninger, R.; Zohm, H.

    2017-03-01

    In this comment, we discuss the arguments raised in two recent papers (Costley 2016 Nucl. Fusion 56 066003, Costley et al 2015 Nucl. Fusion 55 033001) on the claimed size independence of fusion power, triple product and fusion gain in tokamak reactors, and we show that all these three quantities actually do depend on the size of the tokamak, when distinguishing between independent input parameters (design parameters) and output quantities, and when taking into account technological limitations.

  5. The performance simulation of single cylinder electric power confined piston engine

    Science.gov (United States)

    Gou, Yanan

    2017-04-01

    A new type of power plant. i.e, Electric Power Confined Piston Engine, is invented by combining the free piston engine and the crank connecting rod mechanism of the traditional internal combustion engine. Directly using the reciprocating movement of the piston, this new engine converts the heat energy produced by fuel to electrical energy and output it. The paper expounds the working mechanism of ECPE and establishes the kinematics and dynamics equations. Furthermore, by using the analytic method, the ECPE electromagnetic force is solved at load cases. Finally, in the simulation environment of MARLAB, the universal characteristic curve is obtained in the condition of rotational speed n between 1000 r/min and 2400 r/min, throttle opening α between 30% and 100%.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  7. Status and future prospects of laser fusion and high power laser applications

    Science.gov (United States)

    Mima, Kunioki

    2010-08-01

    In Asia, there are many institutes for the R&D of high power laser science and applications. They are 5 major institutes in Japan, 4 major institutes in China, 2 institutes in Korea, and 3 institutes in India. The recent achievements and future prospects of those institutes will be over viewed. In the laser fusion research, the FIREX-I project in Japan has been progressing. The 10kJ short pulse LFEX laser has completed and started the experiments with a single beam. About 1kJ pulse energy will be injected into a cone target. The experimental results of the FIREX experiments will be presented. As the target design for the experiments, a new target, namely, a double cone target was proposed, in which the high energy electrons are well confined and the heating efficiency is significantly improved. Together with the fusion experiments, Osaka University has carried out laboratory astrophysics experiments on photo ionizing plasmas to observe a unique X-ray spectrum from non-LTE plasmas. In 2008, Osaka university has started a new Photon research center in relation with the new program: Consortium for Photon Science and Technology: C-PhoST, in which ultra intense laser plasmas research and related education will be carried out for 10 years. At APRI, JAEA, the fundamental science on the relativistic laser plasmas and the applications of laser particle acceleration has been developed. The application of laser ion acceleration has been investigated on the beam cancer therapy since 2007. In China, The high power glass laser: Shenguan-II and a peta watt beam have been operated to work on radiation hydro dynamics at SIOFM Shanghai. The laser material and optics are developed at SIOFM and LFRC. The IAPCM and the IOP continued the studies on radiation hydrodynamics and on relativistic laser plasmas interactions. At LFRC in China, the construction of Shenguan III glass laser of 200kJ in blue has progressed and will be completed in 2012. Together with the Korean program, I will

  8. Fill-tube-induced mass perturbations on x-ray-driven, ignition-scale, inertial-confinement-fusion capsule shells and the implications for ignition experiments.

    Science.gov (United States)

    Bennett, G R; Herrmann, M C; Edwards, M J; Spears, B K; Back, C A; Breden, E W; Christenson, P J; Cuneo, M E; Dannenburg, K L; Frederick, C; Keller, K L; Mulville, T D; Nikroo, A; Peterson, K; Porter, J L; Russell, C O; Sinars, D B; Smith, I C; Stamm, R M; Vesey, R A

    2007-11-16

    On the first inertial-confinement-fusion ignition facility, the target capsule will be DT filled through a long, narrow tube inserted into the shell. microg-scale shell perturbations Delta m' arising from multiple, 10-50 microm-diameter, hollow SiO2 tubes on x-ray-driven, ignition-scale, 1-mg capsules have been measured on a subignition device. Simulations compare well with observation, whence it is corroborated that Delta m' arises from early x-ray shadowing by the tube rather than tube mass coupling to the shell, and inferred that 10-20 microm tubes will negligibly affect fusion yield on a full-ignition facility.

  9. Ion microtomography (IMT) and particle-induced x-ray emission (PIXE) analysis direct drive of inertial confinement fusion (ICF) targets

    Energy Technology Data Exchange (ETDEWEB)

    Antolak, A.J.; Pontau, A.E.; Morse, D.H. (Sandia National Labs., Livermore, CA (United States)); Weirup, D.L.; Heikkinen, D.W.; Hornady, R.S. (Lawrence Livermore National Lab., CA (United States)); Cholewa, M.; Bench, G.S.; Legge, G.J.F. (Melbourne Univ. (Australia). Micro Analytical Research Centre)

    1991-11-20

    The complementary techniques of ion microtomography (IMT) and particle-induced x-ray emission (PIXE) are used to provide micro-characterization of inertial confinement fusion (ICF) targets for density uniformity, sphericity, and trace element spatial distributions. ICF target quality control in the laser fusion program is important to ensure that the energy deposition from the lasers results in uniform compression and minimization of Taylor-Rayleigh instabilities. We obtain 1% density determinations using IMT with spatial resolution approaching two microns. Utilizing PIXE, we can map out dopant and impurity distributions with elemental detection sensitivities on the order of a few ppm. We present examples of IMT and PIXE analyses performed on several ICF targets.

  10. Confined ion energy >200 keV and increased fusion yield in a DPF with monolithic tungsten electrodes and pre-ionization

    Science.gov (United States)

    Lerner, Eric J.; Hassan, Syed M.; Karamitsos, Ivana; Von Roessel, Fred

    2017-10-01

    To reduce impurities in the dense plasma focus FF-1 device, we used monolithic tungsten electrodes with pre-ionization. With this new set-up, we demonstrated a three-fold reduction of impurities by mass and a ten-fold reduction by ion number. FF-1 produced a 50% increase in fusion yield over our previous copper electrodes, both for a single shot and for a mean of ten consecutive shots with the same conditions. These results represent a doubling of fusion yield as compared with any other plasma focus device with the same 60 kJ energy input. In addition, FF-1 produced a new single-shot record of 240 ± 20 keV for mean ion energy, a record for any confined fusion plasma, using any device, and a 50% improvement in ten-shot mean ion energy. With a deuterium-nitrogen mix and corona-discharge pre-ionization, we were also able to reduce the standard deviation in the fusion yield to about 15%, a four-fold reduction over the copper-electrode results. We intend to further reduce impurities with new experiments using microwave treatment of tungsten electrodes, followed by the use of beryllium electrodes.

  11. High power microwave diagnostic for the fusion energy experiment ITER

    DEFF Research Database (Denmark)

    Korsholm, Søren Bang; Leipold, Frank; Goncalves, B.

    2016-01-01

    Microwave diagnostics will play an increasingly important role in burning plasma fusion energy experiments like ITER and beyond. The Collective Thomson Scattering (CTS) diagnostic to be installed at ITER is an example of such a diagnostic with great potential in present and future experiments....... The ITER CTS diagnostic will inject a 1 MW 60 GHz gyrotron beam into the ITER plasma and observe the scattering off fluctuations in the plasma — to monitor the dynamics of the fast ions generated in the fusion reactions....

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

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

  13. Power requirements for superior H-mode confinement on Alcator C-Mod: experiments in support of ITER

    Science.gov (United States)

    Hughes, J. W.; Loarte, A.; Reinke, M. L.; Terry, J. L.; Brunner, D.; Greenwald, M.; Hubbard, A. E.; LaBombard, B.; Lipschultz, B.; Ma, Y.; Wolfe, S.; Wukitch, S. J.

    2011-08-01

    Power requirements for maintaining sufficiently high confinement (i.e. normalized energy confinement time H98 >= 1) in H-mode and its relation to H-mode threshold power scaling, Pth, are of critical importance to ITER. In order to better characterize these power requirements, recent experiments on the Alcator C-Mod tokamak have investigated H-mode properties, including the edge pedestal and global confinement, over a range of input powers near and above Pth. In addition, we have examined the compatibility of impurity seeding with high performance operation, and the influence of plasma radiation and its spatial distribution on performance. Experiments were performed at 5.4 T at ITER relevant densities, utilizing bulk metal plasma facing surfaces and an ion cyclotron range of frequency waves for auxiliary heating. Input power was scanned both in stationary enhanced Dα (EDA) H-modes with no large edge localized modes (ELMs) and in ELMy H-modes in order to relate the resulting pedestal and confinement to the amount of power flowing into the scrape-off layer, Pnet, and also to the divertor targets. In both EDA and ELMy H-mode, energy confinement is generally good, with H98 near unity. As Pnet is reduced to levels approaching that in L-mode, pedestal temperature diminishes significantly and normalized confinement time drops. By seeding with low-Z impurities, such as Ne and N2, high total radiated power fractions are possible, along with substantial reductions in divertor heat flux (>4×), all while maintaining H98 ~ 1. When the power radiated from the confined versus unconfined plasma is examined, pedestal and confinement properties are clearly seen to be an increasing function of Pnet, helping to unify the results with those from unseeded H-modes. This provides increased confidence that the power flow across the separatrix is the correct physics basis for ITER extrapolation. The experiments show that Pnet/Pth of one or greater is likely to lead to H98 >= 1 operation

  14. High power microwave diagnostic for the fusion energy experiment ITER

    DEFF Research Database (Denmark)

    Korsholm, Søren Bang; Leipold, Frank; Goncalves, B.

    2016-01-01

    Microwave diagnostics will play an increasingly important role in burning plasma fusion energy experiments like ITER and beyond. The Collective Thomson Scattering (CTS) diagnostic to be installed at ITER is an example of such a diagnostic with great potential in present and future experiments. Th...

  15. Apparatus and method for extracting power from energetic ions produced in nuclear fusion

    Science.gov (United States)

    Fisch, Nathaniel J.; Rax, Jean M.

    1994-01-01

    An apparatus and method of extracting power from energetic ions produced by nuclear fusion in a toroidal plasma to enhance respectively the toroidal plasma current and fusion reactivity. By injecting waves of predetermined frequency and phase traveling substantially in a selected poloidal direction within the plasma, the energetic ions become diffused in energy and space such that the energetic ions lose energy and amplify the waves. The amplified waves are further adapted to travel substantially in a selected toroidal direction to increase preferentially the energy of electrons traveling in one toroidal direction which, in turn, enhances or generates a toroidal plasma current. In an further adaptation, the amplified waves can be made to preferentially increase the energy of fuel ions within the plasma to enhance the fusion reactivity of the fuel ions. The described direct, or in situ, conversion of the energetic ion energy provides an efficient and economical means of delivering power to a fusion reactor.

  16. Reply to ‘Comment “On the fusion triple product and fusion power gain of tokamak pilot plants and reactors”’

    Science.gov (United States)

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

    2017-03-01

    In reply to the Comment by Biel et al (2016 Nucl. Fusion 57 038001) on our recent papers Costley et al (2015 Nucl Fusion 55 033001) and Costley (2016 Nucl. Fusion 56 066003), we point out that the fusion triple product, nTτ E, and fusion power gain, Q fus, cannot be expressed solely in terms of independent engineering design variables such as major radius, R, and toroidal field, B; output performance variables such as normalised beta, β N, safety factor, q, and fusion power P fus, have to be invoked. Further, we show that the density limit has the effect of largely cancelling the size dependence in nTτ E and Q fus, which would otherwise be present, when these parameters are expressed in terms of P fus. Considerations of engineering aspects are also briefly discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-01

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

  18. Flywheel induction motor-generator for magnet power supply in small fusion device

    Energy Technology Data Exchange (ETDEWEB)

    Hatakeyma, S., E-mail: hatakeyama.shoichi@torus.nr.titech.ac.jp; Yoshino, F.; Tsutsui, H.; Tsuji-Iio, S. [Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2016-04-15

    A flywheel motor-generator (MG) for the toroidal field (TF) coils of a small fusion device was developed which utilizes a commercially available squirrel-cage induction motor. Advantages of the MG are comparably-long duration, quick power response, and easy implementation of power control compared with conventional capacitor-type power supply. A 55-kW MG was fabricated, and TF coils of a small fusion device were energized. The duration of the current flat-top was extended to 1 s which is much longer than those of conventional small devices (around 10–100 ms).

  19. Fusion blanket for high-efficiency power cycles

    Energy Technology Data Exchange (ETDEWEB)

    Usher, J.L.; Powell, J.R.; Fillo, J.A.; Horn, F.L.; Lazareth, O.W.; Taussig, R.

    1980-01-01

    The efficiencies of blankets for fusion reactors are usually in the range of 30 to 40%, limited by the operating temperature (500/sup 0/C) of conventional structural materials such as stainless steels. In this project two-zone blankets are proposed; these blankets consist of a low-temperature shell surrounding a high-temperature interior zone. A survey of nucleonics and thermal hydraulic parameters has led to a reference blanket design consisting of a water-cooled stainless steel shell around a BeO, ZrO/sub 2/ interior (cooled by Ar) utilizing Li/sub 2/O for tritium breeding. In this design, approx. 60% of the fusion energy is deposited in the high-temperature interior. The maximum Ar temperature is 2230/sup 0/C leading to an overall efficiency estimate of 55 to 60% for this reference case.

  20. Cost Modeling and Design of Field-Reversed Configuration Fusion Power Plants

    Science.gov (United States)

    Kirtley, David; Slough, John; Helion Team

    2017-10-01

    The Inductively Driven Liner (IDL) fusion concept uses the magnetically driven implosion of thin (0.5-1 mm) Aluminum hoops to magnetically compress a merged Field-Reversed Configuration (FRC) plasma to fusion conditions. Both the driver and the target have been studied experimentally and theoretically by researchers at Helion Energy, MSNW, and the University of Washington, demonstrating compression fields greater than 100 T and suitable fusion targets. In the presented study, a notional power plant facility using this approach will be described. In addition, a full cost study based on the LLNL Z-IFE and HYLIFE-II studies, the ARIES Tokamak concept, and RAND power plant studies will be described. Finally, the expected capital costs, development requirements, and LCOE for 50 and 500 MW power plants will be given. This analysis includes core FRC plant scaling, metallic liner recycling, radiation shielding, operations, and facilities capital requirements.

  1. High-Energy Electron Confinement in a Magnetic Cusp Configuration

    Directory of Open Access Journals (Sweden)

    Jaeyoung Park

    2015-06-01

    Full Text Available We report experimental results validating the concept that plasma confinement is enhanced in a magnetic cusp configuration when β (plasma pressure/magnetic field pressure is of order unity. This enhancement is required for a fusion power reactor based on cusp confinement to be feasible. The magnetic cusp configuration possesses a critical advantage: the plasma is stable to large scale perturbations. However, early work indicated that plasma loss rates in a reactor based on a cusp configuration were too large for net power production. Grad and others theorized that at high β a sharp boundary would form between the plasma and the magnetic field, leading to substantially smaller loss rates. While not able to confirm the details of Grad’s work, the current experiment does validate, for the first time, the conjecture that confinement is substantially improved at high β. This represents critical progress toward an understanding of the plasma dynamics in a high-β cusp system. We hope that these results will stimulate a renewed interest in the cusp configuration as a fusion confinement candidate. In addition, the enhanced high-energy electron confinement resolves a key impediment to progress of the Polywell fusion concept, which combines a high-β cusp configuration with electrostatic fusion for a compact, power-producing nuclear fusion reactor.

  2. On the physical conditions for arising a controlled fusion chain reaction supported by neutrons in fusion facilities with magnetic plasma confinement

    Directory of Open Access Journals (Sweden)

    A.N. Shmelyov

    2015-11-01

    The fusion neutron source is considered to be the “richest”: neutron generation is accompanied by relatively small-scale processes. The thermonuclear facility with low neutron absorption blanket under consideration here could create a high density neutron flux in the blanket. It can be concluded from the above that such thermonuclear facilities could be used for fast transmutation of long-lived fission products with low neutron absorption cross-section, and perhaps even without their preliminary isotopic separation.

  3. Extreme laser pulses for possible development of boron fusion power reactors for clean and lasting energy

    Science.gov (United States)

    Hora, H.; Eliezer, S.; Kirchhoff, G. J.; Korn, G.; Lalousis, P.; Miley, G. H.; Moustaizis, S.

    2017-05-01

    The nuclear reaction of hydrogen (protons) with the boron isotope 11 (HB11) is aneutronic avoiding the production of dangerous neutrons in contrast to any other fusion but it is extremely difficult at thermal equilibrium plasma conditions. There are alternative schemes without thermal equilibrium, e.g. the Tri Alpha reversed magnetic field (RMF) confinement and others, however, the only historical first measurements of HB11 fusion were with lasers interacting with high density plasmas using non-thermal direct conversion of laser energy into ultrahigh acceleration of plasma blocks to avoid the thermal problems. Combining these long studied mechanisms with recently measured ultrahigh magnetic fields for trapping the reacting plasma arrives at a very compact design of an environmentally clean reactor for profitable low cost energy using present technologies.

  4. Z-inertial fusion energy: power plant final report FY 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark (University of Wisconsin, Madison, WI); Kulcinski, Gerald (University of Wisconsin, Madison, WI); Zhao, Haihua (University of California, Berkeley, CA); Cipiti, Benjamin B.; Olson, Craig Lee; Sierra, Dannelle P.; Meier, Wayne (Lawrence Livermore National Laboratories); McConnell, Paul E.; Ghiaasiaan, M. (Georgia Institute of Technology, Atlanta, GA); Kern, Brian (Georgia Institute of Technology, Atlanta, GA); Tajima, Yu (University of California, Los Angeles, CA); Campen, Chistopher (University of California, Berkeley, CA); Sketchley, Tomas (University of California, Los Angeles, CA); Moir, R (Lawrence Livermore National Laboratories); Bardet, Philippe M. (University of California, Berkeley, CA); Durbin, Samuel; Morrow, Charles W.; Vigil, Virginia L (University of Wisconsin, Madison, WI); Modesto-Beato, Marcos A.; Franklin, James Kenneth (University of California, Berkeley, CA); Smith, James Dean; Ying, Alice (University of California, Los Angeles, CA); Cook, Jason T.; Schmitz, Lothar (University of California, Los Angeles, CA); Abdel-Khalik, S. (Georgia Institute of Technology, Atlanta, GA); Farnum, Cathy Ottinger; Abdou, Mohamed A. (University of California, Los Angeles, CA); Bonazza, Riccardo (University of Wisconsin, Madison, WI); Rodriguez, Salvador B.; Sridharan, Kumar (University of Wisconsin, Madison, WI); Rochau, Gary Eugene; Gudmundson, Jesse (University of Wisconsin, Madison, WI); Peterson, Per F. (University of California, Berkeley, CA); Marriott, Ed (University of Wisconsin, Madison, WI); Oakley, Jason (University of Wisconsin, Madison, WI)

    2006-10-01

    This report summarizes the work conducted for the Z-inertial fusion energy (Z-IFE) late start Laboratory Directed Research Project. A major area of focus was on creating a roadmap to a z-pinch driven fusion power plant. The roadmap ties ZIFE into the Global Nuclear Energy Partnership (GNEP) initiative through the use of high energy fusion neutrons to burn the actinides of spent fuel waste. Transmutation presents a near term use for Z-IFE technology and will aid in paving the path to fusion energy. The work this year continued to develop the science and engineering needed to support the Z-IFE roadmap. This included plant system and driver cost estimates, recyclable transmission line studies, flibe characterization, reaction chamber design, and shock mitigation techniques.

  5. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    Science.gov (United States)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  6. Computational modeling of magentically driven liner-on-plasma fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

    Magnetized Target Fusion (MTF) is an approach to controlled fusion which potentially avoids the difficulties of the traditional magnetic and inertial confinement approaches. It appears possible to investigate the critical issues for MTF at low cost, relative to traditional fusion programs, utilizing pulsed power drivers much less expensive than ICF drivers, and plasma configurations much less expensive than those needed for full magnetic confinement. Computational and experimental research into MTF is proceeding at Los Alamos, VNIIEF, and other laboratories.

  7. Development of heat sink concept for near-term fusion power plant divertor

    Science.gov (United States)

    Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

    2017-04-01

    Development of an efficient divertor concept is an important task to meet in the scenario of the future fusion power plant. The divertor, which is a vital part of the reactor has to discharge the considerable fraction of the total fusion thermal power (∼15%). Therefore, it has to survive very high thermal fluxes (∼10 MW/m2). In the present paper, an efficient divertor heat exchanger cooled by helium is proposed for the fusion tokamak. The Plasma facing surface of divertor made-up of several modules to overcome the stresses caused by high heat flux. The thermal hydraulic performance of one such module is numerically investigated in the present work. The result shows that the proposed design is capable of handling target heat flux values of 10 MW/m2. The computational model has been validated against high-heat flux experiments and a satisfactory agreement is noticed between the present simulation and the reported results.

  8. A Multifeature Fusion Approach for Power System Transient Stability Assessment Using PMU Data

    Directory of Open Access Journals (Sweden)

    Yang Li

    2015-01-01

    Full Text Available Taking full advantage of synchrophasors provided by GPS-based wide-area measurement system (WAMS, a novel VBpMKL-based transient stability assessment (TSA method through multifeature fusion is proposed in this paper. First, a group of classification features reflecting the transient stability characteristics of power systems are extracted from synchrophasors, and according to the different stages of the disturbance process they are broken into three nonoverlapped subsets; then a VBpMKL-based TSA model is built using multifeature fusion through combining feature spaces corresponding to each feature subset; and finally application of the proposed model to the IEEE 39-bus system and a real-world power system is demonstrated. The novelty of the proposed approach is that it improves the classification accuracy and reliability of TSA using multifeature fusion with synchrophasors. The application results on the test systems verify the effectiveness of the proposal.

  9. A novel method for modeling the neutron time of flight detector response in current mode to inertial confinement fusion experiments (invited).

    Science.gov (United States)

    Nelson, A J; Ruiz, C L; Cooper, G W; Chandler, G A; Fehl, D L; Hahn, K D; Leeper, R J; Smelser, R; Torres, J A

    2012-10-01

    A novel method for modeling the neutron time of flight (nTOF) detector response in current mode for inertial confinement fusion experiments has been applied to the on-axis nTOF detectors located in the basement of the Z-Facility. It will be shown that this method can identify sources of neutron scattering, and is useful for predicting detector responses in future experimental configurations, and for identifying potential sources of neutron scattering when experimental set-ups change. This method can also provide insight on how much broadening neutron scattering contributes to the primary signals, which is then subtracted from them. Detector time responses are deconvolved from the signals, allowing a transformation from dN/dt to dN/dE, extracting neutron spectra at each detector location; these spectra are proportional to the absolute yield.

  10. Saturation of the two-plasmon decay instability in long-scale-length plasmas relevant to direct-drive inertial confinement fusion.

    Science.gov (United States)

    Froula, D H; Yaakobi, B; Hu, S X; Chang, P-Y; Craxton, R S; Edgell, D H; Follett, R; Michel, D T; Myatt, J F; Seka, W; Short, R W; Solodov, A; Stoeckl, C

    2012-04-20

    Measurements of the hot-electron generation by the two-plasmon-decay instability are made in plasmas relevant to direct-drive inertial confinement fusion. Density-scale lengths of 400 μm at n(cr)/4 in planar CH targets allowed the two-plasmon-decay instability to be driven to saturation for vacuum intensities above ~3.5×10(14) W cm(-2). In the saturated regime, ~1% of the laser energy is converted to hot electrons. The hot-electron temperature is measured to increase rapidly from 25 to 90 keV as the laser beam intensity is increased from 2 to 7×10(14) W cm(-2). This increase in the hot-electron temperature is compared with predictions from nonlinear Zakharov models.

  11. Saturation of the Two-Plasmon Decay Instability in Long-Scale-Length Plasmas Relevant to Direct-Drive Inertial Confinement Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Froula, D. H.; Yaakobi, B.; Hu, S. X.; Chang, P-Y.; Craxton, R. S.; Edgell, D. H.; Follett, R.; Michel, D. T.; Myatt, J. F.; Seka, W.; Short, R. W.; Solodov, A.; Stoeckl, C.

    2012-04-01

    Measurements of the hot-electron generation by the two-plasmon-decay instability are made in plasmas relevant to direct-drive inertial confinement fusion. Density-scale lengths of 400 {micro}m at n{sub cr}/4 in planar CH targets allowed the two-plasmon-decay instability to be driven to saturation for vacuum intensities above ~3.5 x 10{sup 14} W cm{sup -2}. In the saturated regime, ~1% of the laser energy is converted to hot electrons. The hot-electron temperature is measured to increase rapidly from 25 to 90 keV as the laser beam intensity is increased from 2 to 7 x 10{sup 14} W cm{sup -2}. This increase in the hot-electron temperature is compared with predictions from nonlinear Zakharov models.

  12. On_the_implementation_of_the_conditions_of_Inertial_Confinement_ Fusion by bombarding the target a macro particle

    CERN Document Server

    Dolya, S N

    2013-01-01

    The acceleration of lithium tube segments with the length one centimeter, diameter sixteen microns wall thickness one nanometer is considered. These segments are electrically charged by proton beams produced by an electron beam source. Then, they are accelerated by the traveling wave field in a spiral waveguide. The segments are next sent to a target where they are compressed by three hundred times in the longitudinal direction and compressing target radially, so the conditions for thermonuclear fusion are realized.

  13. Discourse, Power, and Knowledge in the Management of "Big Science": The Production of Consensus in a Nuclear Fusion Research Laboratory.

    Science.gov (United States)

    Kinsella, William J.

    1999-01-01

    Extends a Foucauldian view of power/knowledge to the archetypical knowledge-intensive organization, the scientific research laboratory. Describes the discursive production of power/knowledge at the "big science" laboratory conducting nuclear fusion research and illuminates a critical incident in which the fusion research…

  14. The Status of Research Regarding Magnetic Mirrors as a Fusion Neutron Source or Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Simonen, T

    2008-12-23

    experiments have confirmed the physics of effluent plasma stabilization predicted by theory. The plasma had a mean ion energy of 10 keV and a density of 5e19m-3. If successful, the axisymmetric tandem mirror extension of the GDT idea could lead to a Q {approx} 10 power plant of modest size and would yield important applications at lower Q. In addition to the GDT method, there are four other ways to augment stability that have been demonstrated; including: plasma rotation (MCX), diverter coils (Tara), pondermotive (Phaedrus & Tara), and end wall funnel shape (Nizhni Novgorod). There are also 5 stabilization techniques predicted, but not yet demonstrated: expander kinetic pressure (KSTM-Post), Pulsed ECH Dynamic Stabilization (Post), wall stabilization (Berk), non-paraxial end mirrors (Ryutov), and cusp ends (Kesner). While these options should be examined further together with conceptual engineering designs. Physics issues that need further analysis include: electron confinement, MHD and trapped particle modes, analysis of micro stability, radial transport, evaluation and optimization of Q, and the plasma density needed to bridge to the expansion-region. While promising all should be examined through increased theory effort, university-scale experiments, and through increased international collaboration with the substantial facilities in Russia and Japan The conventional wisdom of magnetic mirrors was that they would never work as a fusion concept for a number of reasons. This conventional wisdom is most probably all wrong or not applicable, especially for applications such as low Q (DT Neutron Source) aimed at materials testing or for a Q {approx} 3-5 fusion neutron source applied to destroying actinides in fission waste and breeding of fissile fuel.

  15. Propulsion and Power Generation Capabilities of a Dense Plasma Focus (DPF) Fusion System for Future Military Aerospace Vehicles (POSTPRINT)

    National Research Council Canada - National Science Library

    Knecht, Sean D; Mead, Franklin B; Thomas, Robert E; Miley, George H; Froning, David

    2005-01-01

    ...) fusion power and propulsion technology, with advanced "waverider"-like airframe configurations utilizing air-breathing MHD propulsion and power technology within a reusable single-stage-to-orbit vehicle...

  16. Liquid metals as alternative solution for the power exhaust of future fusion devices: status and perspective

    NARCIS (Netherlands)

    Coenen, J. W.; De Temmerman, G.; G Federici,; Philipps, V.; Sergienko, G.; G Strohmayer,; Terra, A.; Unterberg, B.; Wegener, T.; van den Bekerom, D. C. M.

    2014-01-01

    Applying liquid metals as plasma facing components for fusion power-exhaust can potentially ameliorate lifetime issues as well as limitations to the maximum allowed surface heat loads by allowing for a more direct contact with the coolant. The material choice has so far been focused on lithium (Li),

  17. Fusion plasma physics

    CERN Document Server

    Stacey, Weston M

    2012-01-01

    This revised and enlarged second edition of the popular textbook and reference contains comprehensive treatments of both the established foundations of magnetic fusion plasma physics and of the newly developing areas of active research. It concludes with a look ahead to fusion power reactors of the future. The well-established topics of fusion plasma physics -- basic plasma phenomena, Coulomb scattering, drifts of charged particles in magnetic and electric fields, plasma confinement by magnetic fields, kinetic and fluid collective plasma theories, plasma equilibria and flux surface geometry, plasma waves and instabilities, classical and neoclassical transport, plasma-materials interactions, radiation, etc. -- are fully developed from first principles through to the computational models employed in modern plasma physics. The new and emerging topics of fusion plasma physics research -- fluctuation-driven plasma transport and gyrokinetic/gyrofluid computational methodology, the physics of the divertor, neutral ...

  18. Solar PV Power Forecasting Using Extreme Learning Machine and Information Fusion

    OpenAIRE

    Le Cadre, Hélène; Aravena, Ignacio; Papavasiliou, Anthony

    2015-01-01

    International audience; We provide a learning algorithm combining distributed Extreme Learning Machine and an information fusion rule based on the ag-gregation of experts advice, to build day ahead probabilistic solar PV power production forecasts. These forecasts use, apart from the current day solar PV power production, local meteorological inputs, the most valuable of which is shown to be precipitation. Experiments are then run in one French region, Provence-Alpes-Côte d'Azur, to evaluate ...

  19. A novel technique for single-shot energy-resolved 2D x-ray imaging of plasmas relevant for the inertial confinement fusion.

    Science.gov (United States)

    Labate, L; Köster, P; Levato, T; Gizzi, L A

    2012-10-01

    A novel x-ray diagnostic of laser-fusion plasmas is described, allowing 2D monochromatic images of hot, dense plasmas to be obtained in any x-ray photon energy range, over a large domain, on a single-shot basis. The device (named energy-encoded pinhole camera) is based upon the use of an array of many pinholes coupled to a large area CCD camera operating in the single-photon mode. The available x-ray spectral domain is only limited by the quantum efficiency of scientific-grade x-ray CCD cameras, thus extending from a few keV up to a few tens of keV. Spectral 2D images of the emitting plasma can be obtained at any x-ray photon energy provided that a sufficient number of photons had been collected at the desired energy. Results from recent inertial confinement fusion related experiments will be reported in order to detail the new diagnostic.

  20. Fusion physics

    CERN Document Server

    Lackner, Karl; Tran, Minh Quang

    2012-01-01

    This publication is a comprehensive reference for graduate students and an invaluable guide for more experienced researchers. It provides an introduction to nuclear fusion and its status and prospects, and features specialized chapters written by leaders in the field, presenting the main research and development concepts in fusion physics. It starts with an introduction to the case for the development of fusion as an energy source. Magnetic and inertial confinement are addressed. Dedicated chapters focus on the physics of confinement, the equilibrium and stability of tokamaks, diagnostics, heating and current drive by neutral beam and radiofrequency waves, and plasma–wall interactions. While the tokamak is a leading concept for the realization of fusion, other concepts (helical confinement and, in a broader sense, other magnetic and inertial configurations) are also addressed in the book. At over 1100 pages, this publication provides an unparalleled resource for fusion physicists and engineers.

  1. Low-to-high confinement transition mediated by turbulence radial wave number spectral shift in a fusion plasma

    DEFF Research Database (Denmark)

    Xu, G. S.; Wan, B. N.; Wang, H. Q.

    2016-01-01

    A new model for the low-to-high (L-H) confinement transition has been developed based on a new paradigm for turbulence suppression by velocity shear [G. M. Staebler et al., Phys. Rev. Lett.110, 055003 (2013)]. The model indicates that the L-H transition can be mediated by a shift in the radial wave...... number spectrum of turbulence, as evidenced here, for the first time, by the direct observation of a turbulence radial wave number spectral shift and turbulence structure tilting prior to the L-H transition at tokamak edge by direct probing. This new mechanism does not require a pretransition overshoot...

  2. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    This volume contains chapters on each of the following topics: (1) radioactivity, (2) heat transport and energy conversion, (3) tritium systems, (4) electrical storage and power supplies, (5) support structure, (6) cryogenics, (7) instrumentation and control, (8) maintenance and operation, (9) balance of plant design, (10) safety and environmental analysis, (11) economic analysis, and (12) plant construction.

  3. An experimental investigation of stimulated Brillouin scattering in laser-produced plasmas relevant to inertial confinement fusion

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, Keith Stanley [Univ. of California, Davis, CA (United States)

    1993-02-11

    Despite the apparent simplicity of controlled fusion, there are many phenomena which have prevented its achievement. One phenomenon is laser-plasma instabilities. An investigation of one such instability, stimulated Brillouin scattering (SBS), is reported here. SBS is a parametric process whereby an electromagnetic wave (the parent wave) decays into another electromagnetic wave and an ion acoustic wave (the daughter waves). SBS impedes controlled fusion since it can scatter much or all of the incident laser light, resulting in poor drive symmetry and inefficient laser-plasma coupling. It is widely believed that SBS becomes convectively unstable--that is, it grows as it traverses the plasma. Though it has yet to be definitively tested, convective theory is often invoked to explain experimental observations, even when one or more of the theory`s assumptions are violated. In contrast, the experiments reported here not only obeyed the assumptions of the theory, but were also conducted in plasmas with peak densities well below quarter-critical density. This prevented other competing or coexisting phenomena from occurring, thereby providing clearly interpretable results. These are the first SBS experiments that were designed to be both a clear test of linear convective theory and pertinent to controlled fusion research. A crucial part of this series of experiments was the development of a new instrument, the Multiple Angle Time Resolving Spectrometer (MATRS). MATRS has the unique capability of both spectrally and temporally resolving absolute levels of scattered light at many angles simultaneously, and is the first of its kind used in laser-plasma experiments. A detailed comparison of the theoretical predictions and the experimental observations is made.

  4. Confinement and heating of a deuterium-tritium plasma

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Ashcroft, D.; Barnes, C.W.; Barnes, G.; Batha, S.; Bell, M.G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N.L.; Budny, R.; Bush, C.E.; Camp, R.; Caorlin, M.; Cauffman, S.; Chang, Z.; Cheng, C.Z.; Collins, J.; Coward, G.; Darrow, D.S.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Fisher, R.; Fonck, R.J.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Gentile, C.; Gorelenkov, N.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Herrmann, H.W.; Hill, K.W.; Hosea, J.; Hsuan, H.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kugel, H.; Lam, N.T.; LaMarche, P.H.; Loughlin, M.J.; LeBlanc, B.; Leonard, M.; Levinton, F.M.; Machuzak, J.; Mansfield, D.K.; Martin, A.; Mazzucato, E.; Majeski, R.; Marmar, E.; McChesney, J.; McCormack, B.; McCune, D.C.; McGuire, K.M.; McKee, G.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Muelle

    1994-05-30

    The Tomamak Fusion Test reactor has performed initial high-power experiments with the plasma fueled with nominally equal densities of deuterium and tritium. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by [similar to]20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by [alpha] particles created by the D-T fusion reactions.

  5. CORRIGENDUM: Main safety issues at the transition from ITER to fusion power plants

    Science.gov (United States)

    Gulden, W.; Ciattaglia, S.; Massaut, V.; Sardain, P.

    2007-09-01

    In parallel to the ITER design process and in close cooperation with the designers a fusion-specific safety approach was developed and implemented. Detailed safety assessments have been performed and documented in the ITER Generic Site Safety Report (GSSR). Following the decision on ITER construction in France, results from the GSSR and from on-going safety-related activities tailored to the Cadarache site and the French licensing process are now being used to write the ITER Preliminary Safety Analysis Report. In the most recent European fusion power plant conceptual study (PPCS) inherent fusion favourable features have been exploited, by appropriate design and choice of materials, to provide major safety and environmental advantages. The study focused on five power plant models, which are illustrative of a wider spectrum of possibilities. These span a range from relatively near-term concepts, based on limited technology and plasma physics extrapolations, to a more advanced conception. All five PPCS plant models differ substantially in their plasma physics, blanket and divertor technology, size, fusion power and materials compositions, and these differences lead to differences in economic performance and in the details of safety and environmental impacts. This paper uses the quite detailed information available from ITER safety documents and highlights the differences between ITER and future fusion power plants. The main areas investigated are releases and doses during normal operation and under accidental conditions, occupational radiation exposure and optimization and waste management, including recycling and/or final disposal in repositories. Due to an error, an incorrect version of this paper was published in issue 7. For the convenience of the reader we have included the correct full article below rather than a list of changes.

  6. Analysing the role of fusion power in the future global energy system

    Directory of Open Access Journals (Sweden)

    Grohnheit P.E.

    2012-10-01

    Full Text Available This work presents the EFDA Times model (ETM, developed within the European Fusion Development Agreement (EFDA. ETM is an optimization global energy model which aims at providing the optimum energy system composition in terms of social wealth and sustainability including fusion as an alternative technology in the long term. Two framework scenarios are defined: a Base case scenario with no limits to CO2 emissions, and a 450ppm scenario with a limit of 450ppm in CO2-eq concentrations set by 2100. Previous results showed that in the Base case scenario, with no measures for CO2 emission reductions, fusion does not enter the energy system. However, when CO2 emission restrictions are imposed, the global energy system composition changes completely. In a 450ppm scenario, coal technologies disappear in a few decades, being mainly replaced by nuclear fission technologies which experience a great increase when constrained only by Uranium resources exhaustion. Fission technologies are then replaced by the fusion power plants that start in 2070, with a significant contribution to the global electricity production by 2100. To conclude the work, a sensitivity analysis will be presented on some parameters that may affect the possible role of fusion in the future global energy system.

  7. Analysing the role of fusion power in the future global energy system

    Science.gov (United States)

    Cabal, H.; Lechón, Y.; Ciorba, U.; Gracceva, F.; Eder, T.; Hamacher, T.; Lehtila, A.; Biberacher, M.; Grohnheit, P. E.; Ward, D.; Han, W.; Eherer, C.; Pina, A.

    2012-10-01

    This work presents the EFDA Times model (ETM), developed within the European Fusion Development Agreement (EFDA). ETM is an optimization global energy model which aims at providing the optimum energy system composition in terms of social wealth and sustainability including fusion as an alternative technology in the long term. Two framework scenarios are defined: a Base case scenario with no limits to CO2 emissions, and a 450ppm scenario with a limit of 450ppm in CO2-eq concentrations set by 2100. Previous results showed that in the Base case scenario, with no measures for CO2 emission reductions, fusion does not enter the energy system. However, when CO2 emission restrictions are imposed, the global energy system composition changes completely. In a 450ppm scenario, coal technologies disappear in a few decades, being mainly replaced by nuclear fission technologies which experience a great increase when constrained only by Uranium resources exhaustion. Fission technologies are then replaced by the fusion power plants that start in 2070, with a significant contribution to the global electricity production by 2100. To conclude the work, a sensitivity analysis will be presented on some parameters that may affect the possible role of fusion in the future global energy system. Note to the reader: The article number has been corrected on web pages on November 22, 2013.

  8. Conceptual Design of Low Fusion Power Hybrid System for Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    DRUP (Direct Reuse of Used PWR) fuel has same process with DUPIC (Direct Use of spent PWR fuel Into CANDU reactor). There are 2 big benefits by using DRUP fuel in Hybrid system. One is fissile production during operating period. Required power is decreased by fissile production from DRUP fuel. When the fusion power is reduced, integrity of structure materials is not significantly weakened due to reduction of 14.1MeV high energy neutrons. In addition, required amount of tritium for self-sufficiency TBR (Tritium Breeding Ratio ≥ 1.1) is decreased. Therefore, it is possible to further loading the SNF as much as the amount of lithium decreased. It is effective in transmutation. The other one is that DRUP fuel is also SNF. Therefore, using DRUP fuel is reusing of SNF, as a result it makes reduction of SNF from PWR. However, thermal neutron system is suitable for using DRUP fuel compared to fast neutron system. Therefore, transmutation zone designed (U-TRU)Zr fuel and fissile production zone designed DRUP fuel are separated in this study. In this paper, using DRUP fuel for low fusion power in hybrid system is suggested. Fusion power is decreased by using DRUP fuel. As a result, TBR is satisfied design condition despite of using natural lithium. In addition, not only (U-TRU)Zr fuel but also DRUP fuel are transmuted.

  9. Estimation of Total Fusion Reactivity and Contribution from Suprathermal Tail using 3-parameter Dagum Ion Speed Distribution

    CERN Document Server

    Majumdar, Rudrodip

    2016-01-01

    Thermonuclear fusion reactivity is a pivotal quantity in the studies pertaining to fusion energy production, fusion ignition and energy break-even analysis in both inertially and magnetically confined systems. Although nuclear fusion reactivity and thereafter the power density of a magnetic confinement fusion reactor and the fulfillment of the ignition criterion are quantitatively determined by assuming the ion speed distribution to be Maxwellian, a significant population of suprathermal ions,with energy greater than the quasi-Maxwellian background plasma temperature, is generated by the fusion reactions and auxiliary heating in the fusion devices. In the current work 3-parameter Dagum speed distribution has been introduced to include the effect of suprathermal ion population in the calculation of total fusion reactivity. The extent of enhancement in the fusion reactivity, at different back-ground temperatures of the fusion fuel plasma, due to the suprathermal ion population has also been discussed.

  10. Survey of linear magnetic fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A.

    1978-01-01

    The promise and problems of Linear Magnetic Fusion (LMF) for the generation of electrical power are surveyed. A number of axial confinement schemes are described and compared on an n tau basis. Likewise, the range of heating methods is described. The results of seven conceptual LMF reactor design studies are summarized with an emphasis on the interfaces between reactor operation, confinement scheme, and heating methods.

  11. Mission to Mars by catalyzed nuclear reactions of the commercialized cold fusion power

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Yonsei University, Wonju (Korea, Republic of)

    2016-05-15

    The chemical compound source is deficient to reach to the power as much as the journey to Mars, unless the massive equipment is installed like the nuclear fusion reactor. However, there is very significant limitations of making up the facility due to the propellant power. Therefore, the light and cheap energy source, Low energy nuclear reactions (LENRs), powered rocket has been proposed. In this paper, the power conditions by LENRs are analyzed. After the successful Apollo mission to Moon of the National Aeronautics and Space Administration (NASA) in the U.S. government, the civilian companies have proposed for the manned mission to Mars for the commercial journey purposes. The nuclear power has been a critical issue for the energy source in the travel, especially, by the LENR of LENUCO, Champaign, USA. As the velocity of the rocket increases, the mass flow rate decreases. It could be imaginable to take the reasonable velocity of spacecraft. The energy of the travel system is and will be created for the better one in economical and safe method. There is the imagination of boarding pass for spacecraft ticket shows the selected companies of cold fusion products. In order to solve the limitations of the conventional power sources like the chemical and solar energies, it is reasonable to design LENR concept. Since the economical and safe spacecraft is very important in the long journey on and beyond the Mars orbit, a new energy source, LENR, should be studied much more.

  12. Biological effects of activation products and other chemicals released from fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.A.; Poston, T.M.

    1976-09-01

    Literature reviews indicate that existing information is incomplete, often contradictory, and of questionable value for the prediction and assessment of ultimate impact from fusion-associated activation products and other chemical releases. It is still uncertain which structural materials will be used in the blanket and first wall of fusion power plants. However, niobium, vanadium, vanadium-chromium alloy, vanadium-titanium alloy, sintered aluminum product, and stainless steel have been suggested. The activation products of principal concern will be the longer-lived isotopes of /sup 26/Al, /sup 49/V, /sup 51/Cr, /sup 54/Mn, /sup 55/Fe, /sup 58/Co, /sup 60/Co, /sup 93/Nb, and /sup 94/Nb. Lithium released to the environment either during the mining cycle, from power plant operation or accident, may be in the form of a number of compound types varying in solubility and affinity for biological organisms. The effects of a severe liquid metal fire or explosion involving Na or K will vary according to inherent abiotic and biotic features of the affected site. Saline, saline-alkaline, and sodic soils of arid lands would be particularly susceptible to alkaline stress. Beryllium released to the environment during the mining cycle or reactor accident situation could be in the form of a number of compound types. Adverse effects to aquatic species from routine chemical releases (biocides, corrosion inhibitors, dissolution products) may occur in the discharge of both fission and fusion power plant designs.

  13. Strong doping of the n-optical confinement layer for increasing output power of high- power pulsed laser diodes in the eye safe wavelength range

    Science.gov (United States)

    Ryvkin, Boris S.; Avrutin, Eugene A.; Kostamovaara, Juha T.

    2017-12-01

    An analytical model for internal optical losses at high power in a 1.5 μm laser diode with strong n-doping in the n-side of the optical confinement layer is created. The model includes intervalence band absorption by holes supplied by both current flow and two-photon absorption (TPA), as well as the direct TPA effect. The resulting losses are compared with those in an identical structure with a weakly doped waveguide, and shown to be substantially lower, resulting in a significant improvement in the output power and efficiency in the structure with a strongly doped waveguide.

  14. Anomalous neutron yield in indirect-drive inertial-confinement-fusion due to the formation of collisionless shocks in the corona

    Science.gov (United States)

    Zhang, Wen-Shuai; Cai, Hong-Bo; Shan, Lian-Qiang; Zhang, Hua-Sen; Gu, Yu-Qiu; Zhu, Shao-Ping

    2017-06-01

    Observations of anomalous neutron yield in the indirect-drive inertial confinement fusion implosion experiments conducted at SG-III prototype and SG-II upgrade laser facilities are interpreted. The anomalous mechanism results in a neutron yield which is 100-times higher than that predicted by 1D radiation-hydrodynamic simulations. 2D radiation-hydrodynamic simulations show that the supersonic, radially directed gold (Au) plasma jets arising from the laser-hohlraum interactions can collide with the carbon-deuterium (CD) corona plasma of the compressed pellet. It is found that in the interaction front of the high-Z jet with the low-Z corona, with low density  ∼{{10}20}~\\text{c}{{\\text{m}}-3} and high temperature  ∼keV, kinetic effects become important. Particle-in-cell simulations indicate that an electrostatic shock wave can be driven when the high-temperature Au jet expands into the low-temperature CD corona. Deuterium ions with an amount of  ∼1015 can be accelerated to  ∼25 keV by the collisionless shock wave, thus causing efficient neutron productions though the beam-target method by stopping these energetic ions in the corona. The evaluated neutron yield is consistent with the experiments conducted at SG laser facilities.

  15. Using Wave-Based Cross-Beam Energy Transfer Simulations to Improve the Ray-Based Models Used in Inertial Confinement Fusion Applications

    Science.gov (United States)

    Follett, R. K.; Edgell, D. H.; Froula, D. H.; Goncharov, V. N.; Igumenshchev, I. V.; Shaw, J. G.; Myatt, J. F.

    2017-10-01

    Ray-based models of cross-beam energy transfer (CBET) are used in radiation-hydrodynamics codes to calculate laser-energy deposition for inertial confinement fusion (ICF) experiments. In direct-drive ICF, calculations suggest that CBET is responsible for a 10% to 20% reduction in laser energy absorption. In indirect drive, ray-based calculations predict full pump depletion of the outer cone beams. Ray-based CBET models require artificial limiters to give quantitative agreement with experimental observables. The recent development of a 3-D wave-based solver (LPSE CBET) that does not rely on the paraxial or eikonal approximations allows the limitations of ray-based CBET models to be studied at conditions relevant to laser-driven ICF. The accuracy of ray-based CBET models is limited by uncertainties in the approximations used to account for the experimental realities of beam speckle, polarization smoothing, and interactions at caustics. A physics-based technique is proposed for including the effect of beam speckle in existing ray-based models that gives excellent agreement with the wave-based calculations. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  16. Development of processing and analysis techniques for CR-39-based proton detectors for Inertial Confinement Fusion experiments at the NIF and OMEGA

    Science.gov (United States)

    Manzin, M.; Lahmann, B.; Birkel, A.; Doeg, E.; Frankel, R.; Kabadi, N.; Parker, C. E.; Simpson, R. A.; Sio, H. W.; Sutcliffe, G. D.; Frenje, J. A.; Gatu Johnson, M.; Li, C. K.; Seguin, F. H.; Petrasso, R. D.

    2017-10-01

    CR-39 is a clear plastic polymer used for particle detection in several charged-particle spectrometers, including Step Range Filters (SRF) and the Magnetic Recoil Spectrometer (MRS), at the NIF and OMEGA Inertial Confinement Fusion (ICF) laser facilities. SRFs and MRS have been recently used in basic science experiments on the ICF platform, for pobing stellar-nucleosynthesis-relevant nuclear reactions and astrophysical phenomena such as collisionless shocks and turbulent dynamos. These new applications require extensions of established CR-39 analysis techniques. After exposure, the CR-39 is etched in sodium hydroxide to reveal tracks, and it is subsequently analyzed with a microscope. In this poster, two new developments in CR-39 analysis of proton tracks will be described: (1) the detailed mapping of track size and contrast versus proton energy and etch time for protons in the range 0.2-1 MeV, and (2) the extension of the coincidence counting technique for reduction of intrinsic background in CR-39 analysis to protons at energies >4 MeV. The results will be used to extend the range of information that can be obtained from CR-39 data from the NIF and OMEGA charged-particle spectrometers. This work was supported in part by the U.S. DOE.

  17. Measurements of the differential cross sections for the elastic n-3H and n-2H scattering at 14.1 MeV by using an inertial confinement fusion facility.

    Science.gov (United States)

    Frenje, J A; Li, C K; Seguin, F H; Casey, D T; Petrasso, R D; McNabb, D P; Navratil, P; Quaglioni, S; Sangster, T C; Glebov, V Yu; Meyerhofer, D D

    2011-09-16

    For the first time the differential cross section for the elastic neutron-triton (n-(3)H) and neutron-deuteron (n-(2)H) scattering at 14.1 MeV has been measured by using an inertial confinement fusion facility. In these experiments, which were carried out by simultaneously measuring elastically scattered (3)H and (2)H ions from a deuterium-tritium gas-filled inertial confinement fusion capsule implosion, the differential cross section for the elastic n-(3)H scattering was obtained with significantly higher accuracy than achieved in previous accelerator experiments. The results compare well with calculations that combine the resonating-group method with an ab initio no-core shell model, which demonstrate that recent advances in ab initio theory can provide an accurate description of light-ion reactions.

  18. Accident consequences analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Gomez del Rio, J; Sanz, J

    2000-02-23

    Previous studies of the safety and environmental (S and E) aspects of the HYLIFE-II inertial fusion energy (IFE) power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work a set of computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) has been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here the authors consider a severe lost of coolant accident (LOCA) producing simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the containment) and of the two barriers surrounding the chamber (inner shielding and containment building it self). Even though containment failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product release and transport. The results of these calculations show that the estimated off-site dose is less than 6 mSv (0.6 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

  19. Fusion Technology for ITER, the ITER Project. Further Development Towards a DEMO Fusion Power Plant (3/4)

    CERN Document Server

    CERN. Geneva

    2011-01-01

    This is the second half of a lecture series on fusion and will concentrate on fusion technology. The early phase of fusion development was concentrated on physics. However, during the 1980s it was realized that if one wanted to enter the area of fusion reactor plasmas, even in an experimental machine, a significant advance in fusion technologies would be needed. After several conceptual studies of reactor class fusion devices in the 1980s the engineering design phase of ITER started in earnest during the 1990s. The design team was in the beginning confronted with many challenges in the fusion technology area as well as in physics for which no readily available solution existed and in a few cases it was thought that solutions may be impossible to find. However, after the initial 3 years of intensive design and R&D work in an international framework utilizing basic fusion technology R&D from the previous decade it became clear that for all problems a conceptual solution could be found and further devel...

  20. Progress in accident analysis of the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Latkowski, J F; Gomez del Rio, J; Sanz, J

    2000-10-11

    The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident (LOCA) in which all the liquid flibe (Li{sub 2}BeF{sub 4}) was lost at the beginning of the accident. Results showed that the off-site dose was below the limit given by the DOE Fusion Safety Standards for public protection in case of accident, and that his dose was dominated by the tritium released during the accident.

  1. Systems Modeling For The Laser Fusion-Fission Energy (LIFE) Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W R; Abbott, R; Beach, R; Blink, J; Caird, J; Erlandson, A; Farmer, J; Halsey, W; Ladran, T; Latkowski, J; MacIntyre, A; Miles, R; Storm, E

    2008-10-02

    A systems model has been developed for the Laser Inertial Fusion-Fission Energy (LIFE) power plant. It combines cost-performance scaling models for the major subsystems of the plant including the laser, inertial fusion target factory, engine (i.e., the chamber including the fission and tritium breeding blankets), energy conversion systems and balance of plant. The LIFE plant model is being used to evaluate design trade-offs and to identify high-leverage R&D. At this point, we are focused more on doing self consistent design trades and optimization as opposed to trying to predict a cost of electricity with a high degree of certainty. Key results show the advantage of large scale (>1000 MWe) plants and the importance of minimizing the cost of diodes and balance of plant cost.

  2. Magnetic reconnection in plasma under inertial confinement fusion conditions driven by heat flux effects in Ohm's law

    CERN Document Server

    Joglekar, A S; Fox, W; Bhattacharjee, A

    2015-01-01

    In the interaction of high-power laser beams with solid density plasma there are a number of mechanisms that generate strong magnetic fields. Such fields subsequently inhibit or redirect electron flows, but can themselves be advected by heat fluxes, resulting in complex interplay between thermal transport and magnetic fields.We show that for heating by multiple laser spots reconnection of magnetic field lines can occur, mediated by these heat fluxes, using a fully implicit 2D Vlasov-Fokker-Planck code. Under such conditions, the reconnection rate is dictated by heat flows rather than Alfv\\`enic flows. We find that this mechanism is only relevant in a high $\\beta$ plasma. However, the Hall parameter $\\omega_c \\tau_{ei}$ can be large so that thermal transport is strongly modified by these magnetic fields, which can impact longer time scale temperature homogeneity and ion dynamics in the system.

  3. Magnetic reconnection in plasma under inertial confinement fusion conditions driven by heat flux effects in Ohm's law.

    Science.gov (United States)

    Joglekar, A S; Thomas, A G R; Fox, W; Bhattacharjee, A

    2014-03-14

    In the interaction of high-power laser beams with solid density plasma there are a number of mechanisms that generate strong magnetic fields. Such fields subsequently inhibit or redirect electron flows, but can themselves be advected by heat fluxes, resulting in complex interplay between thermal transport and magnetic fields. We show that for heating by multiple laser spots reconnection of magnetic field lines can occur, mediated by these heat fluxes, using a fully implicit 2D Vlasov-Fokker-Planck code. Under such conditions, the reconnection rate is dictated by heat flows rather than Alfvènic flows. We find that this mechanism is only relevant in a high β plasma. However, the Hall parameter ωcτei can be large so that thermal transport is strongly modified by these magnetic fields, which can impact longer time scale temperature homogeneity and ion dynamics in the system.

  4. Method and apparatus to produce high specific impulse and moderate thrust from a fusion-powered rocket engine

    Science.gov (United States)

    Cohen, Samuel A.; Pajer, Gary A.; Paluszek, Michael A.; Razin, Yosef S.

    2017-11-21

    A system and method for producing and controlling high thrust and desirable specific impulse from a continuous fusion reaction is disclosed. The resultant relatively small rocket engine will have lower cost to develop, test, and operate that the prior art, allowing spacecraft missions throughout the planetary system and beyond. The rocket engine method and system includes a reactor chamber and a heating system for heating a stable plasma to produce fusion reactions in the stable plasma. Magnets produce a magnetic field that confines the stable plasma. A fuel injection system and a propellant injection system are included. The propellant injection system injects cold propellant into a gas box at one end of the reactor chamber, where the propellant is ionized into a plasma. The propellant and fusion products are directed out of the reactor chamber through a magnetic nozzle and are detached from the magnetic field lines producing thrust.

  5. Institut de recherche sur la fusion par confinement magnetique. Association EURATOM-CEA. Annual report 2008 (Full Report)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    The 2008 experimental campaign on Tore-Supra has been especially fruitful with the completion of multi-annual programmes bringing important answers, notably on operational aspects of long duration discharges. Concerning JET, particular effort have been devoted to: -) hydrogen isotopes retention studies, -) advanced scenarios studies such as JT60-JET physics identity experiments in ITB regimes, -) the commissioning of the ITER-like ICRH antenna, and -) high power level commissioning of the LHCD antenna and long distance coupling studies. Concerning integrated modelling, the development of the CRONOS code suite has been carried on, with particular emphasis on increasing its reliability and international coverage. CRONOS has been extensively used to investigate steady-state scenarios for ITER and DEMO. Most of the technological developments carried out by the Association concentrates on key domains that apply to ITER: superconducting conductors, actively cooled PFC (plasma facing components), or diagnostics. This document is divided into 5 main section: 1) Tore-Supra status, 2) theory and modeling, 3) Experimental physics, 4) technological developments, and 5) participation to W7X construction

  6. Fluid mechanics of fusion lasers. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Shwartx, J.; Golik, R.J.; Merkle, C.L.; Ausherman, D.R.; Fishman, E.

    1978-04-01

    The primary objective of this study is to define the fluid mechanical requirements for a repetitively-pulsed high energy laser that may be used as a driver in an inertial confinement fusion system designed for electric power generation. Emphasis was placed on defining conceptual designs of efficient laser flow systems that are capable of conserving gas and minimizing operating power requirements. The development of effective pressure wave suppression concepts to produce acceptable beam quality for fusion applications was also considered.

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

    Science.gov (United States)

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

    1992-03-01

    The Compact Torus Accelerator (CTA), under development at Lawrence Livermore National Laboratory, offers the promise of a low cost, high efficiency, high energy, high power density driver for ICF and MICF (Magnetically Insulated ICF) type fusion systems. A CTA with 100 MJ driver capacitor bank energy is predicted to deliver approximately 30 MJ Compact Torus (CT) kinetic energy to a 1 cm(exp 2) target in several nanoseconds for a power density of approximately 10(exp 16) watts/cm(exp 2). The estimated cost of delivered energy is approximately 3$/Joule, or $100 million for 30 MJ. This driver appears to be cost effective and, in this regard, is virtually alone among IFE drivers. We discuss indirect drive ICF with a DT fusion energy gain Q = 70 for a total yield of 2 GJ. The CT can be guided to the target inside a several meter long disposable cone made of frozen Li2BeF4, the same material as the coolant. We have designed a power plant including CT injection, target emplacement, containment, energy recovery, and tritium breeding. The cost of electricity is predicted to be 4.8 (cents)/kWh, which is competitive with future coal and nuclear costs.

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

    Science.gov (United States)

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

    2014-10-01

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

  9. Masked-backlighter technique used to simultaneously image x-ray absorption and x-ray emission from an inertial confinement fusion plasma.

    Science.gov (United States)

    Marshall, F J; Radha, P B

    2014-11-01

    A method to simultaneously image both the absorption and the self-emission of an imploding inertial confinement fusion plasma has been demonstrated on the OMEGA Laser System. The technique involves the use of a high-Z backlighter, half of which is covered with a low-Z material, and a high-speed x-ray framing camera aligned to capture images backlit by this masked backlighter. Two strips of the four-strip framing camera record images backlit by the high-Z portion of the backlighter, while the other two strips record images aligned with the low-Z portion of the backlighter. The emission from the low-Z material is effectively eliminated by a high-Z filter positioned in front of the framing camera, limiting the detected backlighter emission to that of the principal emission line of the high-Z material. As a result, half of the images are of self-emission from the plasma and the other half are of self-emission plus the backlighter. The advantage of this technique is that the self-emission simultaneous with backlighter absorption is independently measured from a nearby direction. The absorption occurs only in the high-Z backlit frames and is either spatially separated from the emission or the self-emission is suppressed by filtering, or by using a backlighter much brighter than the self-emission, or by subtraction. The masked-backlighter technique has been used on the OMEGA Laser System to simultaneously measure the emission profiles and the absorption profiles of polar-driven implosions.

  10. Measurement and mitigation of X-ray shadow imprint of hydrodynamic instabilities on the surface of Inertial Confinement Fusion capsules due to the fill tube

    Science.gov (United States)

    Macphee, Andrew

    2017-10-01

    Indirectly-driven Inertial Confinement Fusion (ICF) implosions on the National Ignition Facility (NIF) employ a small diameter (10 μm) fill tube to supply the cryogenic deuterium-tritium (DT) fuel to the capsule. Recent experimental observations characterizing the perturbation produced by this fill tube have revealed an unexpected shadow imprinted instability mechanism, whereby several of the x-ray spots formed on the inside wall of the hohlraum cast directional shadows of the fill tube onto the surface of the capsule. Reduced ablation in the corresponding umbrae of these shadows leads to a pattern of radial ridges of excess ablator material measuring 100 nm above the surrounding capsule surface. By the time the capsule has converged 2x from its original radius, the areal density (ρR) perturbation of these spoke-like features becomes comparable to that of central hole due to the fill tube itself. We report both quantitative radiographic measurements of this newly observed perturbation (for several ablator materials) as well as the results of two strategies for mitigating against such shadow imprinted instabilities: 1.) reducing the fill tube diameter and wall thickness to produce a smaller perturbation that blows down to low density more quickly, and 2.) modifying the driving laser pulse for the lower-intensity inner beams to allow more time for the fill tube to blow down to low density prior to the onset of shadow imprint, which is produced by the more-intense outer beams during the later part of the drive. Results and analysis from both focused radiographic experiments as well as the impact on the performance of layered DT ignition implosions will be discussed. Work performed under the auspices of the U.S. D.O.E. by Lawrence Livermore National Laboratory under Contract No. DE-AC52-07NA27344.

  11. Unifying Role of Radial Electric Field Shear in the Confinement Trends of TFTR Supershot Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, D.R.; Scott, S.D. [Princeton University Plasma Physics Laboratory, P.O.B. 451, Princeton, New Jersey 08543 (United States); Coppi, B.; Porkolab, M. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    1998-09-01

    A model is presented to explain the favorable ion thermal confinement trends of supershot plasmas in the Tokamak Fusion Test Reactor (TFTR). Turbulence suppression by radial electric field shear is important to reproduce the measured temperatures. Supershot confinement scalings are reproduced in more than sixty discharges, including favorable core power scaling and variation with isotopic mass, density peakedness, edge recycling, and toroidal rotation. The results connect the transitionless supershot regime with improved confinement regimes which are attained through sharp confinement transitions in the core or edge. {copyright} {ital 1998} {ital The American Physical Society}

  12. New trends in fusion research

    CERN Multimedia

    CERN. Geneva

    2004-01-01

    The efforts of the international fusion community aim at demonstrating the scientific feasibility of thermonuclear fusion energy power plants. Understanding the behavior of burning plasmas, i.e. plasmas with strong self-heating, represents a primary scientific challenge for fusion research and a new science frontier. Although integrated studies will only be possible, in new, dedicated experimental facilities, such as the International Tokamak Experimental Reactor (ITER), present devices can address specific issues in regimes relevant to burning plasmas. Among these are an improvement of plasma performance via a reduction of the energy and particle transport, an optimization of the path to ignition or to sustained burn using additional heating and a control of plasma-wall interaction and energy and particle exhaust. These lectures address recent advances in plasma science and technology that are relevant to the development of fusion energy. Mention will be made of the inertial confinement line of research, but...

  13. Simultaneous catalytic regime of tritium and helium-3 in D–D fusion ...

    Indian Academy of Sciences (India)

    2000). [12] J S Lewis, Space Power 10, 363 (1991). [13] L Wittenberg, J F Santarius and G L Kulcinski, Fusion Technol. 10, 167 (1986). [14] J J Duderstadt and G Moses, Inertial confinement fusion (Wiley, New York, 1981). [15] S Yu Guskov and ...

  14. Characterization of high temperature superconductor cables for magnet toroidal field coils of the DEMO fusion power plant

    CERN Document Server

    Bayer, Christoph M

    2017-01-01

    Nuclear fusion is a key technology to satisfy the basic demand for electric energy sustainably. The official EUROfusion schedule foresees a first industrial DEMOnstration Fusion Power Plant for 2050. In this work several high temperature superconductor sub-size cables are investigated for their applicability in large scale DEMO toroidal field coils. Main focus lies on the electromechanical stability under the influence of high Lorentz forces at peak magnetic fields of up to 12 T.

  15. Characterization of high temperature superconductor cables for magnet toroidal field coils of the DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Bayer, Christoph M.

    2017-05-01

    Nuclear fusion is a key technology to satisfy the basic demand for electric energy sustainably. The official EUROfusion schedule foresees a first industrial DEMOnstration Fusion Power Plant for 2050. In this work several high temperature superconductor sub-size cables are investigated for their applicability in large scale DEMO toroidal field coils. Main focus lies on the electromechanical stability under the influence of high Lorentz forces at peak magnetic fields of up to 12 T.

  16. Research on Reverse Recovery Transient of Parallel Thyristors for Fusion Power Supply

    Science.gov (United States)

    Zha, Fengwei; Song, Zhiquan; Fu, Peng; Dong, Lin; Wang, Min

    2014-07-01

    In nuclear fusion power supply systems, the thyristors often need to be connected in parallel for sustaining large current. However, research on the reverse recovery transient of parallel thyristors has not been reported yet. When several thyristors are connected in parallel, they cannot turn-off at the same moment, and thus the turn-off model based on a single thyristor is no longer suitable. In this paper, an analysis is presented for the reverse recovery transient of parallel thyristors. Parallel thyristors can be assumed as one virtual thyristor so that the reverse recovery current can be modeled by an exponential function. Through equivalent transformation of the rectifier circuit, the commutating over-voltage can be calculated based on Kirchhoff's equation. The reverse recovery current and commutation over-voltage waveforms are measured on an experiment platform for a high power rectifier supply. From the measurement results, it is concluded that the modeling method is acceptable.

  17. Final Focus Shielding Designs for Modern Heavy-Ion Fusion Power Plant Designs

    Energy Technology Data Exchange (ETDEWEB)

    Latkowski, J F; Meier, W R

    2000-07-05

    Recent work in heavy-ion fusion accelerators and final focusing systems shows a trend towards less current per beam, and thus, a greater number of beams. Final focusing magnets are susceptible to nuclear heating, radiation damage, and neutron activation. The trend towards more beams, however, means that there can be less shielding for each magnet, Excessive levels of nuclear heating may lead to magnet quench or an intolerable recirculating power for magnet cooling. High levels of radiation damage may result in short magnet lifetimes and low reliability. Finally, neutron activation of the magnet components may lead to difficulties in maintenance, recycling, and waste disposal. The present work expands upon previous, three-dimensional magnet shielding calculations for a modified version of the HYLIFE-I1 IFE power plant design. We present key magnet results as a function of the number of beams.

  18. Positioning Performance of Power and Manual Drivers in Posterior Spinal Fusion Procedures

    Directory of Open Access Journals (Sweden)

    J. Micah Prendergast

    2017-01-01

    Full Text Available This work presents an analysis and comparison of the efficacy of two methods for pedicle screw placement during posterior spinal fusion surgery. A total of 100 screws (64 manual and 36 power driven, all placed utilizing a surgical navigation system, were analyzed and compared. Final screw placement was compared to initial surgical plans using the navigation system, and the final screw locations were analyzed on the basis of angular deviation from these planned trajectories as well as screw translation within a critical reference plane. The power driver was found to insignificantly decrease the resulting angular deviation of these pedicle screws with a mean deviation of 3.35 degrees compared to 3.44 degrees with the manual driver (p=0.853. Conversely, the power driver was found to increase the translational distance in the critical region, with mean deviations of 2.45 mm for the power driver compared to 1.54 mm with the manual driver. The increase in translational deviation was significant (p=0.002 indicating that there may be some loss in performance from the adoption of the power driver.

  19. Nuclear Fusion Research Understanding Plasma-Surface Interactions

    CERN Document Server

    Clark, Robert E.H

    2005-01-01

    It became clear in the early days of fusion research that the effects of the containment vessel (erosion of "impurities") degrade the overall fusion plasma performance. Progress in controlled nuclear fusion research over the last decade has led to magnetically confined plasmas that, in turn, are sufficiently powerful to damage the vessel structures over its lifetime. This book reviews current understanding and concepts to deal with this remaining critical design issue for fusion reactors. It reviews both progress and open questions, largely in terms of available and sought-after plasma-surface interaction data and atomic/molecular data related to these "plasma edge" issues.

  20. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  1. Body-building without power training : Endogenously regulated pectoral muscle hypertrophy in confined shorebirds

    NARCIS (Netherlands)

    Dietz, MW; Piersma, T; Dekinga, A

    1999-01-01

    Shorebirds such as red knots Calidris canutus routinely make migratory flights of 3000 km or more. Previous studies on this species, based on compositional analyses, suggest extensive pectoral muscle hypertrophy in addition to fat storage before take-off. Such hypertrophy could be due to power

  2. Fusion research principles

    CERN Document Server

    Dolan, Thomas James

    2013-01-01

    Fusion Research, Volume I: Principles provides a general description of the methods and problems of fusion research. The book contains three main parts: Principles, Experiments, and Technology. The Principles part describes the conditions necessary for a fusion reaction, as well as the fundamentals of plasma confinement, heating, and diagnostics. The Experiments part details about forty plasma confinement schemes and experiments. The last part explores various engineering problems associated with reactor design, vacuum and magnet systems, materials, plasma purity, fueling, blankets, neutronics

  3. Low-power resistive random access memory by confining the formation of conducting filaments

    Directory of Open Access Journals (Sweden)

    Yi-Jen Huang

    2016-06-01

    Full Text Available Owing to their small physical size and low power consumption, resistive random access memory (RRAM devices are potential for future memory and logic applications in microelectronics. In this study, a new resistive switching material structure, TiOx/silver nanoparticles/TiOx/AlTiOx, fabricated between the fluorine-doped tin oxide bottom electrode and the indium tin oxide top electrode is demonstrated. The device exhibits excellent memory performances, such as low operation voltage (<±1 V, low operation power, small variation in resistance, reliable data retention, and a large memory window. The current-voltage measurement shows that the conducting mechanism in the device at the high resistance state is via electron hopping between oxygen vacancies in the resistive switching material. When the device is switched to the low resistance state, conducting filaments are formed in the resistive switching material as a result of accumulation of oxygen vacancies. The bottom AlTiOx layer in the device structure limits the formation of conducting filaments; therefore, the current and power consumption of device operation are significantly reduced.

  4. Low-power resistive random access memory by confining the formation of conducting filaments

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Yi-Jen; Lee, Si-Chen, E-mail: sclee@ntu.edu.tw [Graduate Institute of Electronics Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Shen, Tzu-Hsien; Lee, Lan-Hsuan [Department of Materials Science and Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Wen, Cheng-Yen [Department of Materials Science and Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Taiwan Consortium of Emergent Crystalline Materials, Ministry of Science and Technology, Taipei 10617, Taiwan (China)

    2016-06-15

    Owing to their small physical size and low power consumption, resistive random access memory (RRAM) devices are potential for future memory and logic applications in microelectronics. In this study, a new resistive switching material structure, TiO{sub x}/silver nanoparticles/TiO{sub x}/AlTiO{sub x}, fabricated between the fluorine-doped tin oxide bottom electrode and the indium tin oxide top electrode is demonstrated. The device exhibits excellent memory performances, such as low operation voltage (<±1 V), low operation power, small variation in resistance, reliable data retention, and a large memory window. The current-voltage measurement shows that the conducting mechanism in the device at the high resistance state is via electron hopping between oxygen vacancies in the resistive switching material. When the device is switched to the low resistance state, conducting filaments are formed in the resistive switching material as a result of accumulation of oxygen vacancies. The bottom AlTiO{sub x} layer in the device structure limits the formation of conducting filaments; therefore, the current and power consumption of device operation are significantly reduced.

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

    Science.gov (United States)

    Coda, S.; TCV Team

    2015-10-01

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

  6. The Mercury Laser System-A scaleable average-power laser for fusion and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Ebbers, C A; Moses, E I

    2008-03-26

    Nestled in a valley between the whitecaps of the Pacific and the snowcapped crests of the Sierra Nevada, Lawrence Livermore National Laboratory (LLNL) is home to the nearly complete National Ignition Facility (NIF). The purpose of NIF is to create a miniature star-on demand. An enormous amount of laser light energy (1.8 MJ in a pulse that is 20 ns in duration) will be focused into a small gold cylinder approximately the size of a pencil eraser. Centered in the gold cylinder (or hohlraum) will be a nearly perfect sphere filled with a complex mixture of hydrogen gas isotopes that is similar to the atmosphere of our Sun. During experiments, the laser light will hit the inside of the gold cylinder, heating the metal until it emits X-rays (similar to how your electric stove coil emits visible red light when heated). The X-rays will be used to compress the hydrogen-like gas with such pressure that the gas atoms will combine or 'fuse' together, producing the next heavier element (helium) and releasing energy in the form of energetic particles. 2010 will mark the first credible attempt at this world-changing event: the achievement of fusion energy 'break-even' on Earth using NIF, the world's largest laser! NIF is anticipated to eventually perform this immense technological accomplishment once per week, with the capability of firing up to six shots per day - eliminating the need for continued underground testing of our nation's nuclear stockpile, in addition to opening up new realms of science. But what about the day after NIF achieves ignition? Although NIF will achieve fusion energy break-even and gain, the facility is not designed to harness the enormous potential of fusion for energy generation. A fusion power plant, as opposed to a world-class engineering research facility, would require that the laser deliver drive pulses nearly 100,000 times more frequently - a rate closer to 10 shots per second as opposed to several shots per day.

  7. Gas Transport and Control in Thick-Liquid Inertial Fusion PowerPlants

    Energy Technology Data Exchange (ETDEWEB)

    Debonnel, Christophe Sylvain [Univ. of California, Berkeley, CA (United States)

    2006-01-01

    Among the numerous potential routes to a commercial fusion power plant, the inertial path with thick-liquid protection is explored in this doctoral dissertation. Gas dynamics phenomena in such fusion target chambers have been investigated since the early 1990s with the help of a series of simulation codes known as TSUNAMI. For this doctoral work, the code was redesigned and rewritten entirely to enable the use of modern programming techniques, languages and software; improve its user-friendliness; and refine its ability to model thick-liquid protected chambers. The new ablation and gas dynamics code is named “Visual Tsunami” to emphasize its graphics-based pre- and post-processors. It is aimed at providing a versatile and user-friendly design tool for complex systems for which transient gas dynamics phenomena play a key role. Simultaneously, some of these improvements were implemented in a previous version of the code; the resulting code constitutes the version 2.8 of the TSUNAMI series. Visual Tsunami was used to design and model the novel Condensation Debris Experiment (CDE), which presents many aspects of a typical Inertial Fusion Energy (IFE) system and has therefore been used to exercise the code. Numerical and experimental results are in good agreement. In a heavy-ion IFE target chamber, proper beam and target propagation set stringent requirements for the control of ablation debris transport in the target chamber and beam tubes. When the neutralized ballistic transport mode is employed, the background gas density should be adequately low and the beam tube metallic surfaces upstream of the neutralizing region should be free of contaminants. TSUNAMI 2.8 was used for the first simulation of gas transport through the complex geometry of the liquid blanket of a hybrid target chamber and beam lines. Concurrently, the feasibility of controlling the gas density was addressed with a novel beam tube design, which introduces magnetic shutters and a long low

  8. The ab initio equation of state of hydrogen in the warm dense matter and its application to the implosion of targets for the inertial confinement fusion; Equation d'etat ab initio de l'hydrogene dans la matiere dense et tiede et application a l'implosion de cibles pour la fusion par confinement inertiel

    Energy Technology Data Exchange (ETDEWEB)

    Caillabet, L.

    2011-03-25

    In the field of the inertial confinement fusion (ICF), the equation of state (EoS) of the hydrogen and its isotopes is one of the most important properties to know. The EoS based on chemical models have difficulty in giving an unambiguous description of the hydrogen in the strong coupled and partial degenerate regime, called Warm Dense Matter (WDM). Indeed, these models use potential with adjustable parameters to describe the many body interactions which are important in the WDM. On the other hand, the ab initio methods resolve almost exactly the quantum many body problem and are thus particularly relevant in this domain. In the first part of this thesis, we describe how we built a table of a multi-phase EoS of the hydrogen, using ab initio methods in the field of the WDM. We show in particular that this EoS is in very good agreement with most of the available experimental data (principal Hugoniot, sound velocity in the molecular fluid, melting curve at low pressure, measurements of multiple shocks). In the second part, we present a direct application of our EoS by showing its influence on the criteria of ignition and combustion of two target designs for ICF: a self-ignited target which will be used on the Laser MegaJoule (LMJ), and a shock-ignited target. We show in particular that the optimization of the laser pulse allowing maximizing the thermonuclear energy is strongly dependent on the precision of the EoS in the strong coupled and degenerate domain. (author) [French] Dans le domaine de la fusion par confinement inertiel (FCI), l'equation d'etat (EoS) de l'hydrogene et de ses isotopes est tres certainement une des proprietes les plus importantes a connaitre. Les EoS basees sur des modeles chimiques peinent a donner une description univoque de l'hydrogene dans le domaine de couplage et de degenerescence partiels, appele matiere dense et tiede, ou Warm Dense Matter (WDM). En effets, ces modeles utilisent des potentiels ad hoc pour decrire les

  9. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

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

  10. Multiscale integral analysis of a HT leakage in a fusion nuclear power plant

    Science.gov (United States)

    Velarde, M.; Fradera, J.; Perlado, J. M.; Zamora, I.; Martínez-Saban, E.; Colomer, C.; Briani, P.

    2016-05-01

    The present work presents an example of the application of an integral methodology based on a multiscale analysis that covers the whole tritium cycle within a nuclear fusion power plant, from a micro scale, analyzing key components where tritium is leaked through permeation, to a macro scale, considering its atmospheric transport. A leakage from the Nuclear Power Plants, (NPP) primary to the secondary side of a heat exchanger (HEX) is considered for the present example. Both primary and secondary loop coolants are assumed to be He. Leakage is placed inside the HEX, leaking tritium in elementary tritium (HT) form to the secondary loop where it permeates through the piping structural material to the exterior. The Heating Ventilation and Air Conditioning (HVAC) system removes the leaked tritium towards the NPP exhaust. The HEX is modelled with system codes and coupled to Computational Fluid Dynamic (CFD) to account for tritium dispersion inside the nuclear power plants buildings and in site environment. Finally, tritium dispersion is calculated with an atmospheric transport code and a dosimetry analysis is carried out. Results show how the implemented methodology is capable of assessing the impact of tritium from the microscale to the atmospheric scale including the dosimetric aspect.

  11. Lower Hybrid antennas for nuclear fusion experiments

    CERN Document Server

    Hillairet, Julien; Bae, Young-Soon; Bai, X; Balorin, C; Baranov, Y; Basiuk, V; Bécoulet, A; Belo, J; Berger-By, G; Brémond, S; Castaldo, C; Ceccuzzi, S; Cesario, R; Corbel, E; Courtois, X; Decker, J; Delmas, E; Delpech, L; Ding, X; Douai, D; Ekedahl, A; Goletto, C; Goniche, M; Guilhem, D; Hertout, P; Imbeaux, F; Litaudon, X; Magne, R; Mailloux, J; Mazon, D; Mirizzi, F; Mollard, P; Moreau, P; Oosako, T; Petrzilka, V; Peysson, Y; Poli, S; Preynas, M; Prou, M; Saint-Laurent, F; Samaille, F; Saoutic, B

    2015-01-01

    The nuclear fusion research goal is to demonstrate the feasibility of fusion power for peaceful purposes. In order to achieve the conditions similar to those expected in an electricity-generating fusion power plant, plasmas with a temperature of several hundreds of millions of degrees must be generated and sustained for long periods. For this purpose, RF antennas delivering multi-megawatts of power to magnetized confined plasma are commonly used in experimental tokamaks. In the gigahertz range of frequencies, high power phased arrays known as "Lower Hybrid" (LH) antennas are used to extend the plasma duration. This paper reviews some of the technological aspects of the LH antennas used in the Tore Supra tokamak and presents the current design of a proposed 20 MW LH system for the international experiment ITER.

  12. Reactor for boron fusion with picosecond ultrahigh power laser pulses and ultrahigh magnetic field trapping

    CERN Document Server

    Miley, G H; Kirchhoff, G

    2015-01-01

    Compared with the deuterium tritium (DT) fusion, the environmentally clean fusion of protons with 11B is extremely difficult. When instead of nanosecond laser pulses for thermal-ablating driven ignition, picosecond pulses are used, a drastic change by nonlinearity results in ultrahigh acceleration of plasma blocks. This radically changes to economic boron fusion by a measured new avalanche ignition.

  13. Breeding blanket design and systems integration for a helium-cooled lithium-lead fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Li Puma, A. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif sur Yvette (France)]. E-mail: alipuma@cea.fr; Berton, J.L. [CEA Cadarache, Direction de l' Energie Nucleaire, F-13108 Saint Paul Lez Durance (France); Branas, B. [CIEMAT, Madrid (Spain); Buehler, L. [Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Doncel, J. [CIEMAT, Madrid (Spain); Fischer, U. [Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Farabolini, W. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif sur Yvette (France); Giancarli, L. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif sur Yvette (France); Maisonnier, D. [EFDA Close Support Unit, Boltzmannstr. 2, D-85748 Garching (Germany); Pereslavtsev, P. [Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany); Raboin, S. [CEA Cadarache, Direction de l' Energie Nucleaire, F-13108 Saint Paul Lez Durance (France); Salavy, J.-F. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif sur Yvette (France); Sardain, P. [EFDA Close Support Unit, Boltzmannstr. 2, D-85748 Garching (Germany); Szczepanski, J. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif sur Yvette (France); Ward, D. [UKAEA, Culham Science Center, Abingdon, Oxfordshire OX14 3DB (United Kingdom)

    2006-02-15

    This paper describes the work performed for the power plant conceptual studies (PPCS) reactor model AB, a 'near-term' fusion power reactor based on limited extrapolations from today knowledge on technology and plasma physic assumption, featuring a helium-cooled lithium lead (HCLL) blanket concept and a He cooled divertor. The main design rational and characteristics of the PPCS model AB as well as the main assumption and achieved results are reported. The reactor is able to supply to the grid a net electrical power of 1.5 GW{sub e} for a fusion power of 4.2 GW{sub th} and features a major radius of 9.56 m. A tritium breeding ratio (TBR) of 1.13 will assure the tritium self sufficiency.

  14. Creating a Star: The Science of Fusion--Fusion Power Would Not Contribute to Global Warming, Acid Rain, or Other Forms of Air Pollution, nor Would It Create Long-Lived Radioactive Waste

    Science.gov (United States)

    Baird, Stephen L.

    2005-01-01

    Fusion is the process that powers the sun and the stars. Since the 1950s, scientists and engineers in the United States and around the world have been conducting fusion research in pursuit of the creation of a new energy source for our planet and to further our understanding and control of plasma, the fourth state of matter that dominates the…

  15. Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

    Science.gov (United States)

    Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

    2017-12-01

    Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

  16. Frontiers in fusion research

    CERN Document Server

    Kikuchi, Mitsuru

    2011-01-01

    Frontiers in Fusion Research provides a systematic overview of the latest physical principles of fusion and plasma confinement. It is primarily devoted to the principle of magnetic plasma confinement, that has been systematized through 50 years of fusion research. Frontiers in Fusion Research begins with an introduction to the study of plasma, discussing the astronomical birth of hydrogen energy and the beginnings of human attempts to harness the Sun's energy for use on Earth. It moves on to chapters that cover a variety of topics such as: * charged particle motion, * plasma kinetic theory, *

  17. Near-Term IEC Thrusters and Future Fusion Propulsion

    Science.gov (United States)

    Miley, George; Gu, Yibin; Jurczyk, Brian

    1997-11-01

    For space missions beyond orbital and lunar distances, studies indicate that advanced power and propulsion systems will be required. Conceptual fusion rocket design studies using the Inertial Electrostatic Confinement (IEC) concept have predicted excellent performance for a variety of space missions.(Williams, C.H., S.K. Borowski, "Fusion Propulsion System Survey and Desired Operating Parameters," Fusion Propulsion Workshop, Huntsville, AL, 1997.)(Bussard, R.W., L.W. Jameson, "Design Considerations for Clean QED Fusion Propulsion Systems," Proc. 11th Symp. on Space Nuclear Power and Propulsion, 1994.)(Miley, G.H., et al., "Innovative Technology for and Inertial Electrostatic Confinement Fusion Propulsion Unit," Fusion Energy in Space Propulsion, vol. 167, 1995.) Research at the University of Illinois has shifted towards the development of a small-scale xenon jet plasma thruster ( 500W) for satellite adjustment and station-keeping. The scalability of the IEC physics will allow the research on this thruster to contribute to the eventual goal of a high-power fusion propulsion unit, while simultaneously generating a spin-off technology that can be utilized in the near term. A presentation of the experimental setup and preliminary results will be given for the IEC thruster.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-07

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

  19. ORNL fusion power demonstration study: arguments for a vacuum building in which to enclose a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Werner, R.W.

    1976-12-01

    Fusion reactors as presently contemplated are excessively complicated, are virtually inaccessible for some repairs, and are subject to frequent loss of function. This dilemma arises in large part because the closed surface that separates the ''hard'' vacuum of the plasma zone from atmospheric pressure is located at the first wall or between blanket and shield. This closed surface is one containing hundreds to thousands of linear meters of welds or mechanical seals which are subject to radiation damage and cyclic fatigue. In situ repair is extremely difficult. This paper examines the arguments favoring the enclosing of the entire reactor in a vacuum building and thus changing the character of this closed surface from one requiring absolute vacuum integrity to one of high pumping impedance. Two differentially pumped vacuum zones are imagined, one clean zone for the plasma and one for the balance of the volume. Both would be at substantially the same pressure. Other advantages for the vacuum enclosure are also cited and discussed.

  20. Atomic and molecular processes in fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Janev, R.K. [International Atomic Energy Agency, Vienna (Austria)

    1997-01-01

    The role of atomic and molecular processes in achieving and maintaining the conditions for thermonuclear burn in a magnetically confined fusion plasma is described. Emphasis is given to the energy balance and power and particle exhaust issues. The most important atomic and molecular processes which affect the radiation losses and impurity transport in the core plasma, the neutral particle transport in the plasma edge and the radiative cooling of divertor plasmas are discussed in greater detail. (author)

  1. Ablation of a Deuterium Pellet in a Fusion Plasma Viewed as a Stopping Power Problem

    DEFF Research Database (Denmark)

    Chang, C. T.

    1983-01-01

    At present, the most exploited technology to refuel a future fusion reactor is the high speed injection of macroscopic size pellet of solid hydrogen isotopes. The basic idea is that the ablation of a pellet in a fusion reactor is mainly caused by thermal electrons (~ 10 keV) /1/. Due to the low...

  2. Heavy ion fusion--Using heavy ions to make electricity

    Energy Technology Data Exchange (ETDEWEB)

    Celata, C.M.

    2004-03-15

    The idea of using nuclear fusion as a source of commercial electrical power has been pursued worldwide since the 1950s. Two approaches, using magnetic and inertial confinement of the reactants, are under study. This paper describes the difference between the two approaches, and discusses in more detail the heavy-ion-driven inertial fusion concept. A multibeam induction linear accelerator would be used to bring {approx}100 heavy ion beams to a few GeV. The beams would then heat and compress a target of solid D-T. This approach is unique among fusion concepts in its ability to protect the reaction chamber wall from neutrons and debris.

  3. Fusion materials: Technical evaluation of the technology of vandium alloys for use as blanket structural materials in fusion power systems

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-04

    The Committee`s evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is United and will require an extensive development and test program. Because of the chemical reactivity of vanadium the alloy will be less tolerant of system failures, accidents, and off-normal events than most other candidate blanket structural materials and will require more careful handling during fabrication of hardware. Because of the cost of the material more stringent requirements on processes, and minimal historical worlding experience, it will cost an order of magnitude to qualify a vanadium alloy for ITER blanket structures than other candidate materials. The use of vanadium is difficult and uncertain; therefore, other options should be explored more thoroughly before a final selection of vanadium is confirmed. The Committee views the risk as being too high to rely solely on vanadium alloys. In viewing the state and nature of the design of the ITER blanket as presented to the Committee, h is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium allay option is to be pursued, a large program needs to be started immediately. The commitment of funding and other resources needs to be firm and consistent with a realistic program plan.

  4. INSTALLATION OF A POST-ACCIDENT CONFINEMENT HIGH-LEVEL RADIATION MONITORING SYSTEM IN THE KOLA NUCLEAR POWER STATION (UNIT 2) IN RUSSIA

    Energy Technology Data Exchange (ETDEWEB)

    GREENE,G.A.; GUPPY,J.G.

    1998-09-01

    This is the final report on the INSP project entitled, ``Post-Accident Confinement High-Level Radiation Monitoring System'' conducted by BNL under the authorization of Project Work Plan WBS 1.2.2.6 (Attachment 1). This project was initiated in February 1993 to assist the Russians in reducing risks associated with the continued operation of older Soviet-designed nuclear power plants, specifically the Kola VVER-440/230 Unit 2, through improved accident detection capability, specifically by the installation of a dual train high-level radiation detection system in the confinement of Unit 2 of the Kola NPP. The major technical objective of this project was to provide, install and make operational the necessary hardware inside the confinement of the Kola NPP Unit 2 to provide early and reliable warning of the release of radionuclides from the reactor into the confinement air space as an indication of the occurrence of a severe accident at the plant. In addition, it was intended to provide hands-on experience and training to the Russian plant workers in the installation, operation, calibration and maintenance of the equipment in order that they may use the equipment without continued US assistance as an effective measure to improve reactor safety at the plant.

  5. A Particle X-ray Temporal Diagnostic (PXTD) for studies of kinetic, multi-ion effects, and ion-electron equilibration rates in Inertial Confinement Fusion plasmas at OMEGA (invited).

    Science.gov (United States)

    Sio, H; Frenje, J A; Katz, J; Stoeckl, C; Weiner, D; Bedzyk, M; Glebov, V; Sorce, C; Gatu Johnson, M; Rinderknecht, H G; Zylstra, A B; Sangster, T C; Regan, S P; Kwan, T; Le, A; Simakov, A N; Taitano, W T; Chacòn, L; Keenan, B; Shah, R; Sutcliffe, G; Petrasso, R D

    2016-11-01

    A Particle X-ray Temporal Diagnostic (PXTD) has been implemented on OMEGA for simultaneous time-resolved measurements of several nuclear products as well as the x-ray continuum produced in High Energy Density Plasmas and Inertial Confinement Fusion implosions. The PXTD removes systematic timing uncertainties typically introduced by using multiple instruments, and it has been used to measure DD, DT, D(3)He, and T(3)He reaction histories and the emission history of the x-ray core continuum with relative timing uncertainties within ±10-20 ps. This enables, for the first time, accurate and simultaneous measurements of the x-ray emission histories, nuclear reaction histories, their time differences, and measurements of Ti(t) and Te(t) from which an assessment of multiple-ion-fluid effects, kinetic effects during the shock-burn phase, and ion-electron equilibration rates can be made.

  6. A compact proton spectrometer for measurement of the absolute DD proton spectrum from which yield and ρR are determined in thin-shell inertial-confinement-fusion implosions.

    Science.gov (United States)

    Rosenberg, M J; Zylstra, A B; Frenje, J A; Rinderknecht, H G; Johnson, M Gatu; Waugh, C J; Séguin, F H; Sio, H; Sinenian, N; Li, C K; Petrasso, R D; Glebov, V Yu; Hohenberger, M; Stoeckl, C; Sangster, T C; Yeamans, C B; LePape, S; Mackinnon, A J; Bionta, R M; Talison, B; Casey, D T; Landen, O L; Moran, M J; Zacharias, R A; Kilkenny, J D; Nikroo, A

    2014-10-01

    A compact, step range filter proton spectrometer has been developed for the measurement of the absolute DD proton spectrum, from which yield and areal density (ρR) are inferred for deuterium-filled thin-shell inertial confinement fusion implosions. This spectrometer, which is based on tantalum step-range filters, is sensitive to protons in the energy range 1-9 MeV and can be used to measure proton spectra at mean energies of ∼1-3 MeV. It has been developed and implemented using a linear accelerator and applied to experiments at the OMEGA laser facility and the National Ignition Facility (NIF). Modeling of the proton slowing in the filters is necessary to construct the spectrum, and the yield and energy uncertainties are ±DD-neutron yield diagnostics at the NIF.

  7. Modelling and simulation of new generation powerful gyrotrons for the fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Sabchevski, S [Institute of Electronics, Bulgarian Academy of Sciences, BG-1784 Sofia (Bulgaria); Zhelyazkov, I [Faculty of Physics, Sofia University, BG-1164 Sofia (Bulgaria)

    2007-04-15

    One of the important issues related with the cyclotron resonance heating (CRH) and current drive of fusion plasmas in thermonuclear reactors (tokamaks and stellarators) is the development of multi-megawatt class gyrotrons. There are generally three stages of the implementation of that task, notably (i) elaborating a novel generation of software tools for the physical modelling and simulation of such kind of gyrotrons, (ii) their computer aided design (CAD) and construction on the basis of the simulation's results, and finally, (iii) gyrotrons' testing in real experimental conditions. This tutorial paper concerns the first item-the development of software tools. In co-operation with the Institute for Pulsed Power and Microwave Technology at the Forschungszentrum Karlsruhe, Germany, and Centre de Recherches en Physique des Plasmas at Ecole Polytechnique Federale de Lausanne, Switzerland, we work on the conceptual design of the software tools under development. The basic conclusions are that the numerical codes for gyrotrons' modelling should possess the following essential characteristics: (a) portability, (b) extensibility, (c) to be oriented toward the solution of practical problems (i.e., elaborating of computer programs that can be used in the design process), (d) to be based on self-consistent 3D physical models, which take into account the departure from axial symmetry, and (e) ability to simulate time dependent processes (electrostatic PIC simulation) alongside with a trajectory analysis (ray tracing simulation). Here, we discuss how various existing numerical codes have to be improved and implemented via the advanced programming technologies for state-of-the-art computer systems including clusters, grid, parallel platforms, and supercomputers.

  8. Spherical microwave confinement and ball lightning

    Science.gov (United States)

    Robinson, William Richard

    This dissertation presents the results of research done on unconventional energy technologies from 1995 to 2009. The present civilization depends on an infrastructure that was constructed and is maintained almost entirely using concentrated fuels and ores, both of which will run out. Diffuse renewable energy sources rely on this same infrastructure, and hence face the same limitations. I first examined sonoluminescence directed toward fusion, but demonstrated theoretically that this is impossible. I next studied Low Energy Nuclear Reactions and developed methods for improving results, although these have not been implemented. In 2000, I began Spherical Microwave Confinement (SMC), which confines and heats plasma with microwaves in a spherical chamber. The reactor was designed and built to provide the data needed to investigate the possibility of achieving fusion conditions with microwave confinement. A second objective was to attempt to create ball lightning (BL). The reactor featured 20 magnetrons, which were driven by a capacitor bank and operated in a 0.2 s pulse mode at 2.45 GHz. These provided 20 kW to an icosahedral array of 20 antennas. Video of plasmas led to a redesign of the antennas to provide better coupling of the microwaves to the plasma. A second improvement was a grid at the base of the antennas, which provided corona electrons and an electric field to aid quick formation of plasmas. Although fusion conditions were never achieved and ball lightning not observed, experience gained from operating this basic, affordable system has been incorporated in a more sophisticated reactor design intended for future research. This would use magnets that were originally planned. The cusp geometry of the magnetic fields is suitable for electron cyclotron resonance in the same type of closed surface that in existing reactors has generated high-temperature plasmas. Should ball lightning be created, it could be a practical power source with nearly ideal

  9. Controlled Nuclear Fusion.

    Science.gov (United States)

    Glasstone, Samuel

    This publication is one of a series of information booklets for the general public published by The United States Atomic Energy Commission. Among the topics discussed are: Importance of Fusion Energy; Conditions for Nuclear Fusion; Thermonuclear Reactions in Plasmas; Plasma Confinement by Magnetic Fields; Experiments With Plasmas; High-Temperature…

  10. Controlled thermonuclear fusion

    CERN Document Server

    Bobin, Jean Louis

    2014-01-01

    The book is a presentation of the basic principles and main achievements in the field of nuclear fusion. It encompasses both magnetic and inertial confinements plus a few exotic mechanisms for nuclear fusion. The state-of-the-art regarding thermonuclear reactions, hot plasmas, tokamaks, laser-driven compression and future reactors is given.

  11. ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets

    CERN Document Server

    Sorbom, B N; Palmer, T R; Mangiarotti, F J; Sierchio, J M; Bonoli, P; Kasten, C; Sutherland, D A; Barnard, H S; Haakonsen, C B; Goh, J; Sung, C; Whyte, D G

    2014-01-01

    The affordable, robust, compact (ARC) reactor conceptual design study aims to reduce the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Q_p~13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ~63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ~23 T peak field on coil with newly available REBCO superconductor technology. External cu...

  12. ADVANCED FUSION TECHNOLOGY RESEARCH AND DEVELOPMENT ANNUAL REPORT TO THE US DEPARTMENT OF ENERGY

    Energy Technology Data Exchange (ETDEWEB)

    PROJECT STAFF

    2001-09-01

    OAK A271 ADVANCED FUSION TECHNOLOGY RESEARCH AND DEVELOPMENT ANNUAL REPORT TO THE US DEPARTMENT OF ENERGY. The General Atomics (GA) Advanced Fusion Technology Program seeks to advance the knowledge base needed for next-generation fusion experiments, and ultimately for an economical and environmentally attractive fusion energy source. To achieve this objective, they carry out fusion systems design studies to evaluate the technologies needed for next-step experiments and power plants, and they conduct research to develop basic and applied knowledge about these technologies. GA's Advanced Fusion Technology program derives from, and draws on, the physics and engineering expertise built up by many years of experience in designing, building, and operating plasma physics experiments. The technology development activities take full advantage of the GA DIII-D program, the DIII-D facility and the Inertial Confinement Fusion (ICF) program and the ICF Target Fabrication facility.

  13. Magnetized Target Fusion At General Fusion: An Overview

    Science.gov (United States)

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

    2017-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

  15. An overview of the development of the first wall and other principal components of a laser fusion power plant

    Science.gov (United States)

    Sethian, John D.; Raffray, A. Rene; Latkowski, Jeffery; Blanchard, James P.; Snead, Lance; Renk, Timothy J.; Sharafat, Shahram

    2005-12-01

    This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

  16. ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets

    Energy Technology Data Exchange (ETDEWEB)

    Sorbom, B.N., E-mail: bsorbom@mit.edu; Ball, J.; Palmer, T.R.; Mangiarotti, F.J.; Sierchio, J.M.; Bonoli, P.; Kasten, C.; Sutherland, D.A.; Barnard, H.S.; Haakonsen, C.B.; Goh, J.; Sung, C.; Whyte, D.G.

    2015-11-15

    Highlights: • ARC reactor designed to have 500 MW fusion power at 3.3 m major radius. • Compact, simplified design allowed by high magnetic fields and jointed magnets. • ARC has innovative plasma physics solutions such as inboardside RF launch. • High temperature superconductors allow high magnetic fields and jointed magnets. • Liquid immersion blanket and jointed magnets greatly simplify tokamak reactor design. - Abstract: The affordable, robust, compact (ARC) reactor is the product of a conceptual design study aimed at reducing the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a ∼200–250 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Q{sub p} ≈ 13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ∼63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ∼23 T peak field on coil achievable with newly available REBCO superconductor technology. External current drive is provided by two innovative inboard RF launchers using 25 MW of lower hybrid and 13.6 MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent

  17. Fusion Plasma Physics and ITER - An Introduction (1/4)

    CERN Multimedia

    CERN. Geneva

    2011-01-01

    In November 2006, ministers representing the world’s major fusion research communities signed the agreement formally establishing the international project ITER. Sited at Cadarache in France, the project involves China, the European Union (including Switzerland), India, Japan, the Russian Federation, South Korea and the United States. ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. The ITER tokamak is designed to study the “burning plasma” regime in deuterium-tritium (D-T) plasmas by achieving a fusion amplification factor, Q (the ratio of fusion output power to plasma heating input power), of 10 for several hundreds of seconds with a nominal fusion power output of 500MW. It is also intended to allow the study of steady-state plasma operation at Q≥5 by me...

  18. History and status of magnetic fusion research; Evolution et statut des recherches sur la fusion controlee

    Energy Technology Data Exchange (ETDEWEB)

    Jacquinot, J. [CEA Saclay, Cabinet du Haut Commissaire, 91 - Gif-sur-Yvette (France)

    2008-02-15

    Ever since the understanding of the basic process which powers the stars has been elucidated, humanity has been dreaming to master controlled fusion for peaceful purposes. Controlled fusion in a steady state regime must use magnetic confinement of a gas (plasma) heated up to 150 millions degrees. Physics and technology involved in such a state are extremely complex and went through many up and down phases. Nevertheless, the overall progress has been spectacular and a significant amount of energy could be produced in a well controlled manner. On this basis, an international organisation of unprecedented magnitude involving 34 countries has started working in Cadarache for the construction of the ITER project. It aims at the scientific demonstration of controlled fusion at the level of 500 MW and a power gain of 10. (author)

  19. Use of Clearance Indexes to Assess Waste Disposal Issues for the HYLIFE-II Inertial Fusion Energy Power Plant Design

    Energy Technology Data Exchange (ETDEWEB)

    Reyes, S; Latkowski, J F; Sanz, J

    2002-01-17

    Traditionally, waste management studies for fusion energy have used the Waste Disposal Rating (WDR) to evaluate if radioactive material from irradiated structures could qualify for shallow land burial. However, given the space limitations and the negative public perception of large volumes of waste, there is a growing international motivation to develop a fusion waste management system that maximizes the amount of material that can be cleared or recycled. In this work, we present an updated assessment of the waste management options for the HYLIFE-II inertial fusion energy (IFE) power plant, using the concept of Clearance Index (CI) for radioactive waste disposal. With that purpose, we have performed a detailed neutronics analysis of the HYLIFE-II design, using the TART and ACAB computer codes for neutron transport and activation, respectively. Whereas the traditional version of ACAB only provided the user with the WDR as an index for waste considerations, here we have modified the code to calculate Clearance Indexes using the current International Atomic Energy Agency (IAEA) clearance limits for radiological waste disposal. The results from the analysis are used to perform an assessment of the waste management options for the HYLIFE-II IFE design.

  20. Magnetic confinement

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, Paola; De Marco, Francesco; Pieroni, Leonardo (ed.)

    2005-07-01

    The Frascati Tokamak Upgrade (FTU) is a compact, high-magnetic-field tokamak capable of operating at density and magnetic field values similar to, or even encompassing, those of International Thermonuclear Experimental Reactor (ITER) and therefore provides a unique opportunity to explore physics issues that are directly relevant to ITER. During 2004 the experimental activities were focussed on fully exploiting the lower hybrid system (for generating and controlling the plasma current) and the electron cyclotron heating system (joint experiment with the Institute of Plasma Physics of the National Research Council, Milan). With all four gyrotrons in operation, full electron cyclotron power was achieved up to a record level of 1.5 MW. By simultaneously injecting lower hybrid waves, to tailor the plasma current radial profile, and electron cyclotron waves, to heat the plasma centre, good confinement regimes with internal transport barriers were obtained at the highest plasma density values ever achieved for this operation regime (n {approx}1.5X10{sup 20}m{sup -3}). Specific studies were devoted to optimising the coupling of lower hybrid waves to the plasma (by real-time control of the plasma position) and to generating current by electron cyclotron current drive. The new scanning CO{sub 2} interferometer (developed by the Reversed Field Experiment Consortium) for high spatial and time resolution (1 cm/50 {mu}s) density profile measurements was extensively used. The Thomson scattering diagnostic was upgraded and enabled observation of scattered signals associated with the Confinement background plasma dynamics. As for theoretical studies on the dynamics of turbulence in plasmas, the transition from Bohm-like scaling to gyro-Bohm scaling of the local plasma diffusivity was demonstrated on the basis of a generalised four wave model (joint collaboration with Princeton Plasma Physics Laboratory and the University of California at Irvine). The transition from weak to strong

  1. Tokamak Fusion Test Reactor. Final conceptual design report. [Overall cost and scheduling program

    Energy Technology Data Exchange (ETDEWEB)

    1976-02-01

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

  2. Deliberating and communicating the potential of fusion power based on long-term foresight knowledge

    Energy Technology Data Exchange (ETDEWEB)

    Laes, Erik [SCK-CEN - Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium)], E-mail: erik.laes@sckcen.be; Bombaerts, Gunter [SCK-CEN - Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); UGent (Ghent University) (Belgium)

    2007-10-15

    The main aim of this contribution is to provide guidance (in terms of quality criteria) for setting up foresight exercises as a platform for discussion and communication of the benefits and drawbacks of fusion with a broad range of stakeholders. At the same time, we explore conditions that might enhance the resonance of such foresight exercises in the policy sphere. In order to address this dual aim, we first introduce a philosophical framework called 'constructivism'. Next, we give a constructivist reading of scientific foresight as a combined scientific-political practice and point out some of the main points of interest regarding the relationship between foresight knowledge and policy. We illustrate these points of interest with practical case-study examples. Finally, we draw upon our theoretical and case-study research to propose some points of particular interest for the fusion community wishing to develop long-term energy scenarios.

  3. Conceptual study of ferromagnetic pebbles for heat exhaust in fusion reactors with short power decay length

    Directory of Open Access Journals (Sweden)

    N. Gierse

    2015-03-01

    The key results of this study are that very high heat fluxes are accessible in the operation space of ferromagnetic pebbles, that ferromagnetic pebbles are compatible with tokamak operation and current divertor designs, that the heat removal capability of ferromagnetic pebbles increases as λq decreases and, finally, that for fusion relevant values of q∥ pebble diameters below 100 μm are required.

  4. Study of the cooling systems with S-CO2 for the DEMO fusion power reactor.

    Czech Academy of Sciences Publication Activity Database

    Veselý, L.; Dostál, V.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 244-247 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : DEMO * Cooling * Energy conversion * Thermal cycle * Carbon dioxide * SCO2a Subject RIV: JF - Nuclear Energetics Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617305719

  5. INTRODUCTION: Status report on fusion research

    Science.gov (United States)

    Burkart, Werner

    2005-10-01

    members' personal views on the latest achievements in fusion research, including magnetic and inertial confinement scenarios. The report describes fusion fundamentals and progress in fusion science and technology, with ITER as a possible partner in the realization of self-sustainable burning plasma. The importance of the socio-economic aspects of energy production using fusion power plants is also covered. Noting that applications of plasma science are of broad interest to the Member States, the report addresses the topic of plasma physics to assist in understanding the achievements of better coatings, cheaper light sources, improved heat-resistant materials and other high-technology materials. Nuclear fusion energy production is intrinsically safe, but for ITER the full range of hazards will need to be addressed, including minimising radiation exposure, to accomplish the goal of a sustainable and environmentally acceptable production of energy. We anticipate that the role of the Agency will in future evolve from supporting scientific projects and fostering information exchange to the preparation of safety principles and guidelines for the operation of burning fusion plasmas with a Q > 1. Technical progress in inertial and magnetic confinement, as well as in alternative concepts, will lead to a further increase in international cooperation. New means of communication will be needed, utilizing the best resources of modern information technology to advance interest in fusion. However, today the basis of scientific progress is still through journal publications and, with this in mind, we trust that this report will find an interested readership. We acknowledge with thanks the support of the members of the IFRC as an advisory body to the Agency. Seven chairmen have presided over the IFRC since its first meeting in 1971 in Madison, USA, ensuring that the IAEA fusion efforts were based on the best professional advice possible, and that information on fusion developments has

  6. Spatially resolved X-ray emission measurements of the residual velocity during the stagnation phase of inertial confinement fusion implosion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Ruby, J. J.; Pak, A., E-mail: pak5@llnl.gov; Field, J. E.; Ma, T.; Spears, B. K.; Benedetti, L. R.; Bradley, D. K.; Berzak Hopkins, L. F.; Casey, D. T.; Döppner, T.; Eder, D.; Fittinghoff, D.; Grim, G.; Hatarik, R.; Hinkel, D. E.; Izumi, N.; Khan, S. F.; Kritcher, A. L.; Moody, J. D.; Nagel, S. R. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); and others

    2016-07-15

    A technique for measuring residual motion during the stagnation phase of an indirectly driven inertial confinement experiment has been implemented. This method infers a velocity from spatially and temporally resolved images of the X-ray emission from two orthogonal lines of sight. This work investigates the accuracy of recovering spatially resolved velocities from the X-ray emission data. A detailed analytical and numerical modeling of the X-ray emission measurement shows that the accuracy of this method increases as the displacement that results from a residual velocity increase. For the typical experimental configuration, signal-to-noise ratios, and duration of X-ray emission, it is estimated that the fractional error in the inferred velocity rises above 50% as the velocity of emission falls below 24 μm/ns. By inputting measured parameters into this model, error estimates of the residual velocity as inferred from the X-ray emission measurements are now able to be generated for experimental data. Details of this analysis are presented for an implosion experiment conducted with an unintentional radiation flux asymmetry. The analysis shows a bright localized region of emission that moves through the larger emitting volume at a relatively higher velocity towards the location of the imposed flux deficit. This technique allows for the possibility of spatially resolving velocity flows within the so-called central hot spot of an implosion. This information would help to refine our interpretation of the thermal temperature inferred from the neutron time of flight detectors and the effect of localized hydrodynamic instabilities during the stagnation phase. Across several experiments, along a single line of sight, the average difference in magnitude and direction of the measured residual velocity as inferred from the X-ray and neutron time of flight detectors was found to be ∼13 μm/ns and ∼14°, respectively.

  7. Fusion technology annual report of the association EURATOM/CEA 1998; Technologie de la fusion Rapport annuel 1998 Association EURATOM/CEA 1998

    Energy Technology Data Exchange (ETDEWEB)

    Magaud, P.; Le vagueres, F

    1998-07-01

    In this book are found technical and scientific papers on the main works carried out in the frame of the european program of fusion technology, during 1998. The presented activities are: plasma facing components, vacuum vessel and shield, magnets, remote handling, safety (short and long term), european blanket project (long term) with water cooled lithium lead and helium cooled pebble bed blanket, materials for fusion power plant, socio-economic research on fusion, plasma facing components, fuel cycle, inertial confinement. (A.L.B.)

  8. Alternative approaches to plasma confinement

    Science.gov (United States)

    Roth, J. R.

    1978-01-01

    The paper discusses 20 plasma confinement schemes each representing an alternative to the tokamak fusion reactor. Attention is given to: (1) tokamak-like devices (TORMAC, Topolotron, and the Extrap concept), (2) stellarator-like devices (Torsatron and twisted-coil stellarators), (3) mirror machines (Astron and reversed-field devices, the 2XII B experiment, laser-heated solenoids, the LITE experiment, the Kaktus-Surmac concept), (4) bumpy tori (hot electron bumpy torus, toroidal minimum-B configurations), (5) electrostatically assisted confinement (electrostatically stuffed cusps and mirrors, electrostatically assisted toroidal confinement), (6) the Migma concept, and (7) wall-confined plasmas. The plasma parameters of the devices are presented and the advantages and disadvantages of each are listed.

  9. Fusion for Space Propulsion and Plasma Liner Driven MTF

    Science.gov (United States)

    Thio, Y.C. Francis; Rodgers, Stephen L. (Technical Monitor)

    2001-01-01

    in the light of significant development of the enabling pulsed power component technologies that have occurred in the last two decades because of defense and other energy requirements. The extreme states of matter required to produce fusion reactions may be more readily realizable in the pulsed states with less system mass than in steady states. Significant saving in system mass may result in pulsed fusion systems using plasmas in the appropriate density regimes. Magnetized target fusion, which attempts to combine the favorable attributes of magnetic confinement and inertial compression-containment into one single integrated fusion scheme, appears to have benefits that are worth exploring for propulsion application.

  10. Response measurement of single-crystal chemical vapor deposition diamond radiation detector for intense X-rays aiming at neutron bang-time and neutron burn-history measurement on an inertial confinement fusion with fast ignition.

    Science.gov (United States)

    Shimaoka, T; Kaneko, J H; Arikawa, Y; Isobe, M; Sato, Y; Tsubota, M; Nagai, T; Kojima, S; Abe, Y; Sakata, S; Fujioka, S; Nakai, M; Shiraga, H; Azechi, H; Chayahara, A; Umezawa, H; Shikata, S

    2015-05-01

    A neutron bang time and burn history monitor in inertial confinement fusion with fast ignition are necessary for plasma diagnostics. In the FIREX project, however, no detector attained those capabilities because high-intensity X-rays accompanied fast electrons used for plasma heating. To solve this problem, single-crystal CVD diamond was grown and fabricated into a radiation detector. The detector, which had excellent charge transportation property, was tested to obtain a response function for intense X-rays. The applicability for neutron bang time and burn history monitor was verified experimentally. Charge collection efficiency of 99.5% ± 0.8% and 97.1% ± 1.4% for holes and electrons were obtained using 5.486 MeV alpha particles. The drift velocity at electric field which saturates charge collection efficiency was 1.1 ± 0.4 × 10(7) cm/s and 1.0 ± 0.3 × 10(7) cm/s for holes and electrons. Fast response of several ns pulse width for intense X-ray was obtained at the GEKKO XII experiment, which is sufficiently fast for ToF measurements to obtain a neutron signal separately from X-rays. Based on these results, we confirmed that the single-crystal CVD diamond detector obtained neutron signal with good S/N under ion temperature 0.5-1 keV and neutron yield of more than 10(9) neutrons/shot.

  11. Study of high gain spherical shell ICF targets containing uniform layers of liquid deuterium tritium fuel. A numericial model for analyzing thermal layering of liquid mixtures of hydrogen isotopes inside a spherical inertial confinement fusion target: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, E.M.; Kim, Kyekyoon [Lawrence Livermore National Lab., CA (United States)

    1994-05-01

    A numerical model has been developed to describe the thermally induced behavior of a liquid layer of hydrogen isotopes inside a spherical Inertial Confinement Fusion (ICF) target and to calculate the far-field temperature gradient which will sustain a uniform liquid layer. This method is much faster than the trial-and-error method previously employed. The governing equations are the equations of continuity, momentum, energy, mass diffusion-convection, and conservation of the individual isotopic species. Ordinary and thermal diffusion equations for the diffusion of fluxes of the species are included. These coupled equations are solved by a finite-difference method using upwind schemes, variable mesh, and rigorous boundary conditions. The solution methodology unique to the present problem is discussed in detail. in particular, the significance of the surface tension gradient driven flows (also called Marangoni flows) in forming uniform liquid layers inside ICF targets is demonstrated. Using the theoretical model, the values of the externally applied thermal gradients that give rise to uniform liquid layers of hydrogen inside a cryogenic spherical-shell ICF target are calculated, and the results compared with the existing experimental data.

  12. Materials needs for compact fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A.

    1983-01-01

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

  13. Fusion Electra: A Krypton Fluoride Laser for Fusion Energy

    National Research Council Canada - National Science Library

    Sethian, J; Friedman, M; Giuliani, J; Lehmberg, R; Myers, M; Obenschain, S; Hegeler, F; Swanekamp, S

    2001-01-01

    .... We will focus on technologies that can be scalable to the 50-150 kJ energy needed for a full-size fusion power plant beam line and are projected to meet the economic requirements for fusion power...

  14. A fusion-driven gas core nuclear rocket

    Science.gov (United States)

    Kammash, T.; Godfroy, T.

    1998-01-01

    A magnetic confinement scheme is investigated as a potential propulsion device in which thrust is generated by a propellant heated by radiation emanating from a fusion plasma. The device in question is the gasdynamic mirror (GDM) machine in which a hot dense plasma is confined long enough to generate fusion energy while allowing a certain fraction of its charged particle population to go through one end to a direct converter. The energy of these particles is converted into electric power which is recirculated to sustain the steady state operation of the system. The injected power heats the plasma to thermonuclear temperatures where the resulting fusion energy appears a charged particle power, neutron power, and radiated power in the form of bremsstrahlung and synchrotron radiation. The neutron power can be converted through a thermal converter to electric power that can be combined with the direct converter power before being fed into the injector. The radiated power, on the other hand, can be used to heat a hydrogen propellant introduced into the system at a specified pressure and mass flow rate. This propellant can be pre-heated by regeneratively cooling the (mirror) nozzle or other components of the system if feasible, or by an electrothermal unit powered by portions of the recirculated power. Using a simple heat transfer model that ignores the heat flux to the wall, and assuming total absorption of radiation energy by the propellant it is shown that such a gas core rocket is capable of producing tens of kilonewtons of thrust and several thousands of seconds of specific impulse. It is also shown that the familiar Kelvin-Helmholtz instability which arises from the relative motion of the neutral hydrogen to the ionized fuel is not likely to occur in this system due to the presence of the confining magnetic field.

  15. Fusion plasma physics during half a century

    Energy Technology Data Exchange (ETDEWEB)

    Lehnert, Bo

    1999-08-01

    A review is given on the potentialities of fusion energy with respect to energy production and related environmental problems, the various approaches to controlled thermonuclear fusion, the main problem areas of research, the historical development, the present state of investigations, and future perspectives. This article also presents a personal memorandum of the author. Thereby special reference will be given to part of the research conducted at the Royal Institute of Technology in Stockholm, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. In large tokamak experiments temperatures above the ignition limit of about 10{sup 8} K have been reached under break-even conditions where the fusion power generation is comparable to the energy loss. A power producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient. The future international research programme has therefore to be conducted along broad lines, with necessary ingredients of basis research and new ideas, and also within lines of magnetic confinement being alternative to that of tokamaks.

  16. Realizing Technologies for Magnetized Target Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Wurden, Glen A. [Los Alamos National Laboratory

    2012-08-24

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

  17. The defects and microstructure in the fusion zone of multipass laser welded joints with Inconel 52M filler wire for nuclear power plants

    Science.gov (United States)

    Li, Gang; Lu, Xiaofeng; Zhu, Xiaolei; Huang, Jian; Liu, Luwei; Wu, Yixiong

    2017-09-01

    The defects and microstructure in the fusion zone of multipass laser welded joints with Inconel 52M filler wire are investigated for nuclear power plants. Experimental results indicate that the incomplete fusion forms as the deposited metals do not completely cover the groove during multipass laser welding. The dendritic morphologies are observed on the inner surface of the porosity in the fusion zone. Many small cellular are found in the zones near the fusion boundary. With solidification preceding, cellular gradually turn into columnar dendrites and symmetrical columnar dendrites are exhibited in the weld center of the fusion zone. The fine equiaxed grains form and columnar dendrites disappear in the remelted zone of two passes. The dendrite arm spacing in the fusion zone becomes widened with increasing welding heat input. Nb-rich carbides/carbonitrides are preferentially precipitated in the fusion zone of multipass laser welded joints. In respect to high cooling rate during multipass laser welding, element segregation could be insufficient to achieve the component of Laves phase.

  18. Superconductivity and fusion energy—the inseparable companions

    Science.gov (United States)

    Bruzzone, Pierluigi

    2015-02-01

    Although superconductivity will never produce energy by itself, it plays an important role in energy-related applications both because of its saving potential (e.g., power transmission lines and generators), and its role as an enabling technology (e.g., for nuclear fusion energy). The superconducting magnet’s need for plasma confinement has been recognized since the early development of fusion devices. As long as the research and development of plasma burning was carried out on pulsed devices, the technology of superconducting fusion magnets was aimed at demonstrations of feasibility. In the latest generation of plasma devices, which are larger and have longer confinement times, the superconducting coils are a key enabling technology. The cost of a superconducting magnet system is a major portion of the overall cost of a fusion plant and deserves significant attention in the long-term planning of electricity supply; only cheap superconducting magnets will help fusion get to the energy market. In this paper, the technology challenges and design approaches for fusion magnets are briefly reviewed for past, present, and future projects, from the early superconducting tokamaks in the 1970s, to the current ITER (International Thermonuclear Experimental Reactor) and W7-X projects and future DEMO (Demonstration Reactor) projects. The associated cryogenic technology is also reviewed: 4.2 K helium baths, superfluid baths, forced-flow supercritical helium, and helium-free designs. Open issues and risk mitigation are discussed in terms of reliability, technology, and cost.

  19. Simulations of burn dynamics in tokamak fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

    The global dynamics of tokamak reactors is investigated with the time-dependent, volume-averaged (0D) particle and power balance code FRESCO (Fusion REactor Simulation COde). The main emphasis is on studies of reactivity transients during tokamak start-up and shut down, as well as after sudden changes in plasma and tokamak parameters. In particular, the plasma responses to changes in the confinement, fuelling rates and impurity concentrations are considered. 76 refs.

  20. Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Neu, R., E-mail: Rudolf.Neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Fakultät für Maschinenbau, Technische Universität München, D-85748 Garching (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Coenen, J.W. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Brinkmann, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Calvo, A. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Elgeti, S. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); García-Rosales, C. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Greuner, H.; Hoeschen, T.; Holzner, G. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Klein, F. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Koch, F. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); and others

    2016-11-01

    Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W{sub f}/W) has been developed incorporating extrinsic toughening mechanisms. Small W{sub f}/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO{sub 3} compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

  1. Experimental results of a sheet-beam, high power, FEL amplifier with application to magnetic fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, S.; Destler, W.W.; Granatstein, V.L. [Univ. of Maryland, College Park, MD (United States)] [and others

    1995-12-31

    The experimental study of sheet-beam FELs as candidate millimeter-wave sources for heating magnetic fusion plasmas has achieved a major milestone. In a proof-of-principle, pulsed experiment, saturated FEL amplifier operation was achieved with 250 kW of output power at 86 GHz. Input microwave power was 1 kW, beam voltage was 450 kV and beam current was 17 A. The planar wiggler had a peak value of 3.8 kG, a period of 0.96 cm and was 71 cm long. The linear gain of 30 dB, saturated gain of 24 dB and saturated efficiency of 3% all are in good agreement with theoretical prediction. Follow-on work would include development of a thermionic sheet-beam electron-gun compatible with CW FEL operation, adding a section of tapered wiggler to increase the output power to levels in excess of 1 megawatt, and increasing the FEL frequency.

  2. Plasma transport in the Scrape-off-Layer of magnetically confined plasma and the plasma exhaust

    DEFF Research Database (Denmark)

    Rasmussen, Jens Juul; Naulin, Volker; Nielsen, Anders Henry

    An overview of the plasma dynamics in the Scrape-off-Layer (SOL) of magnetically confined plasma is presented. The SOL is the exhaust channel of the warm plasma from the core, and the understanding of the SOL plasma dynamics is one of the key issues in contemporary fusion research. It is essential...... for operation of fusion experiments and ultimately fusion power plants. Recent results clearly demonstrate that the plasma transport through the SOL is dominated by turbulent intermittent fluctuations organized into filamentary structures convecting particles, energy, and momentum through the SOL region. Thus......, the transport cannot be described and parametrized by simple diffusive type models. The transport leads to strong localized power loads on the first wall and the plasma facing components, which have serious lasting influence....

  3. Magnetic fusion technology

    CERN Document Server

    Dolan, Thomas J

    2014-01-01

    Magnetic Fusion Technology describes the technologies that are required for successful development of nuclear fusion power plants using strong magnetic fields. These technologies include: ? magnet systems, ? plasma heating systems, ? control systems, ? energy conversion systems, ? advanced materials development, ? vacuum systems, ? cryogenic systems, ? plasma diagnostics, ? safety systems, and ? power plant design studies. Magnetic Fusion Technology will be useful to students and to specialists working in energy research.

  4. Direct conversion of fusion energy

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, Markus

    2003-03-01

    Deuterium and tritium are expected to be used as fuel in the first fusion reactors. Energy is released as kinetic energy of ions and neutrons, when deuterium reacts with tritium. One way to convert the kinetic energy to electrical energy, is to let the ions and neutrons hit the reactor wall and convert the heat that is caused by the particle bombardment to electrical energy with ordinary thermal conversion. If the kinetic energy of the ions instead is converted directly to electrical energy, a higher efficiency of the energy conversion is possible. The majority of the fusion energy is released as kinetic energy of neutrons, when deuterium reacts with tritium. Fusion reactions such as the D-D reactions, the D-{sup 3}He reaction and the p-{sup 11}B reaction, where a larger part of the fusion energy becomes kinetic energy of charged particles, appears therefore more suitable for direct conversion. Since they have lower reactivity than the D-T reaction, they need a larger {beta}B{sup 2}{sub 0} to give sufficiently high fusion power density. Because of this, the fusion configurations spherical torus (ST) and field-reversed configuration (FRC), where high {beta} values are possible, appear interesting. Rosenbluth and Hinton come to the conclusion that efficient direct conversion isn't possible in closed field line systems and that open geometries, which facilitate direct conversion, provide inadequate confinement for D-{sup 3}He. It is confirmed in this study that it doesn't seem possible to achieve as high direct conversion efficiency in closed systems as in open systems. ST and FRC fusion power plants that utilize direct conversion seem however interesting. Calculations with the help of Maple indicate that the reactor parameters needed for a D-D ST and a D{sub 3} He ST hopefully are possible to achieve. The best energy conversion option for a D-D or D{sub 3} He ST appears to be direct electrodynamic conversion (DEC) together with ordinary thermal conversion

  5. New results on structure of low beta confinement Polywell cusps simulated by comsol multiphysics

    Directory of Open Access Journals (Sweden)

    B. Mahdavipour

    Full Text Available The Inertial electrostatic confinement (IEC is one of the ways for fusion approaches. It is one of the various methods which can be used to confine hot fusion plasma. The advantage of IEC is that the IEC experiments could be done in smaller size facilities than ITER or NIF, costing less money and moving forward faster. In IEC fusion, we need to trap adequate electrons to confine the desired ion density which is needed for a fusion reactor. Polywell is a device which uses the magnetic cusp system and traps the required amount of electrons for fusion reactions. The purpose of this device is to create a virtual cathode in order to achieve nuclear fusion using inertial electrostatic confinement (Miley and Krupakar Murali, 2014. In this paper, we have simulated the low beta Polywell. Then, we examined the effects of coil spacing, coils current, electron injection energy on confinement time. Keywords: Low beta confinement, Polywell, IEC, Comsol multiphysics

  6. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  7. Transport properties of inertial confinement fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Issanova, M.K.; Kodanova, S.K.; Ramazanov, T.S. [IETP, Al-Farabi Kazakh National University, Almaty (Kazakhstan); Hoffmann, D.H.H. [Institut fuer Kernphysik, Technische Universitaet Darmstadt (Germany)

    2016-06-15

    In this paper the transport properties of non-isothermal dense deuterium-tritium plasmas were studied. Based on the effective interaction potentials between particles, the Coulomb logarithm for a two-temperature nonisothermal dense plasma was obtained. These potentials take into consideration long-range multi-particle screening effects and short-range quantum-mechanical effects in two-temperature plasmas. Transport processes in such plasmas were studied using the Coulomb logarithm. The obtained results were compared with the theoretical works of other authors and with the results of molecular dynamics simulations. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  8. Supplemental heating of conventional Inertial Confinement Fusion

    Science.gov (United States)

    Thomas, B. R.; Hughes, S. J.; Garbett, W. J.; Sircombe, N. J.

    2016-03-01

    We report a new ICF scheme whereby a capsule is imploded to near ignition conditions and subsequently flooded with hot electrons generated from a short-pulse laser- plasma interaction so as to heat the whole assembly by a few hundred eV. The cold dense shell pressure is increased by a larger factor than that of the hot spot at the capsule core, so that further heating and compression of the hot spot occurs. We suggest it may be possible to drive the capsule to ignition by the pressure augmentation supplied by this extra deposition of energy.

  9. Recent progress of high-power negative ion beam development for fusion plasma heating

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Kazuhiro; Akino, Noboru; Aoyagi, Tetsuo [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1997-03-01

    A negative-ion-based neutral beam injector (N-NBI) has been constructed for JT-60U. The N-NBI is designed to inject 500 keV, 10 MW neutral beams using two ion sources, each producing a 500 keV, 22 A D{sup -} ion beam. Beam acceleration test started in July, 1995 using one ion source. In the preliminary experiment, D{sup -} ion beam of 13.5 A has been successfully accelerated with an energy of 400 keV (5.4 MW) for 0.12 s at an operating pressure of 0.22 Pa. This is the highest D{sup -} beam current and power in the world. Co-extracted electron current was effectively suppressed to the ratio of Ie/I{sub D}- <1. The highest energy beam of 460 keV, 2.4 A, 0.44 s has also been obtained. Neutral beam injection starts in March, 1996 using two ion sources. To realize 1 MeV class NBI system for ITER (International Thermonuclear Experimental Reactor), demonstration of ampere class negative ion beam acceleration up to 1 MeV is an important mile stone. To achieve the mile stone, a high energy test facility called MeV Test Facility (MTF) was constructed. The system consists of a 1 MV, 1 A acceleration power supply and a 100 kW power supply system for negative ion production. Up to now, an H{sup -} ion beam was accelerated up to the energy of 805 keV with an acceleration drain current of 150 mA for 1 s in a five stage electrostatic multi-aperture accelerator. (author)

  10. Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

    2016-11-01

    Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

  11. Fusion driver study. Final technical report, April 1, 1978-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Friedman, H.W.

    1980-04-01

    A conceptual design of a multi-megajoule, repetitively pulsed CO/sub 2/ laser system for Inertial Confinement Fusion is presented. System configurations consisting of 50 to 100 kJ modules operating at subatmospheric pressures with multiple pass optical extraction appear feasible with present or near term technology. Overall laser system efficiencies of greater than 10% at repetition rates in excess of 10 Hz are possible with the state-of-the-art pulsed power technology. The synthesis of all the laser subsystems into a specific configuration for a Laser Fusion Driver depends upon the reactor chamber(s) layout, subsystem reliability and restrictions on overall dimensions of the fusion driver. A design is presented which stacks power amplifier modules in series in a large torus with centrally located reactor chamber. Cost estimates of the overall Laser Fusion Driver are also presented.

  12. Safety training and safe operating procedures written for PBFA (Particle Beam Fusion Accelerator) II and applicable to other pulsed power facilities

    Energy Technology Data Exchange (ETDEWEB)

    Donovan, G.L.; Goldstein, S.A.

    1986-12-01

    To ensure that work in advancing pulsed power technology is performed with an acceptably low risk, pulsed power research facilities at Sandia National Laboratories must satisfy general safety guidelines established by the Department of Energy, policies and formats of the Environment, Safety, and Health (ES and H) Department, and detailed procedures formulated by the Pulsed Power Sciences Directorate. The approach to safety training and to writing safe operating procedures, and the procedures presented here are specific to the Particle Beam Fusion Accelerator II (PBFA II) Facility but are applicable as guidelines to other research and development facilities which have similar hazards.

  13. Microstructural Evolution and Creep-Rupture Behavior of Fusion Welds Involving Alloys for Advanced Ultrasupercritical Power Generation

    Science.gov (United States)

    Bechetti, Daniel H., Jr.

    Projections for large increases in the global demand for electric power produced by the burning of fossil fuels, in combination with growing environmental concerns surrounding these fuel sources, have sparked initiatives in the United States, Europe, and Asia aimed at developing a new generation of coal fired power plant, termed Advanced Ultrasupercritical (A-USC). These plants are slated to operate at higher steam temperatures and pressures than current generation plants, and in so doing will offer increased process cycle efficiency and reduced greenhouse gas emissions. Several gamma' precipitation strengthened Ni-based superalloys have been identified as candidates for the hottest sections of these plants, but the microstructural instability and poor creep behavior (compared to wrought products) of fusion welds involving these alloys present significant hurdles to their implementation and a gap in knowledge that must be addressed. In this work, creep testing and in-depth microstructural characterization have been used to provide insight into the long-term performance of these alloys. First, an investigation of the weld metal microstructural evolution as it relates to creep strength reductions in A-USC alloys INCONELRTM 740, NIMONICRTM 263 (INCONEL and NIMONIC are registered trademarks of Special Metals Corporation), and HaynesRTM 282RTM (Haynes and 282 are registered trademarks of Haynes International) was performed. gamma'-precipitate free zones were identified in two of these three alloys, and their development was linked to the evolution of phases that precipitate at the expense of gamma'. Alloy 282 was shown to avoid precipitate free zone formation because the precipitates that form during long term aging in this alloy are poor in the gamma'-forming elements. Next, the microstructural evolution of INCONELRTM 740H (a compositional variant of alloy 740) during creep was investigated. Gleeble-based interrupted creep and creep-rupture testing was used to

  14. Fifty Years of Magnetic Fusion Research (1958–2008: Brief Historical Overview and Discussion of Future Trends

    Directory of Open Access Journals (Sweden)

    Laila A. El-Guebaly

    2010-06-01

    Full Text Available Fifty years ago, the secrecy surrounding magnetically controlled thermonuclear fusion had been lifted allowing researchers to freely share technical results and discuss the challenges of harnessing fusion power. There were only four magnetic confinement fusion concepts pursued internationally: tokamak, stellarator, pinch, and mirror. Since the early 1970s, numerous fusion designs have been developed for the four original and three new approaches: spherical torus, field-reversed configuration, and spheromak. At present, the tokamak is regarded worldwide as the most viable candidate to demonstrate fusion energy generation. Numerous power plant studies (>50, extensive R&D programs, more than 100 operating experiments, and an impressive international collaboration led to the current wealth of fusion information and understanding. As a result, fusion promises to be a major part of the energy mix in the 21st century. The fusion roadmaps developed to date take different approaches, depending on the anticipated power plant concept and the degree of extrapolation beyond ITER. Several Demos with differing approaches will be built in the US, EU, Japan, China, Russia, Korea, India, and other countries to cover the wide range of near-term and advanced fusion systems.

  15. Fifty years of magnetic fusion research (1958-2008): brief historical overview and discussion of future trends

    Energy Technology Data Exchange (ETDEWEB)

    El-Guebaly, L. A. [University of Wisconsin-Madison, 1500 Engineering Dr. Madison, WI 53706 (United States)

    2010-06-15

    Fifty years ago, the secrecy surrounding magnetically controlled thermonuclear fusion had been lifted allowing researchers to freely share technical results and discuss the challenges of harnessing fusion power. There were only four magnetic confinement fusion concepts pursued internationally: tokamak, stellarator, pinch, and mirror. Since the early 1970s, numerous fusion designs have been developed for the four original and three new approaches: spherical torus, field-reversed configuration, and spheromak. At present, the tokamak is regarded worldwide as the most viable candidate to demonstrate fusion energy generation. Numerous power plant studies (>50), extensive research and development programs, more than 100 operating experiments, and an impressive international collaboration led to the current wealth of fusion information and understanding. As a result, fusion promises to be a major part of the energy mix in the 21{sup st} century. The fusion roadmaps developed to date take different approaches, depending on the anticipated power plant concept and the degree of extrapolation beyond ITER. Several Demos with differing approaches will be built in the US, EU, Japan, China, Russia, Korea, India, and other countries to cover the wide range of near-term and advanced fusion systems. (author)

  16. Advanced latent heat of fusion thermal energy storage for solar power systems

    Science.gov (United States)

    Phillips, W. M.; Stearns, J. W.

    1985-01-01

    The use of solar thermal power systems coupled with thermal energy storage (TES) is being studied for both terrestrial and space applications. In the case of terrestrial applications, it was found that one or two hours of TES could shift the insolation peak (solar noon) to coincide with user peak loads. The use of a phase change material (PCM) is attractive because of the higher energy storage density which can be achieved. However, the use of PCM has also certain disadvantages which must be addressed. Proof of concept testing was undertaken to evaluate corrosive effects and thermal ratcheting effects in a slurry system. It is concluded that the considered alkali metal/alkali salt slurry approach to TES appears to be very viable, taking into account an elimination of thermal ratcheting in storage systems and the reduction of corrosive effects. The approach appears to be useful for an employment involving temperatures applicable to Brayton or Stirling cycles.

  17. Spinal fusion

    Science.gov (United States)

    ... Low back pain - fusion; Herniated disk - fusion; Spinal stenosis - fusion; Laminectomy - fusion ... be done: With other surgical procedures for spinal stenosis , such as foraminotomy or laminectomy After diskectomy in ...

  18. Sequential charged-particle and neutron activation of Flibe in the HYLIFE-II inertial fusion energy power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Latkowski, J.F.; Tobin, M.T. [Lawrence Livermore National Lab., CA (United States); Vujic, J.L. [California Univ., Berkeley, CA (United States); Sanz, J. [Universidad Politecnica de Madrid (Spain)

    1996-06-14

    Most radionuclide generation/depletion codes consider only neutron reactions and assume that charged particles, which may be generated in these reactions, deposit their energy locally without undergoing further nuclear interactions. Neglect of sequential charged-particle (x,n) reactions can lead to large underestimation in the inventories of radionuclides. PCROSS code was adopted for use with the ACAB activation code to enable calculation of the effects of (x,n) reactions upon radionuclide inventories and inventory-related indices. Activation calculations were made for Flibe (2LiF + BeF{sub 2}) coolant in the HYLIFE-II inertial fusion energy (IFE) power plant design. For pure Flibe coolant, it was found that (x,n) reactions dominate the residual contact dose rate at times of interest for maintenance and decommissioning. For impure Flibe, however, radionuclides produced directly in neutron reaction dominate the contact dose rate and (x,n) reactions do not make a significant contribution. Results demonstrate potential importance of (x,n) reactions and that the relative importance of (x,n) reactions varies strongly with the composition of the material considered. Future activation calculations should consider (x,n) reactions until a method for pre-determining their importance is established.

  19. SXR-XUV Diagnostics for Edge and Core of Magnetically Confined Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Stutman, Dan [Johns Hopkins University

    2014-09-10

    The present report summarizes the results obtained during a one-year extension of DoE grant “SXR-XUV Diagnostics for Edge and Core of Magnetically Confined Plasmas”, at Johns Hopkins University, aimed at completing the development of a new type of magnetic fusion plasma diagnostic, the XUV Transmission Grating Imaging Radiometer (TGIR). The TGIR enables simultaneous spatially and spectrally resolved measurements of the XUV/VUV radiated power from impurities in fusion plasmas, with high speed. The instrument was successfully developed and qualified in the laboratory and in experiments on a tokamak. Its future applications will be diagnostic of the impurity content and transport in the divertor and edge of advanced magnetic fusion experiments, such as NSTX Upgrade.

  20. FIREBALL: Fusion Ignition Rocket Engine with Ballistic Ablative Lithium Liner

    Science.gov (United States)

    Martin, Adam K.; Eskridge, Richard H.; Fimognari, Peter J.; Lee, Michael H.

    2006-01-01

    Thermo-nuclear fusion may be the key to a high Isp, high specific power propulsion system. In a fusion system energy is liberated within, and imparted directly to, the propellant. In principle, this can overcome the performance limitations inherent in systems that require thermal power transfer across a material boundary, and/or multiple power conversion stages (NTR, NEP). A thermo-nuclear propulsion system, which attempts to overcome some of the problems inherent in the Orion concept, is described. A dense FRC plasmoid is accelerated to high velocity (in excess of 500 km/s) and is compressed into a detached liner (pulse unit). The kinetic energy of the FRC is converted into thermal and magnetic-field energy, igniting a fusion burn in the magnetically confined plasma. The fusion reaction serves as an ignition source for the liner, which is made out of detonable materials. The energy liberated in this process is converted to thrust by a pusher-plate, as in the classic Orion concept. However with this concept, the vehicle does not carry a magazine of autonomous pulse-units. By accelerating a second, heavier FRC, which acts as a piston, right behind the first one, the velocity required to initiate the fusion burn is greatly reduced.

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

    Energy Technology Data Exchange (ETDEWEB)

    1976-11-01

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

  2. Smart tungsten alloys as a material for the first wall of a future fusion power plant

    Science.gov (United States)

    Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch.; Rasinski, M.; Kreter, A.; Unterberg, B.; Coenen, J. W.; Du, H.; Mayer, J.; Garcia-Rosales, C.; Calvo, A.; Ordas, N.

    2017-06-01

    Tungsten is currently deemed as a promising plasma-facing material (PFM) for the future power plant DEMO. In the case of an accident, air can get into contact with PFMs during the air ingress. The temperature of PFMs can rise up to 1200 °C due to nuclear decay heat in the case of damaged coolant supply. Heated neutron-activated tungsten forms a volatile radioactive oxide which can be mobilized into the atmosphere. New self-passivating ‘smart’ alloys can adjust their properties to the environment. During plasma operation the preferential sputtering of lighter alloying elements will leave an almost pure tungsten surface facing the plasma. During an accident the alloying elements in the bulk are forming oxides thus protecting tungsten from mobilization. Good plasma performance and the suppression of oxidation are required for smart alloys. Bulk tungsten (W)-chroimum (Cr)-titanium (Ti) alloys were exposed together with pure tungsten (W) samples to the steady-state deuterium plasma under identical conditions in the linear plasma device PSI 2. The temperature of the samples was ~576 °C-715 °C, the energy of impinging ions was 210 eV matching well the conditions expected at the first wall of DEMO. Weight loss measurements demonstrated similar mass decrease of smart alloys and pure tungsten samples. The oxidation of exposed samples has proven no effect of plasma exposure on the oxidation resistance. The W-Cr-Ti alloy demonstrated advantageous 3-fold lower mass gain due to oxidation than that of pure tungsten. New yttrium (Y)-containing thin film systems are demonstrating superior performance in comparison to that of W-Cr-Ti systems and of pure W. The oxidation rate constant of W-Cr-Y thin film is 105 times less than that of pure tungsten. However, the detected reactivity of the bulk smart alloy in humid atmosphere is calling for a further improvement.

  3. The role of the National Ignition Facility in the development of inertial fusion energy

    Energy Technology Data Exchange (ETDEWEB)

    Logan, B.G.

    1996-06-01

    The authors have completed a conceptual design for a 1.8-MJ, 500-TW, 0.35-{mu}m solid-state laser system for the National Ignition Facility (NIF), which will demonstrate inertial fusion ignition and gain for national security, energy, and science applications. The technical goal of the U.S. Inertial Confinement Fusion (ICF) Program as stated in the current ICF Five-Year Program Plan is {open_quotes}to produce pure fusion ignition and burn in the laboratory, with fusion yields of 200 to 1000 MJ, in support of three missions: (1) to play an essential role in accessing physics regimes of interest in nuclear weapon design...; (2) to provide an above-ground simulation capability for nuclear weapon effects...; and (3) to develop inertial fusion energy for civilian power production.{close_quotes} This article addresses the third goal-- the development of inertial fusion energy (IFE). This article reports a variety of potential contributions the NIF could make to the development of IFE, drawn from a nationally attended workshop held at the University of California at Berkeley in Feb, 1994. In addition to demonstrating fusion ignition as a fundamental basis for IFE, the findings of the workshop, are that the NIF could also provide important data for target physics and fabrication technology, for IFE target chamber phenomena such as materials responses to target emissions, and for fusion power technology-relevant tests.

  4. Some not such wonderful magnetic fusion facts; and their solution

    Science.gov (United States)

    Manheimer, Wallace

    2017-10-01

    The first not such wonderful fusion fact (NSWFF) is that if ITER is successful, it is nowhere near ready to develop into a DEMO. The design Q=10, along with electricity generating efficiency of 1/3 prevents this. Making it smaller and cheaper, increasing the gain by 3 or 4, and the wall loading by an order of magnitude is not a minor detail, it is not at all clear the success with ITER will lead to a similar, pure fusion DEMO. The second NSWFF is that tokamaks are unlikely to improve to the point where they can be effective fusion reactors because their performance is limited by conservative design rules. The third NSWFF is that developing large fusion devices like ITER takes an enormous amount of time and dollars, there are no second chances. The fourth NSWFF is that it is unlikely that alternative confinement configurations will succeed either, at least in this century; they are simply too far behind. There is only a single solution for fusion to become a sustainable, carbon free power source by midcentury or shortly thereafter. This is to develop ITER (assuming it is successful) into a fusion breeder. This work was not supported by any organization, private or public.

  5. Instability studies on a spherical inertial electrostatic confinement

    Science.gov (United States)

    Kim, Hyung Jin

    The spherical inertial electrostatic confinement concept offers an alternative fusion plasma confinement scheme, where charged particles are accelerated and confined electrostatically with a series of biased spherical concentric electrodes. The inertia of the accelerated ions compresses the ions and builds up the space charge at the center of the cathode grid. The space charge of the ions attracts electrons which in turn accumulate a space charge. The accumulation of collective space charge creates a series of deep "virtual" electrostatic potential wells which confine and concentrate ions into a small volume where an appreciable number of nuclear fusion reactions could occur. It is very attractive for a power plant due to its mechanical simplicity and high power-to-mass ratio. However, its beam-plasma interactions are not clearly understood. In order to evaluate the inertial electrostatic confinement concept, it is essential to develop a reliable and flexible instability analysis method for an equilibrium plasma in a potential well. Subsequently stability of this well can be studied. As a part of this study, methods are sought to avoid or suppress any destructive instabilities. Methods to be explored include modification/control of the well profile, control of the electron to ion beam density ratio, control of the angular momentum of the beam, etc. For this purpose, a perturbative (deltaf) particle simulation techniques for a kinetic analysis is applied to simulate completely the dynamic evolution of perturbed Vlasov-Poisson equations and, in addition, to achieve much more accurate simulations of the nonlinear dynamics using less simulation particles compared to conventional particle-in-cell method. This model is used to study the behavior of two-stream-like instabilities related to the trapped spherically converging ions. Results show that steady-state solutions of the self-consistent Vlasov-Poisson equation in which angular momentum of positively charged particle

  6. Progress in Mirror-Based Fusion Neutron Source Development

    Science.gov (United States)

    Anikeev, A. V.; Bagryansky, P. A.; Beklemishev, A. D.; Ivanov, A. A.; Kolesnikov, E. Yu.; Korzhavina, M. S.; Korobeinikova, O. A.; Lizunov, A. A.; Maximov, V. V.; Murakhtin, S. V.; Pinzhenin, E. I.; Prikhodko, V. V.; Soldatkina, E. I.; Solomakhin, A. L.; Tsidulko, Yu. A.; Yakovlev, D. V.; Yurov, D. V.

    2015-01-01

    The Budker Institute of Nuclear Physics in worldwide collaboration has developed a project of a 14 MeV neutron source for fusion material studies and other applications. The projected neutron source of the plasma type is based on the gas dynamic trap (GDT), which is a special magnetic mirror system for plasma confinement. Essential progress in plasma parameters has been achieved in recent experiments at the GDT facility in the Budker Institute, which is a hydrogen (deuterium) prototype of the source. Stable confinement of hot-ion plasmas with the relative pressure exceeding 0.5 was demonstrated. The electron temperature was increased up to 0.9 keV in the regime with additional electron cyclotron resonance heating (ECRH) of a moderate power. These parameters are the record for axisymmetric open mirror traps. These achievements elevate the projects of a GDT-based neutron source on a higher level of competitive ability and make it possible to construct a source with parameters suitable for materials testing today. The paper presents the progress in experimental studies and numerical simulations of the mirror-based fusion neutron source and its possible applications including a fusion material test facility and a fusion-fission hybrid system. PMID:28793722

  7. Uniformity of fuel target implosion in Heavy Ion Fusion

    CERN Document Server

    Kawata, S; Suzuki, T; Karino, T; Barada, D; Ogoyski, A I; Ma, Y Y

    2015-01-01

    In inertial confinement fusion the target implosion non-uniformity is introduced by a driver beams' illumination non-uniformity, a fuel target alignment error in a fusion reactor, the target fabrication defect, et al. For a steady operation of a fusion power plant the target implosion should be robust against the implosion non-uniformities. In this paper the requirement for the implosion uniformity is first discussed. The implosion uniformity should be less than a few percent. A study on the fuel hotspot dynamics is also presented and shows that the stagnating plasma fluid provides a significant enhancement of vorticity at the final stage of the fuel stagnation. Then non-uniformity mitigation mechanisms of the heavy ion beam (HIB) illumination are also briefly discussed in heavy ion inertial fusion (HIF). A density valley appears in the energy absorber, and the large-scale density valley also works as a radiation energy confinement layer, which contributes to a radiation energy smoothing. In HIF a wobbling he...

  8. Confinement and stability of plasmas with externally driven steady-state elevated q-profiles

    Energy Technology Data Exchange (ETDEWEB)

    Bock, Alexander; Stober, Joerg; Fischer, Rainer; Fable, Emiliano; Reich, Matthias [Max-Planck-Institut fuer Plasmaphysik, Garching bei Muenchen (Germany); Collaboration: ASDEX Upgrade Team

    2015-05-01

    The helicity profile of the magnetic field lines is an important quantity for the operation of Tokamak fusion devices and can be expressed as the so-called safety factor q. It has profound influence on both the stability of the fusion plasma, as well as its confinement properties. Operation scenarios with centrally elevated and flat, or even reversed q-profiles promise fewer central instabilities and better core confinement and are thus considered potentially attractive for future fusion power plants. To verify these predictions, centrally elevated q-profiles are created using external counter current drive, with additional heating power added afterwards to explore the stability limits and transport properties of the resulting plasmas. The tailored q-profiles are calculated using magnetic equilibrium reconstruction constrained by internal motional Stark effect data to confirm to the presence of the desired helicities. They are then used as a basis for simulations of the transport properties with the gyro-Landau-fluid code TGLF. The simulation results are then compared to the experimentally measured kinetic profiles.

  9. Lead (Pb) hohlraum: target for inertial fusion energy.

    Science.gov (United States)

    Ross, J S; Amendt, P; Atherton, L J; Dunne, M; Glenzer, S H; Lindl, J D; Meeker, D; Moses, E I; Nikroo, A; Wallace, R

    2013-01-01

    Recent progress towards demonstrating inertial confinement fusion (ICF) ignition at the National Ignition Facility (NIF) has sparked wide interest in Laser Inertial Fusion Energy (LIFE) for carbon-free large-scale power generation. A LIFE-based fleet of power plants promises clean energy generation with no greenhouse gas emissions and a virtually limitless, widely available thermonuclear fuel source. For the LIFE concept to be viable, target costs must be minimized while the target material efficiency or x-ray albedo is optimized. Current ICF targets on the NIF utilize a gold or depleted uranium cylindrical radiation cavity (hohlraum) with a plastic capsule at the center that contains the deuterium and tritium fuel. Here we show a direct comparison of gold and lead hohlraums in efficiently ablating deuterium-filled plastic capsules with soft x rays. We report on lead hohlraum performance that is indistinguishable from gold, yet costing only a small fraction.

  10. Recent Trends in Fusion Gyrotron Development at KIT

    Directory of Open Access Journals (Sweden)

    Gantenbein G.

    2017-01-01

    Full Text Available ECRH&CD is one of the favorite heating system for magnetically confined nuclear fusion plasmas. KIT is strongly involved in the development of high power gyrotrons for use in ECRH systems for nuclear fusion. KIT is upgrading the sub-components of the existing 2 MW, 170 GHz coaxial-cavity short-pulse gyrotron to support long-pulse operation up to 1 s, all components will be equipped with a specific active cooling system. Two important developments for future high power, highly efficient gyrotrons will be discussed: design of gyrotrons with high operating frequency (∼ 240 GHz and efficiency enhancement by using advanced collector designs with multi-staged voltage depression.

  11. Recent Trends in Fusion Gyrotron Development at KIT

    Science.gov (United States)

    Gantenbein, G.; Avramidis, K.; Franck, J.; Illy, S.; Ioannidis, Z. C.; Jin, J.; Jelonnek, J.; Kalaria, P.; Pagonakis, I. Gr.; Ruess, S.; Rzesnicki, T.; Thumm, M.; Wu, C.

    2017-10-01

    ECRH&CD is one of the favorite heating system for magnetically confined nuclear fusion plasmas. KIT is strongly involved in the development of high power gyrotrons for use in ECRH systems for nuclear fusion. KIT is upgrading the sub-components of the existing 2 MW, 170 GHz coaxial-cavity short-pulse gyrotron to support long-pulse operation up to 1 s, all components will be equipped with a specific active cooling system. Two important developments for future high power, highly efficient gyrotrons will be discussed: design of gyrotrons with high operating frequency (˜ 240 GHz) and efficiency enhancement by using advanced collector designs with multi-staged voltage depression.

  12. Engineering design point for a 1MW fusion neutron source

    Science.gov (United States)

    Sieck, Paul; Melnik, Paul; Woodruff, Simon; Stuber, James; Romero-Talamas, Carlos; O'Bryan, John; Miller, Ronald

    2016-10-01

    Compact fusion neutron sources are currently serving important roles in medical isotope production, and could be used for waste transmutation if sufficient fluence can be attained. The engineering design point for a compact neutron source with target rateof e17n/sbased on the adiabatic compression of a spheromak is presented. The compression coils and passive structure are designed to maintain stability during compression. The power supplies consist of 4 separate banks of MJ each; Pspice simulations and power requirement calculations will be shown. We outline the diagnostic set that will be required for an experimental campaign to address issues relating to both formation efficiency and energy confinement scaling during compression. Work supported in part by DARPA Grant N66001-14-1-4044 and IAEA CRP on compac fusion neutron sources.

  13. Polarization: A Must for Fusion

    Directory of Open Access Journals (Sweden)

    Guidal M.

    2012-10-01

    Full Text Available Recent realistic simulations confirm that the polarization of the fuel would improve significantly the DT fusion efficiency. We have proposed an experiment to test the persistence of the polarization in a fusion process, using a terawatt laser hitting a polarized HD target. The polarized deuterons heated in the plasma induced by the laser can fuse producing a 3He and a neutron in the final state. The angular distribution of the neutrons and the change in the corresponding total cross section are related to the polarization persistence. The experimental polarization of DT fuel is a technological challenge. Possible paths for Magnetic Confinement Fusion (MCF and for Inertial Confinement Fusion (ICF are reviewed. For MCF, polarized gas can be used. For ICF, cryogenic targets are required. We consider both, the polarization of gas and the polarization of solid DT, emphasizing the Dynamic Nuclear polarization (DNP of HD and DT molecules.

  14. Spinal Fusion

    Science.gov (United States)

    ... concept of fusion is similar to that of welding in industry. Spinal fusion surgery, however, does not ... are taking for other conditions, and your overall health can affect the rate of healing and fusion, ...

  15. Membrane tension and membrane fusion

    OpenAIRE

    Kozlov, Michael M.; Chernomordik, Leonid V.

    2015-01-01

    Diverse cell biological processes that involve shaping and remodeling of cell membranes are regulated by membrane lateral tension. Here we focus on the role of tension in driving membrane fusion. We discuss the physics of membrane tension, forces that can generate the tension in plasma membrane of a cell, and the hypothesis that tension powers expansion of membrane fusion pores in late stages of cell-to-cell and exocytotic fusion. We propose that fusion pore expansion can require unusually la...

  16. COST-EFFECTIVE TARGET FABRICATION FOR INERTIAL FUSION ENERGY

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-03-01

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

  17. Recent results of H-mode confinement study in JT-60U (April-September, 1995)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    Improvement in the performance of energy confinement is one of the most important issues to realize thermonuclear fusion reactors. The H-mode is one of excellent improved confinement modes. From the view point of steady-state operation, the ELMy H-mode is considered to be a principal operation mode in ITER. For the engineering design of the ITER, there still remain issues to be clarified on the H-mode characteristics. These issues are required to be studied on the present tokamaks as ITER physics research needs. In order to satisfy the above request, experiments of the H-mode confinement have been carried out on JT-60U. Recent results of H-mode confinement study in JT-60U during April to September, 1995 are summarized in the present report. The scaling of high T{sub i} H-mode confinement is described in section 2. The time behaviour of transport properties are shown in sections 3 and 4. Result of the non-dimensional transport experiment is presented in section 5. The H-mode transition is investigated in sections 6, 7, 8 and 9; threshold power scaling, parametric study on edge local quantities, effect of edge neutrals, and H-L back transition. The onset condition of ELMs is studied in section 10. (author).

  18. Particle beam fusion progress report for 1989

    Energy Technology Data Exchange (ETDEWEB)

    Sweeney, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States). Pulsed Power Sciences Center

    1994-08-01

    This report summarizes the progress on the pulsed power approach to inertial confinement fusion. In 1989, the authors achieved a proton focal intensity of 5 TW/cm{sup 2} on PBFA-II in a 15-cm-radius applied magnetic-field (applied-B) ion diode. This is an improvement by a factor of 4 compared to previous PBFA-II experiments. They completed development of the three-dimensional (3-D), electromagnetic, particle-in-cell code QUICKSILVER and obtained the first 3-D simulations of an applied-B ion diode. The simulations, together with analytic theory, suggest that control of electromagnetic instabilities could reduce ion divergence. In experiments using a lithium fluoride source, they delivered 26 kJ of lithium energy to the diode axis. Rutherford-scattered ion diagnostics have been developed and tested using a conical foil located inside the diode. They can now obtain energy density profiles by using range filters and recording ion images on nuclear track recording film. Timing uncertainties in power flow experiments on PBFA-II have been reduced by a factor of 5. They are investigating three plasma opening switches that use magnetic fields to control and confine the injected plasma. These new switches provide better power flow than the standard plasma erosion switch. Advanced pulsed-power fusion drivers will require extraction-geometry applied-B ion diodes. During this reporting period, progress was made in evaluating the generation, transport, and focus of multiple ion beams in an extraction geometry and in assessing the probable damage to a target chamber first wall.

  19. Neutron and gamma flux distributions and their implications for radiation damage in the shielded superconducting core of a fusion power plant

    Science.gov (United States)

    Windsor, Colin G.; Morgan, J. Guy

    2017-11-01

    The neutron and gamma ray fluxes within the shielded high-temperature superconducting central columns of proposed spherical tokamak power plants have been studied using the MCNP Monte-Carlo code. The spatial, energy and angular variations of the fluxes over the shield and superconducting core are computed and used to specify experimental studies relevant to radiation damage and activation. The mean neutron and gamma fluxes, averaged over energy and angle, are shown to decay exponentially through the shield and then to remain roughly constant in the core region. The mean energy of neutrons is shown to decay more slowly than the neutron flux through the shield while the gamma energy is almost constant around 2 MeV. The differential neutron and gamma fluxes as a function of energy are examined. The neutron spectrum shows a fusion peak around 1 MeV changing at lower energies into an epithermal E -0.85 variation and at thermal energies to a Maxwellian distribution. The neutron and gamma energy spectra are defined for the outer surface of the superconducting core, relevant to damage studies. The inclusion of tungsten boride in the shield is shown to reduce energy deposition. A series of plasma scenarios with varying plasma major radii between 0.6 and 2.5 m was considered. Neutron and gamma fluxes are shown to decay exponentially with plasma radius, except at low shield thickness. Using the currently known experimental fluence limitations for high temperature superconductors, the continuous running time before the fluence limit is reached has been calculated to be days at 1.4 m major radius increasing to years at 2.2 m. This work helps validate the concept of the spherical tokamak route to fusion power by demonstrating that the neutron shielding required for long lifetime fusion power generation can be accommodated in a compact device.

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

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K. M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, C. W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M; Bell, M. G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C. S.; Cheng, C. Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D. S; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P. C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Fujita, T.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Hawryluk, R. J.; Heidbrink, W.; Herrmann, H. W.; Hill, K. W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Loughlin, M. J.; Machuzak, J.; Majeski, R.; Mansfield, D. K.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M. P.; McChesney, J.; McCormack, B.; McCune, D. C.; McKee, G.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Norris, M.; O`Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Parks, P.; Paul, S. F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G.; Scillia, R.; Scott, S. D.; Semenov, I.; Senko, T.; Sesnic, S.; Sissingh, R.; Skinner, C. H.; Snipes, J.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Tighe, W.; Timberlake, J. R.; Tobita, K.; Towner, H. H.; Tuszewski, M.; Halle, A. Von; Vannoy, C.; Viola, M.; Goeler, S. Von; Voorhees, D.; Walters, R. T.; Wester, R.; White, R.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Winston, J.; Wright, K.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zavereev, V.; Zweben, S. J.

    1995-01-01

    The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≈ 2.8 MW m₋3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni (0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

  1. Understanding fuel magnetization and mix using secondary nuclear reactions in magneto-inertial fusion.

    Science.gov (United States)

    Schmit, P F; Knapp, P F; Hansen, S B; Gomez, M R; Hahn, K D; Sinars, D B; Peterson, K J; Slutz, S A; Sefkow, A B; Awe, T J; Harding, E; Jennings, C A; Chandler, G A; Cooper, G W; Cuneo, M E; Geissel, M; Harvey-Thompson, A J; Herrmann, M C; Hess, M H; Johns, O; Lamppa, D C; Martin, M R; McBride, R D; Porter, J L; Robertson, G K; Rochau, G A; Rovang, D C; Ruiz, C L; Savage, M E; Smith, I C; Stygar, W A; Vesey, R A

    2014-10-10

    Magnetizing the fuel in inertial confinement fusion relaxes ignition requirements by reducing thermal conductivity and changing the physics of burn product confinement. Diagnosing the level of fuel magnetization during burn is critical to understanding target performance in magneto-inertial fusion (MIF) implosions. In pure deuterium fusion plasma, 1.01 MeV tritons are emitted during deuterium-deuterium fusion and can undergo secondary deuterium-tritium reactions before exiting the fuel. Increasing the fuel magnetization elongates the path lengths through the fuel of some of the tritons, enhancing their probability of reaction. Based on this feature, a method to diagnose fuel magnetization using the ratio of overall deuterium-tritium to deuterium-deuterium neutron yields is developed. Analysis of anisotropies in the secondary neutron energy spectra further constrain the measurement. Secondary reactions also are shown to provide an upper bound for the volumetric fuel-pusher mix in MIF. The analysis is applied to recent MIF experiments [M. R. Gomez et al., Phys. Rev. Lett. 113, 155003 (2014)] on the Z Pulsed Power Facility, indicating that significant magnetic confinement of charged burn products was achieved and suggesting a relatively low-mix environment. Both of these are essential features of future ignition-scale MIF designs.

  2. A GDT-based fusion neutron source for academic and industrial applications

    Science.gov (United States)

    Anderson, J. K.; Forest, C. B.; Mirnov, V. V.; Peterson, E. E.; Waleffe, R.; Wallace, J.; Harvey, R. W.

    2017-10-01

    The design of a fusion neutron source based on the gas dynamic trap (GDT) configuration is underway. The motivation is both the ends and the means. There are immediate applications for neutrons including medical isotope production and actinide burners. Taking the next step in the magnetic mirror path will leverage advances in high-temperature superconducting magnets and additive manufacturing in confining a fusion plasma, and both the technological and physics bases exist. Recent breakthrough results at the GDT facility in Russia demonstrate stable confinement of a beta 60% mirror plasma at high Te ( 1 keV). These scale readily to a fusion neutron source with an increase in magnetic field, mirror ratio, and ion energy. Studies of a next-step compact device focus on calculations of MHD equilibrium and stability, and Fokker-Planck modeling to optimize the heating scenario. The conceptualized device uses off-the-shelf MRI magnets for a 1 T central field, REBCO superconducting mirror coils (which can currently produce fields in excess of 30T), and existing 75 keV NBI and 140 GHz ECRH. High harmonic fast wave injection is damped on beam ions, dramatically increasing the fusion reactivity for an incremental bump in input power. MHD stability is achieved with the vortex confinement scheme, where a biasing profile imposes optimal ExB rotation of the plasma. Liquid metal divertors are being considered in the end cells. Work supported by the Wisconsin Alumni Research Foundation.

  3. Materials-related issues in the safety and licensing of nuclear fusion facilities

    Science.gov (United States)

    Taylor, N.; Merrill, B.; Cadwallader, L.; Di Pace, L.; El-Guebaly, L.; Humrickhouse, P.; Panayotov, D.; Pinna, T.; Porfiri, M.-T.; Reyes, S.; Shimada, M.; Willms, S.

    2017-09-01

    Fusion power holds the promise of electricity production with a high degree of safety and low environmental impact. Favourable characteristics of fusion as an energy source provide the potential for this very good safety and environmental performance. But to fully realize the potential, attention must be paid in the design of a demonstration fusion power plant (DEMO) or a commercial power plant to minimize the radiological hazards. These hazards arise principally from the inventory of tritium and from materials that become activated by neutrons from the plasma. The confinement of these radioactive substances, and prevention of radiation exposure, are the primary goals of the safety approach for fusion, in order to minimize the potential for harm to personnel, the public, and the environment. The safety functions that are implemented in the design to achieve these goals are dependent on the performance of a range of materials. Degradation of the properties of materials can lead to challenges to key safety functions such as confinement. In this paper the principal types of material that have some role in safety are recalled. These either represent a potential source of hazard or contribute to the amelioration of hazards; in each case the related issues are reviewed. The resolution of these issues lead, in some instances, to requirements on materials specifications or to limits on their performance.

  4. Method of controlling fusion reaction rates

    Science.gov (United States)

    Kulsrud, Russell M.; Furth, Harold P.; Valeo, Ernest J.; Goldhaber, Maurice

    1988-01-01

    A method of controlling the reaction rates of the fuel atoms in a fusion reactor comprises the step of polarizing the nuclei of the fuel atoms in a particular direction relative to the plasma confining magnetic field. Fusion reaction rates can be increased or decreased, and the direction of emission of the reaction products can be controlled, depending on the choice of polarization direction.

  5. Controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

  6. Magnetic-mirror principle as applied to fusion research

    Energy Technology Data Exchange (ETDEWEB)

    Post, R.F.

    1983-08-11

    A tutorial account is given of the key physics issues in the confinement of high temperature plasma in magnetic mirror systems. The role of adiabatic invariants and particle drifts and their relationship to equilibrium and stability are discussed, in the context of the various forms of mirror field geometry. Collisional effects and the development and the control of ambipolar potentials are reviewed. The topic of microinstabilities is discussed together with the means for their control. The properties and advantages for fusion power purposes of various special embodiments of the mirror idea, including tandem mirrors, are discussed.

  7. Novel, spherically-convergent ion systems for neutron source and fusion energy production

    Science.gov (United States)

    Barnes, D. C.; Nebel, R. A.; Ribe, F. L.; Schauer, M. M.; Schranck, L. S.; Umstadter, K. R.

    1999-06-01

    Combining spherical convergence with electrostatic or electro-magnetostatic confinement of a nonneutral plasma offers the possibility of high fusion gain in a centimeter-sized system. The physics principles, scaling laws, and experimental embodiments of this approach are presented. Steps to development of this approach from its present proof-of-principle experiments to a useful fusion power reactor are outlined. This development path is much less expensive and simpler, compared to that for conventional magnetic confinement and leads to different and useful products at each stage. Reactor projections show both high mass power density and low to moderate wall loading. This approach is being tested experimentally in PFX-I (Penning Fusion eXperiment-Ions), which is based on the following recent advances: 1) Demonstration, in PFX (our former experiment), that it is possible to combine nonneutral electron plasma confinement with nonthermal, spherical focussing; 2) Theoretical development of the POPS (Periodically Oscillating Plasma Sphere) concept, which allows spherical compression of thermal-equilibrium ions; 3) The concept of a massively-modular approach to fusion power, and associated elimination of the critical problem of extremely high first wall loading. PFX-I is described. PFX-I is being designed as a small (<1.5 cm) spherical system into which moderate-energy electrons (up to 100 kV) are injected. These electrons are magnetically insulated from passing to the sphere and their space charge field is then used to spherically focus ions. Results of initial operation with electrons only are presented. Deuterium operation can produce significant neutron output with unprecedented efficiency (fusion gain Q).

  8. EURATOM-CEA association contributions to the 18. IAEA fusion energy conference

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Peysson, Y.; Hoang, G.T. [and others

    2000-12-01

    The 9 contributions of EURATOM-Cea association to the fusion energy conference hold at Sorrento are gathered in this document with 7 additional papers. The different titles are: 1) Ergodic divertor experiments on the route to steady state operation of Tore-Supra, 2) High power lower hybrid current drive experiments in Tore-Supra tokamak, 3) Electron transport and improved confinement on Tore-Supra, 4) ECRH experiments and developments for long pulse in Tore-Supra, 5) Impurity penetration and contamination in Tore-Supra ergodic divertor experiments, 6) Real time plasma feed-back control: an overview of Tore-Supra achievements, 7) Numerical assessment of the ion turbulent thermal transport scaling laws, 8) Design of next step tokamak: consistent analysis of plasma flux consumption and poloidal, 9) Large superconducting conductors and joints for fusion magnets: from conceptual design to test at full size scale, 10) Burst-prone transport in tokamaks with internal transport barriers, 11) Electrostatic turbulence with finite parallel correlation length and radial electric field generation, 12) Theoretical issues in tokamak confinement: internal-edge transport barriers and runaway avalanche confinement, 13) Core and edge confinement studies with different heating methods in JET, 14) Confinement and transport studies of conventional scenarios in ASDEX upgrade, 15) First test results for the ITER central solenoid model coil, and 16) Progress of the ITER central solenoid model coil program.

  9. FIRE, A Next Step Option for Magnetic Fusion

    Energy Technology Data Exchange (ETDEWEB)

    Meade, D.M.

    2002-09-12

    The next major frontier in magnetic fusion physics is to explore and understand the strong nonlinear coupling among confinement, MHD stability, self-heating, edge physics, and wave-particle interactions that is fundamental to fusion plasma behavior. The Fusion Ignition Research Experiment (FIRE) Design Study has been undertaken to define the lowest cost facility to attain, explore, understand, and optimize magnetically confined fusion-dominated plasmas. The FIRE is envisioned as an extension of the existing Advanced Tokamak Program that could lead to an attractive magnetic fusion reactor. The FIRE activities have focused on the physics and engineering assessment of a compact, high-field tokamak with the capability of achieving Q approximately equal to 10 in the ELMy H-mode for a duration of about 1.5 plasma current redistribution times (skin times) during an initial burning-plasma science phase, and the flexibility to add Advanced Tokamak hardware (e.g., lower-hybrid current drive) later. The configuration chosen for FIRE is similar to that of ARIES-RS, the U.S. Fusion Power Plant study utilizing an Advanced Tokamak reactor. The key ''Advanced Tokamak'' features are: strong plasma shaping, double-null pumping divertors, low toroidal field ripple (<0.3%), internal control coils, and space for wall stabilization capabilities. The reference design point is R subscript ''o'' = 2.14 m, a = 0.595 m, B subscript ''t''(R subscript ''o'') = 10 T, I subscript ''p'' = 7.7 MA with a flattop time of 20 s for 150 MW of fusion power. The baseline magnetic fields and pulse lengths can be provided by wedged BeCu/OFHC toroidal-field (TF) coils and OFHC poloidal-field (PF) coils that are pre-cooled to 80 K prior to the pulse and allowed to warm up to 373 K at the end of the pulse. A longer-term goal of FIRE is to explore Advanced Tokamak regimes sustained by noninductive current drive

  10. Massachusetts Institute of Technology, Plasma Fusion Center, Technical Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Davidson, Ronald C.

    1980-08-01

    A review is given of the technical programs carried out by the Plasma Fusion Center. The major divisions of work areas are applied plasma research, confinement experiments, fusion technology and engineering, and fusion systems. Some objectives and results of each program are described. (MOW)

  11. Fabrication of polystyrene hollow microspheres as laser fusion ...

    Indian Academy of Sciences (India)

    Centre for Advanced Technology (CAT), Indore 452 013, India. MS received 26 June 2001; revised 7 December 2001. Abstract. Inertial confinement fusion, frequently referred to as ICF, inertial fusion, or laser fu- sion, is a means of producing energy by imploding small hollow microspheres containing thermo- nuclear fusion ...

  12. Fusion in the energy system

    DEFF Research Database (Denmark)

    Fusion energy is the fundamental energy source of the Universe, as the energy of the Sun and the stars are produced by fusion of e.g. hydrogen to helium. Fusion energy research is a strongly international endeavor aiming at realizing fusion energy production in power plants on Earth. Reaching...... this goal, mankind will have a sustainable base load energy source with abundant resources, having no CO2 release, and with no longlived radioactive waste. This presentation will describe the basics of fusion energy production and the status and future prospects of the research. Considerations...... of integration into the future electricity system and socio-economic studies of fusion energy will be presented, referring to the programme of Socio-Economic Research on Fusion (SERF) under the European Fusion Energy Agreement (EFDA)....

  13. Planning for U.S. Fusion Community Participation in the ITER Program

    Energy Technology Data Exchange (ETDEWEB)

    Baker, Charles [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Berk, Herbert [Univ. of Texas, Austin, TX (United States); Greenwald, Martin [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mauel, Michael E. [Columbia Univ., New York, NY (United States); Najmabadi, Farrokh [Univ. of California, San Diego, CA (United States); Nevins, William M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Stambaugh, Ronald [General Atomics, La Jolla, CA (United States); Synakowski, Edmund [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Batchelor, Donald B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fonck, Raymond [Univ. of Wisconsin, Madison, WI (United States); Hawryluk, Richard J. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Meade, Dale M. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Neilson, George H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Parker, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Strait, Ted [General Atomics, La Jolla, CA (United States)

    2006-06-07

    A central step in the mission of the U.S. Fusion Energy Sciences program is the creation and study of a fusion-powered "star on earth", where the same energy source that drives the sun and other stars is reproduced and controlled for sustained periods in the laboratory. This “star” is formed by an ionized gas, or plasma, heated to fusion temperatures in a magnetic confinement device known as a tokamak, which is the most advanced magnetic fusion concept. The ITER tokamak is designed to be the premier scientific tool for exploring and testing expectations for plasma behavior in the fusion burning plasma regime, wherein the fusion process itself provides the dominant heat source to sustain the plasma temperature. It will provide the scientific basis and control tools needed to move toward the fusion energy goal. The ITER project confronts the grand challenge of creating and understanding a burning plasma for the first time. The distinguishing characteristic of a burning plasma is the tight coupling between the fusion heating, the resulting energetic particles, and the confinement and stability properties of the plasma. Achieving this strongly coupled burning state requires resolving complex physics issues and integrating challenging technologies. A clear and comprehensive scientific understanding of the burning plasma state is needed to confidently extrapolate plasma behavior and related technology beyond ITER to a fusion power plant. Developing this predictive understanding is the overarching goal of the U.S. Fusion Energy Sciences program. The burning plasma research program in the U.S. is being organized to maximize the scientific benefits of U.S. participation in the international ITER experiment. It is expected that much of the research pursued on ITER will be based on the scientific merit of proposed activities, and it will be necessary to maintain strong fusion research capabilities in the U.S. to successfully contribute to the success of ITER and optimize

  14. Office of Fusion Energy Sciences. A ten-year perspective (2015-2025)

    Energy Technology Data Exchange (ETDEWEB)

    None

    2015-12-01

    The vision described here builds on the present U.S. activities in fusion plasma and materials science relevant to the energy goal and extends plasma science at the frontier of discovery. The plan is founded on recommendations made by the National Academies, a number of recent studies by the Fusion Energy Sciences Advisory Committee (FESAC), and the Administration’s views on the greatest opportunities for U.S. scientific leadership.This report highlights five areas of critical importance for the U.S. fusion energy sciences enterprise over the next decade: 1) Massively parallel computing with the goal of validated whole-fusion-device modeling will enable a transformation in predictive power, which is required to minimize risk in future fusion energy development steps; 2) Materials science as it relates to plasma and fusion sciences will provide the scientific foundations for greatly improved plasma confinement and heat exhaust; 3) Research in the prediction and control of transient events that can be deleterious to toroidal fusion plasma confinement will provide greater confidence in machine designs and operation with stable plasmas; 4) Continued stewardship of discovery in plasma science that is not expressly driven by the energy goal will address frontier science issues underpinning great mysteries of the visible universe and help attract and retain a new generation of plasma/fusion science leaders; 5) FES user facilities will be kept world-leading through robust operations support and regular upgrades. Finally, we will continue leveraging resources among agencies and institutions and strengthening our partnerships with international research facilities.

  15. A Plan for the Development of Fusion Energy. Final Report to Fusion Energy Sciences Advisory Committee, Fusion Development Path Panel

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2003-03-05

    This report presents a plan for the deployment of a fusion demonstration power plant within 35 years, leading to commercial application of fusion energy by mid-century. The plan is derived from the necessary features of a demonstration fusion power plant and from the time scale defined by President Bush. It identifies critical milestones, key decision points, needed major facilities and required budgets.

  16. Development of Laser Based Plasma Diagnostics for Fusion Research on NSTX-U

    Science.gov (United States)

    Barchfeld, Robert Adam

    Worldwide demand for power, and in particular electricity, is growing. Increasing population, expanding dependence on electrical devices, as well as the development of emerging nations, has created significant challenges for the power production. Compounding the issue are concerns over pollution, natural resource supplies, and political obstacles in troubled parts of the world. Many believe that investment in renewable energy will solve the expected energy crisis; however, renewable energy has many shortfalls. Consequently, additional sources of energy should be explored to provide the best options for the future. Electricity from fusion power offers many advantages over competing technologies. It can potentially produce large amounts of clean energy, without the serious concerns of fission power plant safety and nuclear waste. Fuel supplies for fusion are plentiful. Fusion power plants can be operated as needed, without dependence on location, or local conditions. However, there are significant challenges before fusion can be realized. Many factors currently limit the effectiveness of fusion power, which prevents a commercial power plant from being feasible. Scientists in many countries have built, and operate, experimental fusion plants to study the fusion process. The leading examples are magnetic confinement reactors known as tokamaks. At present, reactor gain is near unity, where the fusion power output is nearly the same as the power required to operate the reactor. A tenfold increase in gain is what reactors such as ITER hope to achieve, where 50 MW will be used for plasma heating, magnetic fields, and so forth, with a power output of 500 MW. Before this can happen, further research is required. Loss of particle and energy confinement is a principal cause of low performance; therefore, increasing confinement time is key. There are many causes of thermal and particle transport that are being researched, and the prime tools for conducting this research are

  17. Rugged Packaging for Damage Resistant Inertial Fusion Energy Optics

    Energy Technology Data Exchange (ETDEWEB)

    Stelmack, Larry

    2003-11-17

    The development of practical fusion energy plants based on inertial confinement with ultraviolet laser beams requires durable, stable final optics that will withstand the harsh fusion environment. Aluminum-coated reflective surfaces are fragile, and require hard overcoatings resistant to contamination, with low optical losses at 248.4 nanometers for use with high-power KrF excimer lasers. This program addresses the definition of requirements for IFE optics protective coatings, the conceptual design of the required deposition equipment according to accepted contamination control principles, and the deposition and evaluation of diamondlike carbon (DLC) test coatings. DLC coatings deposited by Plasma Immersion Ion Processing were adherent and abrasion-resistant, but their UV optical losses must be further reduced to allow their use as protective coatings for IFE final optics. Deposition equipment for coating high-performance IFE final optics must be designed, constructed, and operated with contamination control as a high priority.

  18. Characteristics of a high-power RF source of negative hydrogen ions for neutral beam injection into controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Abdrashitov, G. F.; Belchenko, Yu. I.; Gusev, I. A.; Ivanov, A. A.; Kondakov, A. A.; Sanin, A. L.; Sotnikov, O. Z., E-mail: O.Z.Sotnikov@inp.nsk.su; Shikhovtsev, I. V. [Russian Academy of Sciences, Budker Institute of Nuclear Physics, Siberian Branch (Russian Federation)

    2017-01-15

    An injector of hydrogen atoms with an energy of 0.5–1 MeV and equivalent current of up to 1.5 A for purposes of controlled fusion research is currently under design at the Budker Institute of Nuclear Physics, Siberian Branch, Russian Academy of Sciences. Within this project, a multiple-aperture RF surface-plasma source of negative hydrogen ions is designed. The source design and results of experiments on the generation of a negative ion beam with a current of >1 A in the long-pulse mode are presented.

  19. Osteoclast Fusion

    DEFF Research Database (Denmark)

    Marie Julie Møller, Anaïs; Delaissé, Jean-Marie; Søe, Kent

    2017-01-01

    suggesting that fusion partners may specifically select each other and that heterogeneity between the partners seems to play a role. Therefore, we set out to directly test the hypothesis that fusion factors have a heterogenic involvement at different stages of nuclearity. Therefore, we have analyzed...... on the nuclearity of fusion partners. While CD47 promotes cell fusions involving mono-nucleated pre-osteoclasts, syncytin-1 promotes fusion of two multi-nucleated osteoclasts, but also reduces the number of fusions between mono-nucleated pre-osteoclasts. Furthermore, CD47 seems to mediate fusion mostly through......Investigations addressing the molecular keys of osteoclast fusion are primarily based on end-point analyses. No matter if investigations are performed in vivo or in vitro the impact of a given factor is predominantly analyzed by counting the number of multi-nucleated cells, the number of nuclei per...

  20. Advanced particle-in-cell simulation techniques for modeling the Lockheed Martin Compact Fusion Reactor

    Science.gov (United States)

    Welch, Dale; Font, Gabriel; Mitchell, Robert; Rose, David

    2017-10-01

    We report on particle-in-cell developments of the study of the Compact Fusion Reactor. Millisecond, two and three-dimensional simulations (cubic meter volume) of confinement and neutral beam heating of the magnetic confinement device requires accurate representation of the complex orbits, near perfect energy conservation, and significant computational power. In order to determine initial plasma fill and neutral beam heating, these simulations include ionization, elastic and charge exchange hydrogen reactions. To this end, we are pursuing fast electromagnetic kinetic modeling algorithms including a two implicit techniques and a hybrid quasi-neutral algorithm with kinetic ions. The kinetic modeling includes use of the Poisson-corrected direct implicit, magnetic implicit, as well as second-order cloud-in-cell techniques. The hybrid algorithm, ignoring electron inertial effects, is two orders of magnitude faster than kinetic but not as accurate with respect to confinement. The advantages and disadvantages of these techniques will be presented. Funded by Lockheed Martin.