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Laser Fusion: The First Ten Years 1962-1972

Description: This account of the beginning of the program on laser fusion at Livermore in 1962, and its subsequent development during the decade ending in 1972, was originally prepared as a contribution to the January 1991 symposium 'Achievements in Physics' honoring Professor Keith Brueckner upon his retirement from the University of San Diego at La Jolla. It is a personal recollection of work at Livermore from my vantage point as its scientific leader, and of events elsewhere that I thought significant. This period was one of rapid growth in which the technology of high-power short-pulse lasers needed to drive the implosion of thermonuclear fuel to the temperature and density needed for ignition was developed, and in which the physics of the interaction of intense light with plasmas was explored both theoretically and experimentally.
Date: July 6, 2006
Creator: Kidder, R. E.
Partner: UNT Libraries Government Documents Department

Tritium-Management Requirements for D-T Fusion Reactors (ETF, INTOR, FED)

Description: The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design.
Date: October 1981
Creator: Finn, P. A.; Clemmer, Robert G. & Misra, B.
Partner: UNT Libraries Government Documents Department

The TRIO Experiment

Description: The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.
Date: September 1984
Creator: Clemmer, Robert G.
Partner: UNT Libraries Government Documents Department

Design of an electronic charged particle spectrometer to measure ({rho}R), yield, and implosion symmetry on the OMEGA Upgrade

Description: The preliminary design for a state-of-the-art diagnostic that will measure a broad energy spectrum of charged particles generated in the OMEGA Upgrade facility is investigated. Using a set of photodiodes ({approximately}10) and a 0.8 Tesla permanent magnet, the diagnostic will uniquely determine particle energies and identities from 0.2 MeV up to the maximum charged particle energies (10.6 MeV tritons, 12.5 MeV deuterons and 17.4 MeV protons). With its high density picture elements, each photodiode has 10{sup 6} single-hit detectors, giving the spectrometer a dynamic range of 1 {minus} 10{sup 5} particles/shot. For example, in the case of a DT yield of 10{sup 9} neutrons, about 100 knock-on charged particles will be detected when the spectrometer aperture is 60 cm from the implosion. Furthermore, the measurement of knock-on D and T spectra will allow {rho}R`s up to 0.15 g/cm{sup 2} to be measured (for a 1 keV plasma), or 0.3 g/cm{sup 2}2 if hydrogen doping is used. In addition, the yield and slowing down of secondary protons may be used to determine {rho}R up to 0.3 g/cm{sup 2}. Significantly, this diagnostic will also directly measure the DD fusion yield and energy degradation of nascent 3 MeV protons. By using two such compact spectrometers to measure the yield and spectra on widely separated ports around the OMEGA Upgrade target chamber, the implosion and bum symmetry can be determined. Furthermore, the ion temperature, and, in principle, even the electron temperature can be measured. The diagnostic and its development will be fully tested at several critical steps, utilizing 0.2-16 MeV protons (and several other charged particles and neutrons) from our absolutely calibrated Cockcroft-Walton facility.
Date: November 1, 1994
Creator: Hicks, D.G.; Li, C.K.; Petrasso, R.D.; Wenzel, K.W. & Knauer, J.P.
Partner: UNT Libraries Government Documents Department

Transfer operations with tritium -- A review

Description: Controlled thermonuclear reactors will involve pumping operations with tritium that may involve pressures ranging from submillipascals to megapascals. A variety of pumps is available that can cover portions of this range, and these can be staged to cove the entire pressure range. Some of these pumps can be adapted to virtually any size requirement currently anticipated. Special attention must be paid to operating features and construction materials.
Date: December 31, 1975
Creator: Folkers, C.L. & Gede, V.P.
Partner: UNT Libraries Government Documents Department

Recent high-speed ballistics experiments at ORNL

Description: Oak Ridge National Laboratory (ORNL) has been developing pellet injectors for plasma fueling experiments on magnetic confinement devices for almost 20 years. With these devices, pellets (1 to 8 mm in diameter) composed of hydrogen isotopes are formed (at temperatures <20 K) and typically accelerated to speeds of {approximately} 1.0 to 2.0 km/s for injection into plasmas of experimental fusion devices. A variety of pellet injector designs have been developed at ORNL, including repeating pneumatic injectors (single- and multiple-barrel light gas guns) that can inject up to hundreds of pellets for long-pulse plasma operation. The repeating pneumatic injectors are of particular importance because long-pulse fueling is required for present large experimental fusion devices, with steady-state operation the objective for future fusion reactors. In this paper, recent advancements in the development of repeating pneumatic injectors are described, including (1) a small-bore (1.8-mm), high-firing-rate (10-Hz) version of a single-stage light gas gun; (2) a repeating single-stage light gas gun for 8-mm-diam tritium pellets; (3) a repeating two-stage light gas gun for operation at higher pellet velocities; and (4) a steady-state hydrogen extruder feed system.
Date: December 31, 1994
Creator: Combs, S.K.; Gouge, M.J.; Baylor, L.R.; Fisher, P.W.; Foster, C.A.; Foust, C.R. et al.
Partner: UNT Libraries Government Documents Department

Tritium experience in the Tokamak Fusion Test Reactor

Description: Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.
Date: July 1, 1998
Creator: Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N. et al.
Partner: UNT Libraries Government Documents Department

Analysis of integrating sphere performance for IR enhanced DT layering

Description: Absorbed IR energy can supplement the beta decay energy from DT ice to improve the driving force toward uniform layers. A significant problem with this approach has been to deliver the added IR energy with sufficient uniformity to enhance rather than destroy the uniformity of the ice layers. Computer modeling has indicated that one can achieve {approximately}1% uniformity in the angular variation of the absorbed power using an integrating sphere containing holes large enough to allow external inspection of the ice layer uniformity. The power required depends on the integrating sphere size, a 25 mm diameter sphere requires {approximately}35 mW of IR to deposit as much energy in the ice as the 50 mW/cm{sup 3}(35 pW total) received from tritium decay in DT. Power absorbed in the plastic can cause unacceptable ice-layer non-uniformities for the integrating sphere design considered here.
Date: June 1, 1997
Creator: Stephens, R.B., & Collins, G.W.
Partner: UNT Libraries Government Documents Department

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

Description: Fundamental physics issues facing development of fusion power on a small-scale are assessed with emphasis on the idea of Inertial Electrostatic Confinement (IEC). The authors propose a new concept of accelerator-driven IEC fusion, termed Converging Beam Inertial Electrostatic Confinement (CB-IEC). CB-IEC offers a number of innovative features that make it an attractive pathway toward resolving fundamental physics issues and assessing the ultimate viability of the IEC concept for power generation.
Date: December 8, 1999
Creator: Bauer, T. H. & Wigeland, R. A.
Partner: UNT Libraries Government Documents Department

Cryogenic Distillation: a Fuel Enrichment System for Near-Term Tokamak-Type D-T Fusion Reactors

Description: The successful operation and economic viability of deuterium-tritium- (D-T-) fueled tokamak-type commercial power fusion reactors will depend to a large extent on the development of reliable tritium-containment and fuel-recycle systems. Of the many operating steps in the fuel recycle scheme, separation or enrichment of the isotropic species of hydrogen by cryogenic distillation is one of the most important. A parametric investigation was carried out to study the effects of the various operating conditions and the composition of the spent fuel on the degree of separation. A computer program was developed for the design and analysis of a system of interconnected distillation columns for isotopic separation such that the requirements of near-term D-T-fueled reactors are met. The analytical results show that a distillation cascade consisting of four columns is capable of reprocessing spent fuel varying over a wide range of compositions to yield reinjection-grade fuel with essentially unlimited D/T ratio.
Date: February 1980
Creator: Misra, B. & Davis, J. F.
Partner: UNT Libraries Government Documents Department

Computational modeling of pulsed-power-driven magnetized target fusion experiments

Description: Direct magnetic drive using electrical pulsed power has been considered impractically slow for traditional inertial confinement implosion of fusion targets. However, if the target contains a preheated, magnetized plasma, magnetothermal insulation may allow the near-adiabatic compression of such a target to fusion conditions on a much slower time scale. 100-MJ-class explosive flux compression generators with implosion kinetic energies far beyond those available with conventional fusion drivers, are an inexpensive means to investigate such magnetized target fusion (MTF) systems. One means of obtaining the preheated and magnetized plasma required for an MTF system is the recently reported {open_quotes}MAGO{close_quotes} concept. MAGO is a unique, explosive-pulsed-power driven discharge in two cylindrical chambers joined by an annular nozzle. Joint Russian-American MAGO experiments have reported D-T neutron yields in excess of 10{sup 13} from this plasma preparation stage alone, without going on to the proposed separately driven NM implosion of the main plasma chamber. Two-dimensional MED computational modeling of MAGO discharges shows good agreement to experiment. The calculations suggest that after the observed neutron pulse, a diffuse Z-pinch plasma with temperature in excess of 100 eV is created, which may be suitable for subsequent MTF implosion, in a heavy liner magnetically driven by explosive pulsed power. Other MTF concepts, such as fiber-initiated Z-pinch target plasmas, are also being computationally and theoretically evaluated. The status of our modeling efforts will be reported.
Date: August 1, 1995
Creator: Sheehey, P.; Kirkpatrick, R. & Lindemuth, I.
Partner: UNT Libraries Government Documents Department

Frozen ammonia micropellet generator for Baseball II-T

Description: A ''startup'' plasma at the center of the Baseball II-T magnet was studied. This startup plasma will be used as a target for high energy neutral beams to achieve the required build-up. The target plasma will be created by irradiating a solid pellet with a laser beam. Although a deuterium pellet would be superior because of purity, the development of an ammonia pellet was undertaken because it requires a simpler technology. The ammonia target plasma is physically acceptable for the initial experiment. A frozen ammonia pellet, about 100 $mu$m in diameter, will be irradiated with 300-J CO$sub 2$ laser, to produce a density of about 10$sup 13$ cm$sup -3$ and about 1 kV temperature. (auth)
Date: October 28, 1975
Creator: Denhoy, B.S.
Partner: UNT Libraries Government Documents Department

Pellet trajectory correction

Description: The injection of frozen ammonia droplets in the Baseball II magnetic confinement system is discussed. Two ways to correct for random particle trajectories are described: (1) design non-turbulent orifices, and (2) design a trajectory correction system. (MOW)
Date: November 18, 1975
Creator: Bogdanoff, A.
Partner: UNT Libraries Government Documents Department


Description: Magnetized target fusion (MTF) takes advantage of (1) the reduction of the electron thermal conductivity in a plasma due to magnetization and (2) the efficient heating through bulk compression. MTF proposes to create a warm plasma with an embedded magnetic field and to compress it using an imploded liner or shell. The minimum energy required for fusion in an optimized target is directly proportional to the mass of the ignited fusion fuel. Simple theoretical arguments and parameter studies have demonstrated that MTF has the potential for significantly reducing the power and intensity of a target driver needed to achieve fusion. In order to acquire a comprehensive understanding of MTF and its potential applications it is prudent to develop more complete and reliable computational techniques. This paper briefly reviews the progress toward that goal.
Date: January 1, 2001
Creator: Kirkpatrick, R. C. (Ronald C.); Lindemuth, I. R. (Irvin R.); Barnes, D. C. (Daniel C.); Faehl, R. J. (Rickey J.); Sheehey, P. T. (Peter T.) & Knapp, C. E. (Charles E.)
Partner: UNT Libraries Government Documents Department

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

Description: The Laser Inertial Fusion Energy (LIFE) concept is being designed to operate as either a pure fusion or hybrid fusion-fission system. A key component of a LIFE engine is the fusion chamber subsystem. The present work details the chamber design for the pure fusion option. The fusion chamber consists of the first wall and blanket. This integrated system must absorb the fusion energy, produce fusion fuel to replace that burned in previous targets, and enable both target and laser beam transport to the ignition point. The chamber system also must mitigate target emissions, including ions, x-rays and neutrons and reset itself to enable operation at 10-15 Hz. Finally, the chamber must offer a high level of availability, which implies both a reasonable lifetime and the ability to rapidly replace damaged components. An integrated LIFE design that meets all of these requirements is described herein.
Date: December 7, 2010
Creator: Latkowski, J F; Kramer, K J; Abbott, R P; Morris, K R; DeMuth, J; Divol, L et al.
Partner: UNT Libraries Government Documents Department

TAE modes and MHD activity in TFTR DT plasmas

Description: The high power deuterium and tritium experiments on TFTR have produced fusion a parameters similar to those expected on ITER. The achieved {beta}{sub {alpha}}/{beta} and the R{triangledown}{beta}{sub {alpha}} in TFRR D-T shots are 1/2 to 1/3 those predicted in the ITER EDA. Studies of the initial TFTR D-T plasmas find no evidence that the presence of the fast fusion {alpha} population has affected the stability of MHD, with the possible exception of Toroidal Alfven Eigenmodes (TAE`s). The initial TFTR DT plasmas had MHD activity similar to that commonly seen in deuterium plasmas. Operation of TFTR at plasma currents of 2.0--2.5 MA has greatly reduced the deleterious effects of MHD commonly observed at lower currents. Even at these higher currents, the performance of TFTR is limited by {beta}-limit disruptions. The effects of MHD on D-T fusion {alpha}`s was similar to effects observed on other fusion products in D only plasmas.
Date: March 1, 1995
Creator: Fredrickson, E.; Batha, S. & Bell, M.
Partner: UNT Libraries Government Documents Department

Stability time of a DT-filled cryogenic ICF target in a high vacuum environment

Description: Following the successful pressure loading with DT of a thin-walled plastic inertial fusion target shell (such as those designed for use at the OMEGA facility at the University of Rochester`s Laboratory for Laser Energetics (UR/LLE)), continual care must be taken to safeguard the shell from being exposed to unacceptable pressure differentials across its wall. In particular, once the DT has been condensed into a liquid or solid phase and the outside pressure has been reduced, the target must be maintained below some upper cutoff temperature such that the vapor pressure of the DT is below the bursting pressure for the shell. Through the process of {beta}-decay the DT self-heats, but while the shell is in a high vacuum environment (P {much_lt} 0.8 Pa (6 mtorr) for the OMEGA layering sphere) there is only a negligible heat loss mechanism. This will cause the temperature to increase. A calculation has been done to estimate the rate of temperature increase of the loaded target under high vacuum conditions. A functional form for calculating the target`s temperature increase given its starting temperature is presented. An overall result is that under high vacuum conditions the DT changes from a solid at 10 K to a liquid at 37 K (T{sub c} = 39.4 K) in about 19 minutes. This holding time is significantly less if the initial temperature is higher, the initial state is liquid, or the upper allowed temperature is lower. Simplifying assumptions which were made and their impact on interpreting the results of this calculation are discussed.
Date: December 1, 1998
Creator: Ebey, P.S. & Hoffer, J.K.
Partner: UNT Libraries Government Documents Department

A fusion power plant without plasma-material interactions

Description: A steady-state fusion power plant is described which avoids the deleterious plasma-material interactions found in D-T fueled tokamaks. It is based on driven p-{sup 11}B fusion in a high-beta closed-field device, the field-reversed configuration (FRC), anchored in a gas-dynamic trap (GDT). The plasma outflow on the open magnetic-field lines is cooled by radiation in the GDT, then channeled through a magnetic nozzle, promoting 3-body recombination in the expansion region. The resulting supersonic neutral exhaust stream flows through a turbine, generating electricity.
Date: April 1, 1997
Creator: Cohen, S.A.
Partner: UNT Libraries Government Documents Department