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  • 1
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 49-58 
    ISSN: 1572-9591
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Type of Medium: Electronic Resource
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  • 2
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 15-48 
    ISSN: 1572-9591
    Keywords: Plasma fueling ; pellet injection ; tokamak
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract The injection of frozen pellets composed of the isotopes of hydrogen has become the leading candidate for refueling fusion power reactors based on the tokamak concept. This lofty position has been reached partly as a result of efforts to find an attractive solution to the perplexing problem of depositing atoms of fuel deep within the magnetically confined, hot plasma, and because of some recent experimental successes. To some extent, the relative merits of this technique will depend upon the distance that the cryogenic pellet will penetrate such a plasma, and the early exploratory research has addressed this problem on both theoretical and experimental fronts. The conclusion from the theoretical effort is that a protective blanket consisting of hydrogenic gas or cold plasma will envelope the pellet and partially shield the surface from the intense plasma heat flux. The blanket prolongs pellet lifetime, but penetration to the plasma center might require pellet injection velocities in excess of 10 km/s. The need for central penetration has not yet been established either theoretically or experimentally. The experiments performed to date have verified the existence of a shielding mechanism in general, and pellet ablation models that incorporate neutral gas shielding in particular are in adequate agreement with the experiments. Magnetic shielding effects are expected to contribute to, but not dominate, self-shielding in the higher plasma temperature regimes of the future. The tokamak plasma has demonstrated a surprising resilience even to massive density perturbations caused by the large refueling pellets used in present experiments. The characteristic discharge behavior is qualitatively not unlike that observed with gas puffing; but, for the first time, central plasma fueling has been studied, and this does not appear to be superior to refueling by partial pellet penetration. If relatively large pellets containing a significant fraction of the total plasma charge are acceptable in the present resistive plasma regimes, then it can be argued that they should have little impact on the gross stability of a hot thermonuclear tokamak plasma. Large pellets are preferable from the standpoint of attaining deep penetration, and this has important implications for the technology of pellet injection. The interesting velocity regime of 1 km/s has already been achieved with simple gun-type devices and this should be adequate for near-term tokamak experiments. Further improvements are anticipated, but the 10 km/s and above regime is uncertain; and, if current theory and experiments extrapolate to the future, such velocities might be desirable but unnecessary.
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  • 3
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 59-67 
    ISSN: 1572-9591
    Keywords: Fusion-fission ; breeder ; tokamak breeder ; decentralized system
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract A decentralized nuclear energy system is proposed comprising mass-produced pressurized water reactors in the size range 10 to 300 MW (thermal), to be used for the production of process heat, space heat, and electricity in applications where petroleum and natural gas are presently used. Special attention is given to maximizing the refueling interval with no interim batch shuffling in order to minimize fuel transport, reactor downtime, and opportunity for fissile diversion. The smallest reactors could be deployed as “nuclear batteries,” kept in the equivalent of spent-fuel shipping casks and returned to nuclear fuel centers for refueling. These objectives demand a substantial fissile enrichment (7 to 15%). The preferred fissile fuel is U-233, which offers an order of magnitude savings in ore requirements (compared with U-235 fuel), and whose higher conversion ratio in thermal reactors serves to extend the period of useful reactivity and relieve demand on the fissile breeding plants (compared with Pu-239 fuel). Application of the neutral-beam-driven tokamak fusion-neutron source to a U-233 breeding pilot plant is examined. This scheme can be extended in part to a decentralized fusion energy system, wherein remotely located large fusion reactors supply excess tritium to a distributed system of relatively smallnonbreeding D-T reactors.
    Type of Medium: Electronic Resource
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  • 4
    ISSN: 1572-9591
    Keywords: Ignition ; copper magnets ; tokamak
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Design considerations have been developed for a compact ignition test reactor (CITR). The objectives of this tokamak device are to achieve ignition, to study the characteristics of plasmas that are self-heated by alpha particles, and to investigate burn control. To achieve a compact design, the toroidal field magnet consists of copper-stainless steel plates to accommodate relatively high stresses; it is inertially cooled by liquid nitrogen. No neutron shielding is provided between the plasma and the toroidal field magnet. The flat-top of the toroidal field magnet is ∼10 s. Strong auxiliary heating is employed. In one design option, adiabatic compression in major radius is employed to reduce the neutral beam energy required for adequate penetration; thiscompression boosted design option has a horizontally elongated vacuum chamber; illustrative parameters are a compressed plasma witha=0.50 m, R=1.35 m,B T =9.1 T, and a neutral beam power of ∼15 MW of 160 keVD 0 beams. A design option has also been developed for alarge bore device, which utilizes a circular vacuum chamber. Thelarge bore design provides increased margin and flexibility; both direct heating with RF or neutral beam injection and compression boosted startup are possible. The large bore design also facilitates the investigation of high-Q driven operation. Illustrative plasma parameters for full use of the large bore area=0.85 m,R=1.90 m, andB T =7.5 T.
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  • 5
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 211-213 
    ISSN: 1572-9591
    Keywords: thermonuclear fusion ; bose degeneracy ; WKB ; intermolecular barrier ; tunneling ; deuterium molecule ; zero-point energy
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract An estimate is made of the fusion probability for an aggregate of D2 molecules that have suffered bose degeneracy. A WKB analysis for the tunneling of deuterons through the electron intermolecular barrier gives the penetration rate $$R = (2E/h)\exp \left[ { - \frac{8}{3}\sqrt {\frac{{2mV_0 x_0^2 }}{\hbar }} \left( {1 - \frac{E}{{V_0 }}} \right)^{3/2} } \right]$$ In this expression,V 0 andx 0 are the potential barrier height and width, respectively, andE is the zero point energy of a deuteron in its ground molecular state. With molecules assumed effectively to be adjacent in the bose condensate, the number of fusion events per mole is found to be vanishingly small.
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  • 6
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 217-217 
    ISSN: 1572-9591
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Type of Medium: Electronic Resource
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  • 7
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 253-257 
    ISSN: 1572-9591
    Keywords: Pellet refueling ; drag effect ; penetration depth ; ablation models
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract A refueling pellet is subjected mainly to two kinds of drags: (1) inertial drag caused by the motion of the pellet relative to the surrounding plasma, and (2) ablation drag caused by an uneven ablation rate of the front and the rear surface of the pellet in an inhomogeneous plasma. Computational results showed that for reasonable combinations of pellet size and injection speed, the drag effect is hardly detectable for plasma conditions prevailing in current large tokamaks.
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  • 8
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 307-307 
    ISSN: 1572-9591
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Type of Medium: Electronic Resource
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  • 9
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 341-356 
    ISSN: 1572-9591
    Keywords: EBT reactor ; volumetric efficiency ; field symmetrizing coils ; aspect ratio enhancement
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Optimization of the vacuum magnetic field of an ELMO Bumpy Torus (EBT) reactor is investigated. Several methods of improving reactor volume utilization and single particle confinement are analyzed. These include the use of (a) a large number of sectors and/or a large mirror ratio, (b) high field Nb3Sn or Nb3Sn/NbTi hybrid mirror coils, (c) split-wedge mirror coils, (d) axis-encircling aspect ratio enhancement (ARE) coils, and (e) recently developed field symmetrizing (SYM) coils. Of these, particle drift orbit and three-dimensional tensor pressure equilibrium calculations show that the use of SYM coils in conjunction with high field mirror magnets offers the most promise of good plasma performance in reactors that are smaller (by up to 50%) than previous reference designs that did not employ supplementary coils. Aspect ratio enhancement coils also offer an attractive alternative for improved confinement, but they do not have many of the advantages of SYM coils, particularly for reactor applications. Split-wedge mirror coils improve volume utilization and trapped particle confinement, but they do not enhance the confinement of transitional and passing particles. High field magnets improve confinement by permitting a larger mirror ratio and a larger plasma radius by virtue of their smaller cross-sectional area and higher current density. The relative merits of each magnetics configuration are discussed, including the effects on single particle confinement, reactor volume utilization, materials requirements, engineering design considerations, and reactor assembly, maintenance, and accessibility.
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  • 10
    ISSN: 1572-9591
    Keywords: Fusion reactors ; hybrid reactors ; Monte Carlo method ; three-dimensional calculations ; breeding ratio ; breeding blankets
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Optimization of fissile and fusile production in the SOLASE-H laser-fusion fissile-enrichment fuel-factory blanket is carried out. The objective is maximizing fissile breeding with the constraints of maintaining self-sufficiency in tritium production, and realistically accounting in the modeling for structural and coolant compositions and configurations imposed by the thermal-hydraulic and mechanical designs. The effect of radial and axial blanket zone thicknesses on fusile and fissile breeding is studied using a procedure which modifies the zones' effective optical thicknesses, rather than the actual three-dimensional geometrical configurations. A tritium yield per source neutron of 1.08 and a Th (n, γ) reaction yield per source neutron of 0.43 can be obtained in such a concept, where ThO2 Zircaloy-clad fuel assemblies for light water reactors (LWRs) are enriched in the233U isotope by irradiating them in a lead flux trap. This corresponds to 0.77 kg/[MW(th)-year] of fissile fuel production, and 1.94 years of irradiation in the fusion reactor to attain an average 3 w/o fissile enrichment in the fuel assemblies. For a once-through LWR cycle, a support ratio of 2–3 is estimated. However, with fuel recycling, more attractive support ratios of 4–6 may be attainable for a conversion ratio of 0.55, and of 5–8 for a conversion ratio of 0.70. These estimates are lower than those reported, around 20, for related designs.
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