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  • Articles  (5)
  • fusion  (5)
  • 2010-2014
  • 1990-1994  (4)
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  • 1925-1929
  • 1993  (4)
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  • Energy, Environment Protection, Nuclear Power Engineering  (5)
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  • Articles  (5)
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  • 2010-2014
  • 1990-1994  (4)
  • 1980-1984  (1)
  • 1975-1979
  • 1925-1929
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  • 1
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 2 (1982), S. 131-143 
    ISSN: 1572-9591
    Keywords: fusion ; direct energy conversion ; neutral beams
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract A direct-energy converter was developed for use on neutral-beam injectors. The purpose of the converter is to raise the efficiency of the injector by recovering the portion of the ion beam not converted to neutrals. In addition to increasing the power efficiency, direct conversion reduces the requirements on power supplies and eases the beam dump problem. The converter was tested at Lawrence Berkeley Laboratory on a reduced-area version of a neutral-beam injector developed for use on the Tokamak Fusion Test Reactor at Princeton. The conversion efficiency of the total ion power was 65 ±7% at the beginning of the pulse, decaying to just over 50% by the end of the 0.6-s pulse. Once the electrode surfaces were conditioned, the decay was due to the rise in pressure of only the beam gas and not to outgassing. The direct converter was tested with 1.7 A of hydrogen ions and with 1.5 A of helium ions through the aperture with similar efficiencies. At the midplane through the beam, the line power density was 0.7 MW/m, for comparison with our calculations of slab beams and the prediction of 2–4 MW/m in some reactor studies. Over 98 kV was developed at the ion collector when the beam energy was 100 keV. When electrons were suppressed magnetically, rather than electrostatically, the efficiency dropped to 40%. However, a better designed electron catcher could improve this efficiency. New electrode material released gas (mostly H2 and CO) in amounts that exceeded the input of primary gas from the beam. The electrodes were all made of 0.51-mm-thick molybdenum cooled only by radiation. This allowed the heating by the beam to outgas the electrodes and for them to stay hot enough to avoid the reabsorption of gas between shots. By minor redesign of the electrodes, adding cryopanels near the electrodes, and grounding the ion source, these results extrapolate with high confidence to an efficiency of 70–80% at a power density of 2–4 MW/m. Higher power may be possible with magnetic electron suppression.
    Type of Medium: Electronic Resource
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  • 2
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 12 (1993), S. 209-214 
    ISSN: 1572-9591
    Keywords: fusion ; radioactive waste ; final disposal
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Within the European Fusion Technology programs Studsvik RadWaste AB has performed studies on fusion waste treatment and disposal for several years. This paper deals with the treatment and geological disposal of radioactive waste from NET operation and decommissioning. Results from calculations on radioactive waste fluxes for the operation and decommissioning of NET are reported. The calculations are based on the NET predesign report published 1993 and include results for the exchangeable in-vessel and external parts of the machine as well as permanent reactor components. Different aspects of treatment, packaging, transportation, and interim storage of the waste are discussed. The volumes of waste conditioned for final disposal are preliminarily quantified, according to German and Swedish scenarios for radioactive waste disposal. A total repository volume of approximately 45,000 m3 is required in the German Scenario and 35,000 m3 is required in the Swedish Scenario. Results from dose rate calculations for NET waste in final repositories are presented for the Swedish Scenario. This work was financially supported by the Swedish Natural Science Research Council (NFR) and the European Atomic Energy Community, under an association contract between Euratom and Sweden.
    Type of Medium: Electronic Resource
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  • 3
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 12 (1993), S. 41-45 
    ISSN: 1572-9591
    Keywords: fusion ; blanket ; passive safety
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accidents such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This paper proposes a new helium-cooled, tritium breeding blanket concept which uses a metallic structure, and which performs significantly better during such accidents than related designs. The proposed blanket uses modified, reduced-activation HT-9 steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. This concept uses novel features such as: (1) a “beryllium-joint” design which allows beryllium to be used to conduct heat away from the first wall, while accommodating swelling of the beryllium, and (2) a shield cooled by naturally circulating water. These features help the blanket passively withstand a worst-case undercooling accident scenario.
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  • 4
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 12 (1993), S. 47-51 
    ISSN: 1572-9591
    Keywords: fusion ; reliability ; failure rates
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Fusion facility safety and reliability/availability analyses require accurate component failure rate information to provide meaningful results. While fission reactor operating experience data may be adequate for some types of components, there are some data needs that are fusion-specific, such as tritium fueling and handling system information. This paper summarizes data analysis of tritium glovebox confinement systems and an air detritiation system from the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. These analyses benefit fusion design work by highlighting weak areas in designs to allow for modifications and upgrades, making future designs more robust. The TSTA results are generally smaller failure rates than the information from other industries, thus showing one of the benefits of gathering these fusion-specific data.
    Type of Medium: Electronic Resource
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  • 5
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 12 (1993), S. 83-87 
    ISSN: 1572-9591
    Keywords: fusion ; water cooling ; loss of coolant accidents ; containment ; pressure transients
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract The NET cooling systems for in-vessel components and vessel are generally based on low pressure and low temperature water. However, the cooling loops for the breeder blanket are intended to operate at a water temperature of about 250°C. A pipe break in a loop with such data would pressurize the compartment where the break takes place. Therefore, as a basis for proper compartment design, it is important to analyze possible pressure increases following pipe breaks. It may also be necessary to introduce equipment for pressure relief or pressure suppression. The objective of the parameter study presented is to determine the relationship between allowed maximum containment pressure following postulated large pipe break in breeding blanket loop and required containment volume. Parameters varied are: blanket loop temperature and pressure (within the range of burn and baking), and pressure suppression system inclusion/exclusion. The analysis has been performed by means of the Swedish containment code COPTA. The results of the analysis are summarized in a plot showing the influence of the varied parameters on required containment volume. In addition, the results presented include required vent areas, heat sink capacities, etc.
    Type of Medium: Electronic Resource
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