Reduced Enrichment for Research and Test Reactors PDF Download

Author: Peter von der Hardt
Publisher: Springer
ISBN:
Category : Science
Languages : en
Pages : 446

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Proceedings of an International Meeting, Petten, The Netherlands, October 14-16, 1985

Author: United States. Congress. House. Committee on Science and Technology. Subcommittee on Energy Development and Applications
Publisher:
ISBN:
Category : Nuclear engineering
Languages : en
Pages : 1968

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Author: M. P. Johnson
Publisher:
ISBN:
Category : Molybdenum alloys
Languages : en
Pages : 38

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A series of experiments was designed to assess the suitability of uranium-molybdenum alloys as high-temperature, high-burnup fuels for advanced sodium cooled reactors. Specimens with molybdenum contents between 3 and 10% were subjected to capsule irradiation tests in the Materials Testing Reactor, to burnups up to 10,000 Mwd/MTU at temperatures between 800 and 1500 deg F. The results indicated that molybdenum has a considerable effect in reducing the swelling due to irradiation. For example. 3% molybdemum reduces the swelling from 25%, for pure uranium. to 7% at approximates 3,000 Mwd/MTU at 1270 deg F. Further swelling resistance can be gained by increasing the molybdenum content, but the amount gained becomes successively smaller. At higher irradiation levels, the amount of swelling rapidly becomes greater, and larger amounts of molybdenum are required to provide similar resistance. A limit of 7% swelling, at 900 deg F and an irradiation of 7,230 Mwd/ MTU, requires the use of 10% Nonemolybdenum in the alloy. The burnup rates were in the range of 2.0 to 4.0 x 10p13s fissiom/cc-sec. Small ternary additions of silicon and aluminum were shown to have a noticeable effect in reducing swelling when added to a U-3% Mo alloy base. Under the conditions of the present experiment, 0.26% silicon or 0.38% aluminum were equivalent to 1 to 1 1/2% molybdenum. The Advanced Sodium Cooled Reactor requires a fuel capable of being irradiated to 20,000 Mwd/MTU at temperatures up to 1500 deg C in metal fuel, or equivalent in ceramic fuel. It is concluded that even the highest molybdenum contents considered did not produce a fuel capable of operating satisfactorily under these conditions. The alloys would be useful, however, for less exacting conditions. The U-3% Mo alloy is capable of use up to 3,000 Mwd/MTU at temperatures of 1300 deg F before swelling becomes excessive. The addition of silicon and aluminum would increase this limit to at least 3,000 Mwd/MTU, and possibly more if the

Author: P. J. Peterson
Publisher:
ISBN:
Category : Graphite as fuel
Languages : en
Pages : 50

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Two type-AUC graphite fuel elements loaded by solution impregnation to an average concentration of 0.115 g/cc of 93.13% enriched U converted to UC and UC2 were irradiated at temperatures of about 1500 deg C to a 10.2% maximum burnup, corresponding to an irradiation level of 219 kwh/cc or 2.45 x 101 fissions/cc of fuel element. Post-irradiation measurements of the elements showed dimensional changes of -4.3 and -4.8% with the grain, and --0.8 to -2.5% across the grain. Weight losses were 3.2 and 5.1% for the individual elements with approximately 11% of the total U being lost from the elements. With-the- grain thermal conductivity at nominal room temperature was reduced by a factor of approximates 7 and electrical conductivities by factors of 3.4 to 8.3, also at room temperature. Impact strength appeared to be somewhat improved by irradiation. Migration of U within the element was detected by radiographic density observations but not evaluated quantitatively. As anticipated, fission product release was high.

Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 14

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Uranium bismuth solutions or dispersions of USn3 in liquid bismuth- lead-tin are proposed as fluid fuels for a power breeder or converter. These fuels offer advantages of high specific power, unlimited burn up, good neutron economy, and low over all operating costs which might lead to truly economical nuclear power. A simple and effective chemical process based on the extraction of liquid metal with fused salts will remove fission products in a concentrated form for use or storage. A process for the continuous separation of U233 from thorium in the blanket is described. Several design possibilities are discussed.

Author: James A. Lane
Publisher:
ISBN:
Category : Fluid fuel reactors
Languages : en
Pages : 1016

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