Fundamental and Applied Research and Development in Metallurgy PDF Download

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Languages : en
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A total of 35 specimens of U-Pu-fissium alloy and 2 specimens of U-10 wt% Pu-5 wt% Mo alloy were irradiated as a part of the fue1-alloy development program for fast breeder reactors at Argonne National Laboratory. Total atom burnups ranged from 1.0 to 1.8% at maximum fuel temperatures ranging from 230 to 470 deg C. Emphasis was placed on the EBR-II Core-III reference fuel material, which is an injection-cast, U-20 wt% Pu-10 wt% fissium alloy. lt was found that this material begins to swell catastrophically at irradiation temperatures above 370 deg C. The ability of the fuel to resist swelling did not appear to vary appreciably with minor changes in Zr or fissium content. Decreasing the Pu to 10 wt%, however, significantly improved the swelling behavior of the alloy. Both pourcast and thermally cycled material and pour-cast, extruded, and thermally cycled material appeared to be more stable under irradiation than injection-cast material. Under comparable irradiation conditions, the specimens of U-20 wt% Pu- 5 wt% Mo alloy were less dimensionally stable than the U-Pu-fissium alloys investigated. (auth).

Author: A. R. Kaufmann
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Category : Uranium alloys
Languages : en
Pages : 88

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Languages : en
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Irradiations were made on small specimens of U-5 wt.% Zr-1.5 wt.% Nb alloy with a wide variety of fabrication histories and heat treatments in order to determine the optimum heat treatment for the fuel plates for the Experimental Boiling Water Reactor (EBW). In the time available, a heat treatment could not be found which simultaneously conferred dimensional stability and corrosion resistance to the alloy. The most effective heat treatment for dimensionally stabilizing swaged or round-rolled material was a 24-hr isothermal transformation from the gamma phase at 650 deg C. This heat treatment was subsequently used as a basis for the heat-treatment specifications for the EBWR fuel plates. In later studies on specimens cut from plates it was learned that the alloy could be adequately stabilized against irradiation growth and also made corrosion- resistent by first reducing the plate 12% in thickness by cold rolling followed by a 24-hr isothermal transformation from the gamma phase at 665 deg C, and finally quenching from 800 deg C. Irradiation growth rates of plate specimens were effectively reduced by the presence of metallurgically bonded Zircaloy-2 cladding. Flat-rolled material under irradiation generally increased in length and width and decreased in thickness. (auth).

Author: M. P. Johnson
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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: A. A. Bauėr
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Category : Nuclear fuel elements
Languages : en
Pages : 150

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Category : Uranium as fuel
Languages : en
Pages : 140

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Languages : en
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Author: J. A. Horak
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Category : Alloys
Languages : en
Pages : 40

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A total of 35 specimens of U-Pu-fissium alloy and 2 specimens of U-10 wt% Pu-5 wt% Mo alloy were irradiated as a part of the fuel-alloy development program for fast breeder reactors at Argonne National Laboratory. Total atom burnups ranged from 1.0 to 1.8% at maximum fuel temperatures ranging from 230 to 470 deg C. Emphasis was placed on the EBR-II Core-III reference fuel material, which is an injection-cast, U-20 wt% Pu-10 wt% fissium alloy. It was found that this material begins to swell catastrophically at irradiation temperatures above 370 deg C. The ability of the fuel to resist swelling did not appear to vary appreciably with minor changes in zirconium or fissium content. Decreasing the Pu to 10 wt%, however, significantly improved the swelling behavior of the alloy. Both pour-cast and thermally cycled material and pour-cast, extruded, and thermally cycled material appeared to be more stable under irradiation than injection-cast material. Under comparable irradiation conditions, the specimens of U-20 wt% Pu- 5 wt% Mo alloy were less dimensionally stable than the U-Pu-fissium alloys investigated.

Author: John J. Burke
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ISBN:
Category : Materials
Languages : en
Pages : 1034

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Languages : en
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A study was made of the phase relations and the properties of uranium-- fissium alloys which have compositions bracketing that intpnded for the first core loading of Experimental Breeder Reactor II. The fissium aggregate in the alloys consisted of the elements Zr, Nb, Mo, Ru, Rh, and Pd. Phase relations are shown to parallel closely those in the dominant U--Mo--Ru ternary system. The uranium gamma phase is stabilized down to 552 deg C, while the beta phase is entirely suppressed at high fissium contents. Certain crystallographic data are given and the minor phases that occur in the alloys are identified. In cast and gammaquenched alloys the retention of the high-temperature gamma phase produced low hardness and low density. The thermal expsnsion behavior of the alloys is shown to be dependent upon composition and prior thermal history. Thermal conductivity data are presented for uranium and the uranium-- fission alloys. The thermal conductivities of the alloys decrease with increasing fissium concentration. (auth).