Author: R. Carbonneau
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ISBN:
Category :
Languages : en
Pages : 0
Book Description
A proposal for the irradiation of fuel elements clad in a zirconium base alloy at rating and maximum temperature representative of a superheat power reactor
Author: R. Carbonneau
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Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
A proposal for the first comparative irradiation of various high performance zirconium alloy clad fuel at conditions representative of low temperature dryout
Author: R. H. Hu
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Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
The first comparative irradiation of various high performance zirconium alloy clad fuel elements, at conditions representative of low temperature dryout in a fuel bundle
Author: R. H. Hu
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
A proposal for the first irradiaiton of fuel elements clad in zirconium base alloys operating at 500 degrees c in superheated steam
Author: Z. Domaratzki
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ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
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Category :
Languages : en
Pages : 0
Book Description
Zirconium Moderator Cladding Irradiated in the Sodium Reactor Experiment
Author: J. J. Gill
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 42
Book Description
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 42
Book Description
Irradiation data and post-irradiation examination of zirconium alloy clad fuel elements inadvertently operated at high sheath temperatures
Author: E. Proudfoot
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ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Hydrofluoric Acid Decladding of Zirconium-clad Power Reactor Fuel Elements
Author: W. E. Clark
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ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 22
Book Description
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 22
Book Description
Irradiation of U-Mo Base Alloys
Author: M. P. Johnson
Publisher:
ISBN:
Category : Molybdenum alloys
Languages : en
Pages : 38
Book Description
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
Publisher:
ISBN:
Category : Molybdenum alloys
Languages : en
Pages : 38
Book Description
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
A proposal for the irradiation of zircaloy-4 clad, uo2 fuel elements at post-dryout sheath temperatures in the 625-675 degrees c range in the x-4 loop
Author: V. J. Langman
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
A proposal to irradiate u-si-al alloy fuel clad in zr-1.2 wt percent cr-0.1 wt percent fe and zr-2.5 wt percent nb at 500 degrees c sheath temperatures in superheated steam
Author: G. R. Dimmick
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description