A proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2 PDF Download

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A proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2

A proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2 PDF Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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A proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2

A proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2 PDF Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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Book Description

Addendum to the proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(24) n/m2

Addendum to the proposal to irradiate fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(24) n/m2 PDF Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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Addendum to proposal exp-NRX-53901 : the irradiation of fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2

Addendum to proposal exp-NRX-53901 : the irradiation of fuel elements containing unirradiated uo2 pellets clad in zirconium alloy tubes previously irradiated to 8 x 10 sup(20) n/cm2 PDF Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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Book Description

Irradiation of UO2 Fuel Rods

Irradiation of UO2 Fuel Rods PDF Author: John D. Eichenberg
Publisher:
ISBN:
Category : Nuclear fuel rods
Languages : en
Pages : 28

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The irradiation of stainless steel clad uo2 fuel elements in the NRX x-4 steam-cooled loop. (i.e. 260/408). a final proposal to irradiate fuel elements containing annular pellets. appendix : a preliminary proposal for the irradiation of 0.57 in. diameter stainless steel tubes for subsequent out-of-pile Creep rupture tests

The irradiation of stainless steel clad uo2 fuel elements in the NRX x-4 steam-cooled loop. (i.e. 260/408). a final proposal to irradiate fuel elements containing annular pellets. appendix : a preliminary proposal for the irradiation of 0.57 in. diameter stainless steel tubes for subsequent out-of-pile Creep rupture tests PDF Author: H. N. Jones
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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A proposal to irradiate three zirconium alloy clad uo2 fuel elements in the x-6 loop of NRX using the he-3 power cycling facility

A proposal to irradiate three zirconium alloy clad uo2 fuel elements in the x-6 loop of NRX using the he-3 power cycling facility PDF Author: P. J. Fehrenbach
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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Proposal to irradiate zircaloy-clad fuel elements containing uo2 pellets of different densities

Proposal to irradiate zircaloy-clad fuel elements containing uo2 pellets of different densities PDF Author: F. R. Campbell
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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Final Report

Final Report PDF Author: J. W. Weber
Publisher:
ISBN:
Category :
Languages : en
Pages : 58

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Irradiation of zircaloy clad uo2 fuel elements containing discs of different materials and geometries to study their efficiency in cooling the fuel pellet

Irradiation of zircaloy clad uo2 fuel elements containing discs of different materials and geometries to study their efficiency in cooling the fuel pellet PDF Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

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Book Description

Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III

Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III PDF Author: J. H. Kittel
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 40

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Book Description
The fuel material specified for the Mark III core of EBR-I was uranium-2 wt. % zirconium alloy coextruded with Zircaloy-2 cladding. From previous work on swaged or rolled uranium-2 wt% zirconium alloy, it was anticipated that the extruded alloy would be dimensionally unstable under irradiation unless stabilized by suitable heat treatment. In order to determine an effective heat treatment, irradiation studies were made on both clad and unclad extruded uranium-2 wt.% zirconium alloy specimens at irradiation temperature estimated at 200 to 750 deg C. The irradiation specimens included material with three different heat treatments, selected on the basis of previous studies, and material transient melted in its cladding. For unclad specimens, it was found that the irradiation temperature strongly influenced the various irradiation growth rates resulting from different heat treatments. Growth rates of the clad specimens were relatively insensitive to either irradiation temperature or prior heat treatment. An exception was the transient-melted material, which shortened under irradiation. The cladding had only limited ability to restrain the swelling rates of specimens irradiated at the more elevated temperatures. Clad transient-melted material was found to be most resistant to high-temperature swelling under irradiation. The results of the present study combined with observations in earlier investigations resulted in a recommendation that the reference heat treatment for the core consist of gamma solution at 800 deg C followed by isothermal transformation at 690 deg C.