Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
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
Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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
Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
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
Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Irradiation of UO2 Fuel Rods
Author: John D. Eichenberg
Publisher:
ISBN:
Category : Nuclear fuel rods
Languages : en
Pages : 28
Book Description
Publisher:
ISBN:
Category : Nuclear fuel rods
Languages : en
Pages : 28
Book Description
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
Author: H. N. Jones
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
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
Author: P. J. Fehrenbach
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Proposal to irradiate zircaloy-clad fuel elements containing uo2 pellets of different densities
Author: F. R. Campbell
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Final Report
Author: J. W. Weber
Publisher:
ISBN:
Category :
Languages : en
Pages : 58
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 58
Book Description
Irradiation of zircaloy clad uo2 fuel elements containing discs of different materials and geometries to study their efficiency in cooling the fuel pellet
Author: R. D. Macdonald
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Irradiation of Extrusion-clad Uranium-2 W/o Zirconium Alloy for EBR-1, Mark III
Author: J. H. Kittel
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 40
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.
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 40
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.