Author: J. W. Weber
Publisher:
ISBN:
Category :
Languages : en
Pages : 58
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
The Development and Testing of the UO2 Fuel Element System
Author:
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 112
Book Description
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 112
Book Description
Oxide Particle Size Distribution from Shearing Irradiated and Unirradiated LWR Fuels in Zircaloy and Stainless Steel Cladding
Author: Wallace Davis
Publisher:
ISBN:
Category : Government publications
Languages : en
Pages : 52
Book Description
Publisher:
ISBN:
Category : Government publications
Languages : en
Pages : 52
Book Description
Interdiffusion in Zircaloy-2 Clad U-2 W/o Zr Fuel Materials and Its Effect Upon Corrosion Behavior
Author: A. L. Geary
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 28
Book Description
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 28
Book Description
Fabrication of Teflon Critical Experiment Fuel Elements
Author: J. L. Swanson
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 20
Book Description
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 20
Book Description
Hot Extrusion of UO2 Fuel Elements
Author: J. G. Hunt
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 52
Book Description
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 52
Book Description
Internal Zirconium Hydride Formation in Zircaloy Fuel Element Cladding Under Irradiation
Author: J. M. Markowitz
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 52
Book Description
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 52
Book Description
Reactor Irradiation of Uranium-impregnated Graphite at 1500°C to 10% Burnup
Author: P. J. Peterson
Publisher:
ISBN:
Category : Graphite as fuel
Languages : en
Pages : 50
Book Description
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.
Publisher:
ISBN:
Category : Graphite as fuel
Languages : en
Pages : 50
Book Description
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.
The Zircaloy-2 In-pile Tube for the NRX Central Thimble
Author: Richard M. Lieberman
Publisher:
ISBN:
Category : Nuclear fuel rods
Languages : en
Pages : 126
Book Description
Publisher:
ISBN:
Category : Nuclear fuel rods
Languages : en
Pages : 126
Book Description
Specifications for Vibrationally Compacted UO2 Nested Tubular Fuel Element (PRTR MARK II-C)
Author: M. K. Millhollen
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 72
Book Description
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 72
Book Description