Final Report

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

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Final Report

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

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Book Description


The Development and Testing of the UO2 Fuel Element System

The Development and Testing of the UO2 Fuel Element System PDF Author:
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 112

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Oxide Particle Size Distribution from Shearing Irradiated and Unirradiated LWR Fuels in Zircaloy and Stainless Steel Cladding

Oxide Particle Size Distribution from Shearing Irradiated and Unirradiated LWR Fuels in Zircaloy and Stainless Steel Cladding PDF Author: Wallace Davis
Publisher:
ISBN:
Category : Government publications
Languages : en
Pages : 52

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Interdiffusion in Zircaloy-2 Clad U-2 W/o Zr Fuel Materials and Its Effect Upon Corrosion Behavior

Interdiffusion in Zircaloy-2 Clad U-2 W/o Zr Fuel Materials and Its Effect Upon Corrosion Behavior PDF Author: A. L. Geary
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 28

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Fabrication of Teflon Critical Experiment Fuel Elements

Fabrication of Teflon Critical Experiment Fuel Elements PDF Author: J. L. Swanson
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 20

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Hot Extrusion of UO2 Fuel Elements

Hot Extrusion of UO2 Fuel Elements PDF Author: J. G. Hunt
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 52

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Internal Zirconium Hydride Formation in Zircaloy Fuel Element Cladding Under Irradiation

Internal Zirconium Hydride Formation in Zircaloy Fuel Element Cladding Under Irradiation PDF Author: J. M. Markowitz
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 52

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Reactor Irradiation of Uranium-impregnated Graphite at 1500°C to 10% Burnup

Reactor Irradiation of Uranium-impregnated Graphite at 1500°C to 10% Burnup PDF Author: P. J. Peterson
Publisher:
ISBN:
Category : Graphite as fuel
Languages : en
Pages : 50

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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

The Zircaloy-2 In-pile Tube for the NRX Central Thimble PDF Author: Richard M. Lieberman
Publisher:
ISBN:
Category : Nuclear fuel rods
Languages : en
Pages : 126

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Specifications for Vibrationally Compacted UO2 Nested Tubular Fuel Element (PRTR MARK II-C)

Specifications for Vibrationally Compacted UO2 Nested Tubular Fuel Element (PRTR MARK II-C) PDF Author: M. K. Millhollen
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 72

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