The Development and Testing of the UO2 Fuel Element System

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

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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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Exponential Experiments with Graphite Lattices Containing Multirod Slightly Enriched Uranium Fuel Clusters

Exponential Experiments with Graphite Lattices Containing Multirod Slightly Enriched Uranium Fuel Clusters PDF Author: W. W. Brown
Publisher:
ISBN:
Category : Graphite
Languages : en
Pages : 50

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Critical Experiments on Slightly Enriched Uranium Metal Fuel Elements in Graphite Lattices

Critical Experiments on Slightly Enriched Uranium Metal Fuel Elements in Graphite Lattices PDF Author:
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ISBN:
Category :
Languages : en
Pages : 44

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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.

Summary Report

Summary Report PDF Author:
Publisher:
ISBN:
Category : Graphite
Languages : en
Pages : 142

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Preliminary Irradiation of Fused UO2

Preliminary Irradiation of Fused UO2 PDF Author: G. Rolland Cole
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Category : Irradiation
Languages : en
Pages : 36

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Symposium on Effects of Irradiation on Fuel and Fuel Elements

Symposium on Effects of Irradiation on Fuel and Fuel Elements PDF Author: W. D. Manly
Publisher:
ISBN:
Category : Metallurgy
Languages : en
Pages : 114

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Effects of Irradiation on Bulk UO2

Effects of Irradiation on Bulk UO2 PDF Author: John D. Eichenberg
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 198

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Metallography of Irradiated UO2-containing Fuel Elements

Metallography of Irradiated UO2-containing Fuel Elements PDF Author: W. K. Barney
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Category : Nuclear energy
Languages : en
Pages : 64

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Radiation Behavior of Metallic Fuels for Sodium Graphite Reactors

Radiation Behavior of Metallic Fuels for Sodium Graphite Reactors PDF Author: B. R. Hayward
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 40

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