Author:
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 112
Book Description
The Development and Testing of the UO2 Fuel Element System
Exponential Experiments with Graphite Lattices Containing Multirod Slightly Enriched Uranium Fuel Clusters
Summary Report
Critical Experiments on Slightly Enriched Uranium Metal Fuel Elements in Graphite Lattices
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.
Preliminary Irradiation of Fused UO2
Author: G. Rolland Cole
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 36
Book Description
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 36
Book Description
Symposium on Effects of Irradiation on Fuel and Fuel Elements
Author: W. D. Manly
Publisher:
ISBN:
Category : Metallurgy
Languages : en
Pages : 114
Book Description
Publisher:
ISBN:
Category : Metallurgy
Languages : en
Pages : 114
Book Description
Effects of Irradiation on Bulk UO2
Author: John D. Eichenberg
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 198
Book Description
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 198
Book Description
Metallography of Irradiated UO2-containing Fuel Elements
Author: W. K. Barney
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 64
Book Description
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 64
Book Description
Nuclear Characteristics of UO2 and ThO2 Fuel for Phase III NPR Operation
Author: G. J. Busselman
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 122
Book Description
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 122
Book Description