Thermal conductivity of fresh and irradiated U-Mo fuels

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a...

Full description

Saved in:
Bibliographic Details
Published in:Journal of nuclear materials 2018-05, Vol.503 (C), p.304-313
Main Authors: Huber, Tanja K., Breitkreutz, Harald, Burkes, Douglas E., Casella, Amanda J., Casella, Andrew M., Elgeti, Stefan, Reiter, Christian, Robinson, Adam. B., Smith, Frances. N., Wachs, Daniel. M., Petry, Winfried
Format: Article
Language:English
Subjects:
Composite materials
Crystal lattices
Fuel consumption
Heat
Heat conductivity
Heat transfer
irradiated
Irradiation
nuclear fuel
Nuclear fuel burnup
Nuclear reactor materials
Quantum physics
Quantum theory
Radiation
Temperature effects
Thermal analysis
Thermal conductivity
thermal properties
Thermo-physical properties
Thermodynamics
Uranium-Molybdenum
Citations: Items that this one cites
Items that cite this one
Online Access:Get full text
Tags: Add Tag
No Tags, Be the first to tag this record!
Description
Summary:The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.
ISSN:0022-3115
1873-4820
DOI:10.1016/j.jnucmat.2018.01.056