Thermal conductivity of zirconia based inert matrix fuel: Use and abuse of the formal models for testing new experimental data

C. Degueldre, T. Arima, Y. W. Lee

Research output: Contribution to journalConference articlepeer-review

40 Citations (Scopus)

Abstract

An inert matrix fuel material based on yttria-stabilized cubic zirconia: ErxYyPuzZr1-x-y-zO2-(x + y)/2 (x + y = 0.15, z: [0.05-0.15]) was proposed for burning excess plutonium in light water reactors. The studied inert matrix fuel is made of cubic stabilized zirconia. The limited number of experimental thermal conductivity data justifies this formal and intensive study. Approaches derived from Klemens theory were revisited and the derived conductivity model applied for zirconia, accounting the effects of phononic scattering centers. The hyperbolic thermal conductivity trend with temperature known for pure zirconia, is reduced by isotopes, impurities, dopants and oxygen vacancies, which act as scattering centers and contribute to conductivity reduction to a flat plot with temperature for stabilized zirconia. It is experimentally observed that the thermal conductivity derived from laser flash measurements for ErxYyMzZr1-x-y-zO2-(x + y)/2 (with M = Ce or Pu, z = 0 or ∼ 0.1 and x + y = 0.15) is rather constant as a function of temperature in the range 300-1000 K. The thermal conductivity was observed to depend on the concentration of dopants such as YO1.5 and/or ErO1.5, CeO2 (analogous of PuO2) or PuO2. The bulk material conductivity of Er0.05Y0.10Pu0.10Zr0.75O1.925 is about 2 Wm-1K-1. In this study, the thermal conductivity data of both monoclinic and stabilized cubic zirconia based IMF are tested with the model approach in order to understand the experimental data in a semi-quantitative way.

Original languageEnglish
Pages (from-to)6-14
Number of pages9
JournalJournal of Nuclear Materials
Volume319
DOIs
Publication statusPublished - Jun 1 2003
EventProceedings of the 8th Inert Matrix Fuel Workshop - Tokai, Japan
Duration: Oct 16 2002Oct 18 2002

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Materials Science(all)
  • Nuclear Energy and Engineering

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