Thermal conductivity of uranium–plutonium oxide fuel for fast reactors-Reference-Cited by-同舟云学术

Thermal conductivity of uranium–plutonium oxide fuel for fast reactors

Published:2000-12 Issue:2-3 Volume:282 Page:186-195
ISSN:0022-3115
Container-title:Journal of Nuclear Materials
language:en
Short-container-title:Journal of Nuclear Materials

Author:

Inoue Masaki

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics

Reference32 articles.

1. C. Duriez, J-P. Alessandri, T. Gervais, Y. Philipponneau, G. Chaigne, A. Chotard, International Symposium on MOX Fuel Cycle Technologies for Medium and Long-Term Deployment, Vienna, Austria, IAEA-SM-358/4P, 17–21 May 1999

2. A re-appraisal of the thermal conductivity of UO2 and mixed (U, Pu) oxide fuels

3. Thermal conductivity of (U, Pu)O2−x mixed oxide fuel

4. The electronic transport properties of UO2

5. J.M. Bonnerot, Rapport CEA-R-5450, 1988

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