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1.The thermal conductivity of UO2 sphere-pac beds
R.O.A. Hall*, a, D.G. Martin** Journal of Nuclear Materials Volume 173, Issue 2, 2 September 1990, Pages 130-141

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dx.doi.org/10.1016/0022-3115(90)90251-H 

2.Analytical and experimental performance of sphere-pacnext term nuclear fuels
K.L. Peddicord*, R.W. Stratton† and J.K. Thomas Progress in Nuclear Energy Volume 18, Issue 3, 1986, Pages 265-299

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dx.doi.org/10.1016/0149-1970(86)90036-3  

3.Nitrides as a nuclear fuel option
Marco and Franz Ingold Journal of the European Ceramic Society Volume 25, Issue 12, 2005, Pages 2687-2692

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dx.doi.org/10.1016/j.jeurceramsoc.2005.03.181   

4.Diffusion creep in oxide ceramics
Atul H. Chokshi Journal of the European Ceramic Society Volume 22, Issues 14-15, 2002, Pages 2469-2478

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dx.doi.org/10.1016/S0955-2219(02)00105-X  

5.Creep properties of oxide and carbide fuels under irradiation
D.J. Clough Journal of Nuclear Materials Volume 65, March 1977, Pages 24-36

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dx.doi.org/10.1016/0022-3115(77)90038-1    

6.Irradiation induced creep of ceramic nuclear fuels
W. Diensta Journal of Nuclear Materials Volume 65, March 1977, Pages 1-8

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dx.doi.org/10.1016/0022-3115(77)90035-6 

7.A New Code for Predicting the Thermo-Mechanical and Irradiation Behavior of Metallic Fuels in Sodium Fast Reactors
Aydın Karahan Journal of Nuclear Materials

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dx.doi.org/10.1016/j.jnucmat.2009.11.022    

8.Modeling of thermo-mechanical and irradiationnext term behavior of mixed oxide fuel for sodium fast reactors
Aydın Karahan Journal of Nuclear Materials

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dx.doi.org/10.1016/j.jnucmat.2009.11.021   

9.Metallic fuels for advanced reactors
W.J.Carmack Journal of Nuclear Materials Volume 392, Issue 2, 15 July 2009, Pages 139-150

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dx.doi.org/10.1016/j.jnucmat.2009.03.007    

10.A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design
Journal of Nuclear Materials Volume 389, Issue 3, 1 June 2009, Pages 458-469

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dx.doi.org/10.1016/j.jnucmat.2009.02.035 

11.Analysis of intergranular fission-gas bubble-size distributions in irradiated uraniumnext term–molybdenum alloy fuel
J. Rest Journal of Nuclear Materials Volume 385, Issue 3, 15 April 2009, Pages 563-571

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dx.doi.org/10.1016/j.jnucmat.2009.01.001    

12.An interdiffusional model for prediction of the interaction layer growth in the system uranium–molybdenum/aluminum
A. Soba Journal of Nuclear Materials Volume 360, Issue 3, 1 March 2007, Pages 231-241

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dx.doi.org/10.1016/j.jnucmat.2006.09.014    

13.Modeling of multicomponent systems: Role of Mo and Si on the interaction of Al and γ-UMo
Jorge E. Garcés Computational Materials Science Volume 40, Issue 1, July 2007, Pages 6-19

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dx.doi.org/10.1016/j.commatsci.2006.10.017    

14.Theoretical description of the interdiffusion of Al in the U–Mo solid solution
Jorge E. Garcés Applied Surface Science Volume 219, Issues 1-2, 15 October 2003, Pages 47-55

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dx.doi.org/10.1016/S0169-4332(03)00631-7  

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P.S. Можно попросить еще одну статью..
Microstructure evolution during irradiation
M. Kiritani Journal of Nuclear Materials Volume 216, October 1994, Pages 220-264

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http://dx.doi.org/10.1016/0022-3115(94)90014-0

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