1. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors;Aydın Karahan;J. Nucl. Mater.,2010
2. A perspective on fast reactor fuel cycle in India;Baldev Raj;Prog. Nucl. Energy,2005
3. In-pile measurement of the thermal conductivity of irradiated metallic fuel;Bauer;Nucl. Technol.,1995
4. M.C. Billone, Y.Y.L., 1986. Status of the fuel element modeling codes for metallic fuels. International Conference Reliable Fuels for Liquid Metal Reactors, American Nuclear Society Tucson, Arizona, Sep. 7-11.
5. BOOTH, A.H., 1957. A Method of Calculating Fission Gas Diffusion from UO2 Fuel and Its Application to the X-2-f Loop Test. Atomic Energy of Canada Ltd.