Papers by Keyword: Irradiation Creep

Abstract: Article discusses experimental data on creep of (U,Pu)N and other uranium compounds, and possible mechanism of mass-transfer. Proposed equation describes the following creep features: weak temperature dependence at T < 1000°C, creep acceleration in a fuel with micron-sized grains, and acceleration with the content of second phases formed by impurities and fission products. The difference in creep behavior in reactors with thermal and fast neutrons environmentsis discussed. Comparison of irradiation creep of nitride fuel and properties of cladding materials shows that under parameters of fast reactors and typical design of fuel element it is impossible to implement restraining of external nitride swelling. As initial porosity in the fuel will not compensate the nitride swelling, the cladding of fuel element will work in a mode of following the changing of fuel size. Some suggestions on the cladding material properties are done.

Abstract: Miniature and sub-miniature samples were used for determination of mechanical properties of materials for advanced fission plants. Results from indentation and focused ion beam prepared micro-samples, punch tests and thin strip (irradiation) creep tests are shown. The results allow conclusions concerning materials damage. Irradiation damage profiles were determined with indentation. Results from micro-pillar tests showed a good agreement with results from conventional samples in case of oxide dispersion strengthened steels. Thin strip irradiation creep experiments revealed a negligible influence of dispersoid size/distribution on creep rates. Punch tests of fibre reinforced materials showed consistent results which still need quantitative analysis.