Analyzing the causes for the dispersion of the fast reactor spent fuel pin cladding properties
The swelling, corrosion and high-temperature embrittlement behavior of the fast-neutron sodium-cooled reactor standard and test fuel pin cladding was studied following the operation up to a damaging dose of 55 to 69 dpa. The tested characteristics were found to differ sensitively in conditions similar to irradiationfor the claddings of the experimental tube conversion technology. Unlike the standard fuel pin claddings, the test pin claddings were additionally heated in the process of fabrication to homogenize the solid solution at different temperatures and austenitization times. On the whole, this led to an increased cladding resistance due to the damaging factor of the reactor environment. The positive effect is explained by the influence of carbon and the morphology of swelling-reducing alloying elements, as well as by the nature of the carbide and intermetallide phase separation. However, the dispersion of the post-irradiation properties which remained significant and was also earlier observed in the standard pins, is explained by potential differences in the heat treatment technology and the irradiation temperature in conditions of a hard-to-control coolant flow velocity. The swelling rate and the in-fuel corrosion depth for the test technology tubes were respectively 0.04 to 0.058 %/dpa and 20 to 47 μm; similar values for the test material are 0.036 to 0.056 %/dpa and 15 to 35 μ m respectively. The short-term mechanical properties of the test fuel pins at a temperature of 600°C showed a smaller tendency towards high-temperature embrittlement. The dispersion of the properties was caused by the chemical and structural heterogeneity as the result of the tube fabrication.
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