Open AccessPRELIMINARY IRRADIATIONS OF PuC AND UC-PuC
Author(s) -
Jean Kittel,
L.A. Neimark,
R. Carlander,
O.L. Kruger,
R.C. Lied
Publication year - 1963
Publication title -
osti oai (u.s. department of energy office of scientific and technical information)
Language(s) - English
Resource type - Reports
DOI - 10.2172/4127983
Subject(s) - pellets , materials science , irradiation , rod , carbide , burnup , metallurgy , radiochemistry , metal , fission , graphite , composite material , nuclear chemistry , chemistry , neutron , nuclear physics , medicine , physics , alternative medicine , pathology
Preliminary irradiations were made of arc-cast specimens of RaC andd UC- PuC andd of pressed and sintered pellets of UC--RaC to determine the suitability of these materials for further development as possible fuels for liquid metal- cooled fast reactors. The PuC content in the mixedcarbide specimens was near 20 wt%. The specimens were irradiated as fuel rods in the fast reactor EBR-I. Maximum fuel burnup and temperature during irradiation were respectively 0. 10 metal atom percent and 650 deg C. The cast PuC specimens were found to have fractured to a greater extent under irradiation than did the cast mixed-carbide specimens. Dimensional chandges in those specimens which did not fracture were negligible. Fission gas release from the cast specimens was 0.24% of the theoretical yield. The pressed andd sintered pellets did not fracture under irradiation and showed negligible dimensional chandges. Fission gas release from the pellets was 12.0% of theoretical. The high gas release from the pressed and sintered specimens is attributed to a relatively high degree of open porosity which existed in the pellets. (auth
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