Open AccessIn-reactor rupture testing of Zircaloy-2 clad seven-rod cluster fuel elements
Author(s) -
R.L. Call,
D.C. Kaulitz
Publication year - 1960
Publication title -
osti oai (u.s. department of energy office of scientific and technical information)
Language(s) - English
Resource type - Reports
DOI - 10.2172/10151632
Subject(s) - cladding (metalworking) , materials science , shearing (physics) , fuel element failure , nuclear fission product , nuclear engineering , control rod , pressurized water reactor , water cooled , nuclear fuel , composite material , fission products , structural engineering , metallurgy , nuclear reactor core , engineering , mechanical engineering , water cooling
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