Open AccessDefect testing of coextruded uranium-zircaloy-II clad fuel material in a 300 C out-of-reactor recirculating water loop: Interim report
Author(s) -
K.D. Hayden,
J.W. Goffard
Publication year - 1959
Language(s) - English
Resource type - Reports
DOI - 10.2172/10169406
Subject(s) - materials science , uranium , water cooling , corrosion , metallurgy , coolant , zirconium alloy , pressurized water reactor , nuclear engineering , waste management , zirconium , nuclear physics , mechanical engineering , engineering , physics