In-Reactor Studies of Low-Cycle Fatigue Properties of a Nuclear Pressure Vessel Steel

摘要: Abstract : An experimental irradiation assembly and associated instrumentation which have been developed successfully utilized for the performance of dynamic in-reactor low cycle fatigue tests reactor pressure vessel steels are described. The equipment provides simultaneous reverse bend testing as many fifteen sheet type specimens representing a range strain amplitudes at controlled temperatures in 300 to 700F. results an exploratory investigation on resistance ASTM A302-B steel during 500F presented compared with data from out-of-reactor control tests. These preliminary do not indicate any pronounced difference strength irradiated versus unirradiated steel. Exploratory investigations continuing. (Author)