Neutronography of Irradiated Reactor Austenitic Steels

Abstract

The neutron diffracttion studies of the samples of fuel element claddings made of EK-164 and ChS-600 austenitic steels in the initial state and after operation in a BN-600 fast neutron reactor are reviewed. The samples were investigated in a wide range of fast neutron doses and irradiation temperatures, up to ∼80 displacements per atoms (dpa) and 628°C, respectively. Various microstructural characteristics, such as microstresses, dislocation density, and crystallographic texture are determined using full-profile analysis of neutron diffraction data. At high doses studied in this work, the irradiation temperature is the dominant factor responsible for the dislocation density. The procedures of analyzing and revealing defect types were preliminary developed on specially prepared nickel samples and austenitic alloys upon irradiation in the IVV-2M reactor core at a temperature near 80°С, which provided a low thermal mobility of lattice defects. The results obtained in some first diffraction studies on samples after their operation in the reactor core show that neutron data can be used to characterize the microstructures formed due to irradiation and, at least, estimate the mechanical properties of irradiated materials without their destruction and with no radiation risk for personnel.