PNNL

Abstract

Various components of pressurized water power reactors (PWRs) and some proposed fusion devices such as ITER will operate at lower temperatures and displacement rates than are encountered in many test reactors such as EBR-II, FFTF and HFIR. The question arises if the presence and magnitude of void swelling can be predicted for such irradiation environments. Data on Russian steel can be used to address part of this question. In reactor applications where Western countries typically use annealed AISI 304 stainless steel, it is the Russian practice to use annealed X18H10T, a titanium-stabilized 18Cr-10Ni stainless steel analogous to AISI 321. Using a flow restrictor component from the low-flux breeder zone of the BN-350 reactor in Kazakhstan, it was possible to examine the behavior of void swelling at relatively low temperatures and low displacement rates after 12 years of irradiation. The temperature of this component ranged from 270-340 degrees centigrade with a peak dose rate of 1.6 x 10-7 power dpa/sec and a peak dose of 56 dpa. Careful sectioning of the component has yielded a large number of microscopy specimens over a ITER-relevant range of temperatures and displacement rates. Microstructural data are presented and show that void swelling at 10 to 50 dpa persists down to ~306 degrees centigrade for dose rates on the order of 1 x 10-7 power dpa/sec.