N Akasaka

Phase stability of oxide dispersion-strengthened ferritic steels in neutron irradiation

Abstract Oxide dispersion-strengthened ferritic steels were irradiated by neutrons up to 21 dpa and studied by microstructural observation and microchemical analysis. The original high dislocation density did not change after neutron irradiation, indicating that the dispersed oxide particles have high stability under neutron irradiation. However, there is potential for recoil resolution of the oxide particles due to ballistic ejection at high dose. From the microchemical analysis, it was implied that some of the complex oxides have a double-layer structure, such that TiO2 occupied the core region and Y2O3 the outer layer. Such a structure may be more stable than the simple mono-oxides. Under high-temperature irradiation, Laves phase was the predominant precipitate occurring at grain boundaries α phase and χ phase were not observed in this study.

Effect of mechanical alloying parameters on irradiation damage in oxide dispersion strengthened ferritic steels

Abstract Issues for developing oxide dispersion strengthened (ODS) steel are anisotropic mechanical properties due to the bamboo-like structure, impurity pick up during the mechanical alloying (MA) process, stability of oxide particles, heat-treatment condition and chemical composition. Several ODS steels were fabricated with a changing gas environment during MA, heat-treatment condition and chemical composition, and were electron-irradiated to 12 dpa at 673–748 K in a high-voltage electron microscope. An ODS martensitic steel (M–Ar) with high dislocation density showed very good swelling resistance. Swelling levels of ODS ferritic steels depended on the gas environment during MA and the recrystallization condition. These indicated that a helium gas environment during MA was more effective to suppress swelling than an argon gas environment and that cold working after recrystallization reduced void formation and swelling. The effect of MA parameters, such as the gas environment, heat-treat condition and cold working on the swelling behavior was evaluated.

Effect of temperature gradients on void formation in modified 316 stainless steel cladding

Focusing on the effect of temperature gradients across the cladding wall on void formation, observation of void distribution was carried out by means of electron microscopic techniques in the direction of the wall thickness in the irradiated fuel claddings made of P, Ti-modified 316 steel. A deformation analysis was also conducted including irradiation creep and swelling deformation, using a finite element method. At the beginning of swelling a secondary stress occurred at the cladding surface due to non-uniform swelling arising from the temperature gradient. Swelling appears to accelerate by this secondary stress, and the swelling developed a non-monotonic distribution, namely the swelling in mid-wall region was lower than that in the surface regions. This non-monotonic swelling distribution disappears with increasing fluence, however. In the deformation analysis, it became clear that swelling is not continuously accelerated by the secondary stress, because irradiation creep deformation relaxes the stress.

Behavior of Irradiated Type 316 Stainless Steels under Low-Strain-Rate Tensile Conditions

The effects of lower strain rate on the tensile behavior of 12 % cold-worked type 316 stainless steels irradiated in the EBR-II reactor under low-dose-rate and moderate temperature conditions were investigated. Tensile tests were carried out at a strain rate of 1 × 10−7/s. Post-test fractography and microstructural characterization were also performed. Irradiation temperature and dose appeared to have the greatest effect on hardening and ductility loss, whereas dose rate appeared to have less apparent effects. In conjunction with earlier work performed on the same material at a strain rate of 1 × 10−3/s, there was no significant effect of strain rate on tensile behavior under the irradiation conditions examined. For fracture behavior, the material after irradiation exhibited typical ductile fracture during both high and low-strain-rate tests.

Swelling and Microstructural Evolution in 316 Stainless Steel Hexagonal Ducts Following Long-Term Irradiation in EBR-II

Swelling behavior and microstructural evolution of 12% cold-worked 316 SS hexagonal ducts following irradiation in the outer rows of EBR-II is described. Immersion density measurements and transmission electron microscopy (TEM) examination were performed on a total of seven irradiation conditions. The samples were irradiated to temperatures between 375 and 430 C to doses between 23 and 51 dpa and at dose-rates ranging from 1.3 x 10{sup -7} to 5.8 x 10{sup -7} dpa/s. Dose-rates and temperatures approach conditions experienced by a variety of components in pressurized water reactors (PWRs) and those which may be present in future advanced reactors designs. TEM analysis was employed to elucidate the effect of radiation on the dislocation, void and precipitate structures as a function of irradiation conditions. A moderate dose-rate effect was observed for samples which were irradiated at dose-rates differing by a factor of two. Lower dose-rate samples contained voids of larger diameter and typically swelled more in the bulk. The dislocation and precipitate structure was not visibly influenced by a dose-rate decrease.