Heishichiro Takahashi

Grain boundary segregation under electron irradiation in austenitic stainless steels modified with oversized elements

Abstract Effects of modification with oversized elements (Ti, Zr, Hf, V, Nb or Ta) on radiation-induced grain boundary segregation (RIS) in austenitic stainless steels were studied by means of electron irradiation in an HVEM and analytical electron microscopy. Solution-annealed 316L steels modified with the oversized elements were electron irradiated up to a dose of 10 dpa at temperatures of 673 to 773 K. Segregation behaviors of Cr and Ni near the grain boundary were strongly affected by the additional elements. Addition of Hf or Zr was particularly effective in preventing RIS. It was deduced from damaged microstructures that when the amount of irradiation-produced point defects decreased as a result of recombination and/or development of secondary defects in the matrix, segregation at the grain boundary was retarded. The change in net concentration of point defects in the matrix was associated with the trapping effect of point defects by the added elements, which depended on their size factor.

Sink effect of grain boundary on radiation-induced segregation in austenitic stainless steel

We have investigated the orientation dependence of grain boundaries in radiation-induced segregation (RIS) in an austenitic stainless steel under irradiation using a high-voltage electron microscope (HVEM) at the Hokkaido University HVEM facility and a new rate equation model for RIS, which incorporates the grain boundary sink strength for point defects. The tilt grain boundaries were chosen for the present study. It was observed that, after electron irradiation at a dose rate of 2.0 x 10 3 dpa s 1 to 10 dpa at temperatures of either 623 or 723 K, RIS was enhanced as the tilt angle increased but was suppressed at coincidence grain boundaries (Σ9 and Σ3). The present result indicates that one needs to explicitly consider the grain boundary sink strength as an important factor affecting radiation-induced grain boundary phenomena (RIGBPH). This work can be thought as a study of the radiation-induced grain boundary phenomena associated with grain boundary engineering related to non-equilibrium phenomena.

Radiation-induced segregation at internal sinks in electron irradiated binary alloys

Abstract Radiation-induced segregation phenomena in copper, nickel and iron based binary alloys were studied by means of a high voltage electron microscope and an energy dispersive X-ray microanalyzer. The segregation was caused during electron irradiation of undersaturated Cu-2 at% Ni alloy. In supersaturated Cu-2 at% Ag and Cu-2 at% Fe alloys precipitates were formed near surfaces and grain boundaries in the early stage of irradiation, subsequently voids nucleated at the central region of foil. On the other hand, alloying elements were depleted at defect sinks in Ni-2 at% Cu, Fe-5,13 at% Cr and Fe1 at% Mn alloys. The changes of solute concentration near sinks in these alloys indicate that the size effects lead to the segregation or depletion and under- and oversize solutes tend to migrate toward and away from the sinks, respectively.

Void swelling and segregation of solute in ion-irradiated ferritic steels

Void formation and radiation induced segregation were investigated through the interaction between defects and solute atoms in pure iron, Fe-13wt%Cr and Si or Ti doped alloys by 200 keV C+ ion irradiation up to 118 dpa at 798 K. Microstructural observation was carried out by transmission electron microscopy, energy dispersive X-ray microanalysis and electron energy loss spectroscopy. The ferritic alloys exhibited significant resistance to void swelling. In Fe-Cr and Fe-Cr-Si alloys, the irradiation produced the precipitates consisting chiefly of chromium and implanted carbon, and chromium was enriched at grain boundaries and voids. In the Fe-Cr-Ti alloy, Ti-rich precipitates-were formed, and chromium was depleted from grain boundaries. These facts suggest that the solute atom-defect interaction which controls void formation and segregation is affected by the presence of chromium and other alloying elements.

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.

Transmission electron microscopy of La0.7Ca0.3MnO3 thin films

La 0.7 Ca 0.3 MnO 3 (LCMO) films grown on (001) surfaces of various (pseudo-) cubic substrates, MgO, SrTiO 3 , LaSrGaO 4 and YAlO 3 , were studied by means of transmission electron microscopy. In all cases, the (001) plane of LCMO perovskite primitive cell is parallel to the substrate surface. The film on SrTiO 3 showed the single-variant structure with many twin boundaries while the films on other substrates had the three-variant structure where three [0 1 0] axes of the orthorhombic unit cell are perpendicular to each other. The LCMO cell on SrTiO 3 was expanded by about 1% in the film plane and shrank by about 1% out of the plane. Many planar defects and intergrown grains were observed in the film on MgO. They originated at the substrate surface.

Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

Heterogeneous dislocation formation and solute redistribution near grain boundaries in austenitic stainless steel under electron irradiation

We investigated the radiation-induced segregation (RIS) near grain boundaries in austenitic stainless steel under electron irradiation, taking into account the simultaneous evolution of faulted dislocation loops and network dislocations. The formation of a dislocation-free zone (DLFZ) in the vicinity of the grain boundary due to in situ irradiation was observed using a high-voltage electron microscope (HVEM). Predicted temperature and dose dependencies of DLFZ formation agreed with the experimental results. The relationship between RIS and DLFZ formation near a grain boundary has been substantially clarified.

The effect of solute content on grain boundary segregation in electron-irradiated Fe-Cr-Mn alloys

Solute distribution and microstructural development in the vicinity of grain boundaries in ferritic Fe-10Cr-xMn-3Al (x = 5, 10 or 15) alloys were studied during electron irradiation to 10 dpa at 723 K. In addition, X-ray diffraction analysis was performed for determination of the volume size factor of solutes. The oversized solute atoms, manganese and aluminum, were depleted at grain boundaries, whereas the concentration of the oversized chromium rose sharply at the boundary. The amount of segregation of the solutes decreased with increasing atomic volume depending on manganese content. Segregation of aluminum, which had the greater volume size factor relative to that of manganese, was higher than that of manganese. The amount of radiation-induced segregation of manganese and aluminum at the grain boundary is consistent with arguments based on the atomic size effect, but the enrichment of chromium at the grain boundary is not and seems to related primarily to the formation of chromium-rich precipitates at the boundary.

Effect of Mn addition on decrease of Cr depletion at grain boundary in austenitic alloys irradiated with electrons

Abstract Radiation-induced Cr depletion at a grain boundary (GB) is known as one of the major factors to degrade corrosion resistance of austenitic stainless steel. The effect of 10% Mn addition on prevention of the Cr depletion was investigated from a viewpoint of volume size factor (VSF) of Cr in the austenitic alloys irradiated with 1 MeV electrons. VSF of Cr in solution-annealed 316L steel added with 10 wt% Mn was + 0.8%, decreased by 4% compared with 316L. Radiation-induced Cr depletion at GB of 316L + 10%Mn was smaller than that of 316L at 723 and 773 K. Decrease of radiation-induced Cr depletion in 316LF + 10%Mn is thought to be derived mainly from the suppression of vacancy—Cr atom interaction.