This study investigates the tensile behaviors of additively manufactured (AM) 17-4PH stainless steels heat-treated within various temperature ranges from 400 °C to 700 °C in order to identify the effective aging temperature. Despite an aging treatment of 400–460 °C increasing the retained austenite content, an enhancement of the tensile properties was achieved without a strength-ductility trade-off owing to precipitation hardening by the Cu particles. Due to the intricate evolution of the microstructure, aging treatments above 490 °C led to a loss in yield strength and ductility. A considerable rise in strength and a decrease in ductility were brought about by the increase in the fraction of precipitation-hardened martensitic matrix in aging treatments over 640 °C. The impact of heat-treatment pathways on aging effectiveness and tensile anisotropy was then examined. Direct aging at 482 °C for an hour had hardly any effect on wrought 17-4PH, but it increased the yield strength of AM counterparts from 436–457 to 588–604 MPa. A solid-solution treatment at 1038 °C for one hour resulted in a significant drop in the austenite fraction, which led to an increase in the yield (from 436–457 to 841–919 MPa) and tensile strengths (from 1106–1127 to 1254–1256 MPa) with a sacrifice in ductility. Improved strength and ductility were realized by a solid-solution followed by an aging treatment, achieving 1371–1399 MPa. The tensile behaviors of AM 17-4PH were isotropic both parallel and perpendicular to the building direction.

Korea has put in a huge effort to develop the future nuclear systems such as a sodium-cooled fast reactor (SFR), a very high temperature reactor (VHTR) and fusion reactor. As the realization of such nuclear systems relies heavily on performance of materials, Korea has made steady and persistent effort to develop structural materials that can withstand the extremely harsh operating conditions. The present paper summarizes the current status of the Korean R&D activities on advanced nuclear materials including oxide dispersion strengthened (ODS) steels, Ni- base alloys and their ODS variations, reduced-activation ferritic-martensitic (RAFM) steels, and SiC composites

The effects of proton irradiation on a beryllium reflector in terms of microstructure evolution have been studied to emulate the effects of neutron irradiation. Protons were irradiated on a beryllium...

In this study, the effect of simple shearing on microstructure evolution and mechanical properties of 316L austenitic stainless steel were investigated. Two different shear strain routes were obtained by twisting cylindrical specimens in the forward and backward directions. The strain-induced martensite phase was effectively obtained by alteration of the routes. Formation of the martensite phase clearly resulted in significant hardening of the steel. Grain-size reduction and strain-induced martensitic transformation within the deformed structures of the strained specimens were characterized by scanning electron microscopy – electron back-scattered diffraction, X-ray diffraction, and the TEM-ASTAR (transmission electron microscopy – analytical scanning transmission atomic resolution, automatic crystal orientation/phase mapping for TEM) system. Significant numbers of twin networks were formed by alteration of the shear strain routes, and the martensite phases were nucleated at the twin interfaces.

In this study, Fe³⁺ self-ion irradiation is used as means of introducing irradiation damage in ferritic/martensitic (F/M) steel and oxide dispersion strengthened (ODS) steel. The ion accelerator named DuET (at Kyoto University, Japan) was used for irradiation with 6.4 MeV Fe³⁺ ions at 300 °C. The total number of accelerated ions was 2.5 × 10²⁰ ions/m², and the maximum damage rates in the F/M and the ODS steels were estimated to be roughly 6 dpa. The irradiation-induced hardness change in the damaged layer was evaluated by using nano-indentation. The F/M steel and the ODS steel commonly exhibited irradiation hardening; however, the irradiation hardening was more active in the F/M steel than in the ODS steel. The microstructure evolutions after the irradiation were investigated; point or line defects were dominantly observed in the F/M steel, while small circular cavities were typically observed in ODS steel.