Operational conditions of Gen. IV concept nuclear reactors will place significant demands on their structural materials as they operate in temperature regimes where irradiation induced swelling is a concern for several proposed reactor types. Nuclear materials development can be accelerated by innovative materials processing such as additive manufacturing (AM). The microstructure imparted through the rapid solidification of the AM process generates a subgrain-like dislocation network structure. The dislocation cell walls are expected to impart strain fields within the cell, and through this, a consistent dislocation bias for radiation defects and a driving force for radiation induced swelling. Concurrently, neutron bombardment results in helium generation which will diffuse throughout the microstructure and trap at different features including dislocations and vacancy clusters which leads to helium-driven cavity nucleation. Thus, the initial state of the dislocation microstructure after AM is likely to significantly influence the distribution of helium throughout the microstructure and with it, swelling. The objective of this work is to evaluate the effectiveness of thermal processing on swelling resistance of additively manufactured 316H steel. We hypothesize that the stress relieved condition, having a balance between dislocation bias and helium trapping at dislocations, will show increased cavity nucleation compared to an as-printed condition where internal stresses drive helium to traps, and a solution annealed condition where the initial dislocation cell microstructure from additive manufacturing is removed. The proposing team seeks use, through the Nuclear Science User Facilities, of the Michigan Ion Beam Laboratory (MIBL) for dual ion irradiation of four variants of LPBF 316H, and of the Michigan Center for Materials Characterization (MC2) for TEM lamella preparation and characterization of dislocation loops, dislocation cells, cavities, and helium distribution. The four variants include as-printed 316H with a high density of dislocation cells, heat treated to relieve internal stresses but retaining the dislocation cells, solidified through hot isostatic pressing (HIP), and solution annealed to remove the dislocation cells. We will perform dual ion irradiations at 575°C using 9 MeV Fe3+ ions and energy degraded He2+ ions to 2, 10, and 25 DPA with 2.0 appm He/dpa co-injection with a damage rate of 7 × 10-4 dpa/s. The TFS Spectra 300 Probe Corrected S/TEM at MC2 is equipped with a high angle double tilt specimen holder which serves to analyze dislocation loops with on-zone axis STEM imaging, cavities with high angle annular dark field, and high energy resolution monochromated EELS to produce maps for the qualitative identification of helium trapping sites. The proposed experiments will require an estimation of about 48 hours for lamella preparation and 96 hours for post-irradiation examination with transmission electron microscopy. The outcome of this work will provide quantitative analysis of the irradiated microstructure including dislocation loops, dislocation cell evolution, and cavities, and qualitative analysis of helium distribution. The availability of this dataset will support ongoing development activities in determining the processing of AM 316H stainless steel for advanced reactor applications.

The mission of the DOE Office of Nuclear Energy is to advance nuclear power to meet the nation's energy, environmental, and national security needs. This RTE focuses on the generation of data related to dislocation cell wall evolution and cavity nucleation as a function of radiation damage dose for several processing methods of additively or advanced manufactured 316H stainless steel. Demonstration of the processing, and resulting swelling data, would directly benefit the DOE-Office of Nuclear Energy (NE) and advanced manufacturing R&D communities, including DOE-EERE and AMO programs. The objectives of this proposal align strongly with the DOE NE Advanced Materials & Manufacturing Technologies (AMMT) program mission to develop cross cutting technologies and to accelerate the development, qualification, demonstration and deployment of materials and manufacturing technologies to enable reliable and economical nuclear energy. While primary interest lies in the AM focused programs, secondary interest may lie in the DOE NE-43 Innovative Nuclear Materials program to capitalize on recent breakthroughs in advanced characterization, manufacturing and processing capabilities to develop next generation in-core materials.