Significant effort has been underway to provide a code case for use of alloy A709 as a structural material in advanced nuclear reactors ranging from Sodium Cooled Fast Reactors (SFR’s) to High-Temperature Gas-cooled Reactors (HTGR’s). The hostile environments of these varied advanced reactor concepts will vary with respect to coolant/cladding compatibility criteria, stress/temperature combinations that may influence necessary experimentation regarding irradiation creep, and irradiation-induced and environmentally-assisted cracking; however, the effect of irradiation on the microstructure and mechanical properties of A709 as a function of temperature is a fundamental requirement for all systems. Unfortunately, the United States’ test reactors are limited to either the Advanced Test Reactor (ATR) or the High Flux Isotope Reactor (HFIR) for any irradiation campaign aimed at qualifying A709. Both have high thermal neutron fluxes and the associated high levels of He-production from the Ni content of A709. Thus, the objective of this work is to quantify differences in cavity distributions following irradiation of advanced austenitic stainless steel A709 as a function of varied Helium (He) production and damage rates. We hypothesize that, due to the significantly higher He-production rate in thermal test reactors, cavity distributions will be dominated by nanoscale He-bubbles that will fundamentally change post-irradiation measurements of strength and ductility following low-dose neutron irradiation compared to anticipated fast reactor spectra. Ion irradiations may provide an alternative solution to examine the differences in expected cavity nucleation in A709 between thermal and harder neutron spectra in an accelerated manner. We propose to perform a systematic ion irradiation study spanning two relevant operating temperature conditions and two He-generation rates for comparison with neutron-irradiated A709 from HFIR. The proposing team seeks use, through the Nuclear Science User Facilities, one partner institution for dual ion irradiation (UM), one partner institution for sample preparation (ORNL), and two partner institutions for post-irradiation examination (UM, CAES). The A709 material is immediately available for ion-irradiations and PIE, while the neutron-irradiated material will be extracted from HFIR in May 2024 and ready for sample preparation in LAMDA by August 2024, providing the high-impact comparative datapoints with low risk during the 12 month period of performance. Microstructure examination will include dislocation loops, precipitates, and cavities conducted in parallel at the university and national laboratory NSUF partner facilities. The outcome of this work will provide a quantitative analysis of the irradiation microstructure, with an emphasis on the cavity/He bubble distribution at low displacement damage levels (~2 dpa) anticipated for A709’s structural use case. The availability of this dataset will inform subsequent decisions on qualification-relevant neutron irradiations for A709 and enable a more realistic interpretation of the advanced austenitic alloy in fast-reactor operating environments.

The Office of Nuclear Energy (DOE-NE) mission is to advance nuclear energy science and technology to meet U.S. energy, environmental, and economic needs. This Super RTE focuses on the generation of radiation response data for a candidate structural alloy in advanced nuclear reactor environments using ion irradiation in lieu of a fast neutron spectrum test facility. Irradiations of Alloy 709 in the nation’s primarily thermal spectrum test reactors will result in helium production not representative of the target application. Microstructure evolution under irradiation would directly benefit the DOE-Office of Nuclear Energy (NE) in several areas. The focus on Alloy 709 for advanced nuclear environments in this Super RTE addresses the goal of the Advanced Reactor Technologies (ART) program to conduct R&D on advanced reactor concepts. The ART program has long supported the development of Alloy 709 through its code qualification campaign for Class A component design in ASME Section III, Division 5. Although radiation performance data are not needed for ASME code qualification of this alloy, this Super RTE fills a critical need as this type of systematic evaluation of radiation-induced evolution in material microstructure and properties will be necessary to support eventual application in nuclear power plant designs. The objectives of this proposal align strongly with the DOE NE Advanced Materials & Manufacturing Technologies (AMMT) program mission to accelerate the development, qualification, and deployment of materials and manufacturing technologies to enable reliable and economical nuclear energy. At the end of FY23, the AMMT program developed an accelerated materials qualification framework combining ion irradiations, neutron irradiations, and modeling. The use of ion irradiation in this Super RTE to generate specimens for comparison with neutron-irradiated material in support of model development fits many aspects of this framework and may serve as part of a demonstration case despite the fact that these are not fast spectrum tests.