Ferritic-martensitic steels are being considered as candidate structural materials for fast reactors and advanced LWR due to their excellent resistance to radiation-induced void swelling, microstructural stability, thermal conductivity, and superior irradiation creep properties. HT-9 was selected as the fuel clad and duct material in FFTF and EBR-II, and it is still the first-choice candidate core material for several advanced reactor concepts due to its service performance and the relatively large database on it. Currently, commercial nuclear power companies such as TerraPower has rejuvenated the manufacturing of HT-9.



To address the issue of low-temperature (~425°C) neutron irradiation hardening and embrittlement, it is necessary to conduct systematic investigations on the mechanical behavior and microstructure of HT-9 with slight variations in chemical composition and heat treatment over a wide range of doses and temperatures. Three HT-9 heats (ORNL, LANL and EBR II) with variations in manufacturing process, chemical composition and heat treatment were neutron irradiated (~4 dpa) in the ATR at different temperatures (241-469C). Our team recently won RTE award (#1687) and completed microhardness and tensile testing of these HT-9 variants. Maximum impact of this work will be obtained by performing TEM and APT characterization of these samples to correlate the measured hardening with microstructural features.



We propose to employ the APT technique to enhance the understanding of the underlying mechanisms for α′ and Ni/Mn/Si precipitation evolution in HT-9 upon neutron irradiation. Currently, efforts are ongoing to perform APT study (#1) at CAES under a funded RTE project#4629 on irradiated (~4 dpa) HT-9 as a function of irradiation temperature (241°C, 388°C and 452°C). The objective of this RTE proposal (follow-up study, #2) is to perform APT study on irradiated HT-9 (EBR II variant) that has a significant variation in heat treatment, alloying elements (Ni, Mn, Si and W) and mechanical properties, when compared to the HT-9 heat (ORNL) being studied now (RTE project#4629). Since both alloys were irradiated together in the ATR, the overall objective is to evaluate the formation of α′ and Ni/Mn/Si precipitation in neutron irradiated (~4 dpa) HT-9 as a function of irradiation temperature (241°C, 291°C, 388°C and 452°C), heat treatment and alloying elements (Ni, Mn, Si, W).



To achieve this objective, we will utilize APT to determine the size, number density, and chemical composition of the α′ and Ni/Mn/Si precipitates. To understand the contributions to irradiation hardening, the dispersed barrier-hardening model will be employed to determine the changes in yield strength induced by each type of obstacles, and then compare the measured hardening (tensile and microhardness data from funded RTE) with microstructure-deduced hardening contributions obtained from APT (proposed RTE; α’ and G phases) and TEM studies (ongoing work). By doing this work, our team can contribute to filling the gap in the literature on understanding the irradiation effects on HT-9 and F-M steels in general. The project performance (sample preparation, APT characterization, and analysis) is expected to take place during June - Dec 2024 and will result in one conference presentation and one journal article publication.

The nuclear energy plays an important role in the nation’s diverse electricity portfolio. To satisfy the fast-growing electricity demand, climate change, safety and waste concerns, efforts are ongoing to design advanced nuclear reactors with greater thermal efficiency, flexibility, safety, and economics than the current generation. These advanced reactors require high performance materials that can operate under more aggressive service conditions (such as higher operating temperatures, higher radiation doses and corrosive environment).



One of the goals of the USDOE-NE research programs is to develop advanced structural materials for the next generation of reactors that have good neutronics, dimensional stability, corrosion resistance, mechanical (yield strength, ductility, fracture toughness, creep strength, etc.,) and thermal properties under irradiation. HT-9 is being considered as a candidate structural material for fast and advanced LWR, due to their superior irradiation resistance. Despite various advantages, HT-9 and other F-M steels have serious issues during low-temperature (~425C) neutron irradiation, and it is a critical concern because temperature drops during reactor shutdowns and temperature transients are inevitable during operation.



To address the issue of low-temperature neutron irradiation hardening and embrittlement, systematic investigations on the mechanical behavior and microstructure of HT-9 with slight variations in chemical composition and heat treatment are needed over a wide range of doses and temperatures. Three HT-9 heats (ORNL, LANL and EBR II) with variations in manufacturing process, chemical composition and heat treatment were neutron irradiated (~4 dpa) in the ATR at different temperatures (241°C-469°C) under an NSUF irradiation experiment. Our team recently won RTE award and completed microhardness and tensile testing of HT-9 variants (~4 dpa).



The objective of this RTE proposal is to evaluate the formation of α′ and Ni/Mn/Si precipitation in neutron irradiated (~4 dpa) HT-9 as a function of irradiation temperature (241°C, 291°C, 388°C and 469°C), alloying content (Ni, Mn, Si, W) and heat treatment using APT. The microstructural information obtained from this proposed work would be analyzed to understand the effects of radiation damage on HT-9 heats at LWR and fast reactor relevant temperatures, and to successfully develop appropriate processing-composition-structure-property-temperature correlations. Limited low-temperature neutron irradiation data exists addressing the above variations. Hence, our team can begin to fill the gap in the literature by successfully completing the proposed work on neutron irradiated HT-9. The results of the proposed work could be extended beyond HT-9, and it would be relevant to many F-M steels in general.



In addition, the proposed work will benefit the AMMT program in developing advanced structural materials (for fast and advanced LWR) with optimized chemical composition and heat treatment for greater radiation resistance, and NEAMS program by providing experimental results that would enable models to extrapolate it to the wider range of in-service conditions of future advanced reactors. Thus, the proposed work is well-aligned with the Office of Nuclear Energy’s missions and vision and will have substantial implications for the deployment of next-generation advanced reactors.