Ferritic-martensitic (F-M) steels are being considered as candidate in-core structural materials for advanced reactors due to their excellent resistance to radiation-induced void swelling, microstructural stability, thermal conductivity, and superior irradiation creep properties. HT-9 has a relatively large irradiation mechanical and microstructural database. HT-9 was utilized in fast reactors, and it is still the first-choice candidate core material for a number of advanced reactor concepts due to its service performance and the relatively large database on it. Currently, commercial nuclear power companies (such as TerraPower) have rejuvenated the manufacturing of HT-9. To address the issue of low-temperature (~425°C) 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 in the ATR as a part of the UW-Madison NSUF pilot irradiation experiment. PNNL received neutron irradiated tensile samples of HT-9 heats (ORNL, LANL and EBRII) as a function of processing conditions (variations in chemical composition and heat treatment) and irradiation temperatures from the NSUF library, under a funded RTE project #1687 and obtained the mechanical properties (tensile testing at irradiation temperatures and Vickers microhardness testing at room temperature). Tensile testing at irradiation temperatures showed the highest increase in yield strength at 241°C, followed by 388°C and 469°C. Maximum impact of this work will be obtained by performing TEM characterization of these neutron irradiated samples present at PNNL to correlate the measured hardening with microstructural features. The proposed project aims to perform a comprehensive TEM characterization of HT-9 heat (ORNL) as a function of irradiation temperature (241°C - 469°C) and dose (3- 8 dpa). TEM will used to evaluate control and irradiated samples and observe the general grain structure, prior austenite grains, martensite packet, lath structure and primary carbide structure, since these parameters can significantly affect the mechanical behavior and radiation resistance of HT-9. TEM will also be employed to study dislocation loops (size and number density), radiation-induced cavities, segregation, and phase transformation. For irradiated F-M steels, the increase in yield strength is quite steep up to around 10 dpa. Hence, a comprehensive understanding of the mechanical behavior and microstructural features of HT-9 after irradiation to lower doses (1-10 dpa) will be beneficial for more accurate extrapolation to higher doses. Efforts will be made to comprehensively understand the effects of radiation damage on HT-9 heats at the LWR and fast reactor relevant temperatures, and to develop appropriate property-structure-dose-temperature correlations. The results of the proposed work could be extended beyond HT-9, and it would be relevant to many F-M steels. Thus, the proposed work will have substantial implications for the deployment of next-generation advanced reactors. The project performance (TEM lamella preparation, TEM characterization, and analysis) is expected to take place during June-Dec 2024 and will result in one conference presentation and one journal article publication.

One of the goals of the DOE-NE R&D programs is to develop advanced structural materials for the next generation of reactors that have good neutronics, dimensional stability, corrosion resistance, mechanical and thermal properties under irradiation. HT-9 is being considered as a candidate structural material for advanced reactors, 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 (NSUF irradiation experiment). The mechanical properties of these alloys were recently evaluated under a funded RTE project. Maximum impact of this work will be obtained by performing TEM characterization of these neutron irradiated samples to correlate the measured hardening with microstructural features. The proposed project aims to perform a comprehensive TEM characterization of HT-9 heat (ORNL) as a function of irradiation temperature (241°C - 469°C) and dose (3- 8 dpa). TEM will used to evaluate control and irradiated samples and observe the general grain structure, prior austenite grains, martensite packet, lath structure and primary carbide structure, since these parameters can significantly affect the mechanical behavior and radiation resistance of HT-9. TEM will also be employed to study dislocation loops (size and number density), radiation-induced cavities, segregation, and phase transformation. For irradiated F-M steels, the increase in yield strength is quite steep up to around 10 dpa. Hence, a comprehensive understanding of the mechanical behavior and microstructural features of HT-9 after irradiation to lower doses (1-10 dpa) will be beneficial for more accurate extrapolation to higher doses. Efforts will be made to comprehensively understand the effects of radiation damage on HT-9 heats at the LWR and fast reactor relevant temperatures, and to develop appropriate property-structure-dose-temperature correlations. The results of the proposed work could be extended beyond HT-9, and it would be relevant to many F-M steels. Thus, the proposed work will have substantial implications for the deployment of next-generation advanced reactors.