The objective of this project is to acquire a multiscale mechanistic understanding of grain boundary cohesion and fracture in irradiated ferritic/martensitic (F/M) alloys, through transmission electron microscopic (TEM) in situ cantilever testing. F/M steels are candidates for cladding and structural components of advanced nuclear reactors. It is well known, however, that these materials exhibit grain boundary radiation-induced segregation (RIS) of Cr and minor alloying elements. RIS changes the grain boundary cohesive energy, which consequently affects the fracture behavior of the material. This is an inherently multiscale problem, as RIS is a point defect-driven process, grain boundary cohesion a microscale challenge, and fracture a macroscopic consequence. Previously, RIS and fracture experiments had to be conducted ex situ. Now, TEM mechanical testing enables in situ testing of RIS, crack propagation, and fracture. Thus, we hypothesize that TEM in situ microcantilever tests are the ideal tool for understanding fracture mechanisms across the length scales. In this work, we propose to conduct TEM in situ microcantilever tests of three irradiated commercial F/M alloys T91, HCM12A, and HT9. We will observe and record video of crack propagation. Grain boundary RIS measurements will connect the cracking to atomistic models of grain boundary chemistry and cohesion. Microcantilever tests will also generate load-displacement-time data that can be used to calculate cohesive laws governing macroscopic fracture. In summation, this experimental approach will link atomic, micro, and macro length scales, providing tremendous insight into the mechanisms governing fracture of irradiated F/M alloys. This work will also provide validation for finite element models in MOOSE.

The primary mission of the Department of Energy Office of Nuclear Energy (DOE-NE) is to advance nuclear power as a resource capable of meeting the nation’s energy, environmental, and national security needs. DOE-NE base programs on Advanced Reactor Technologies (ART) and Small Modular Reactor Technologies (SMRT) help fulfill this mission by coupling high-efficiency power generation with the potential for enhanced safety and security. Another DOE-NE base program on Light Water Reactor Technologies (LWRT) will ensure safe continued operation of our existing fleet of light water reactors. However, life extensions of light water reactors and the promise of advanced reactor and small modular reactor designs, are accompanied by the challenge of finding suitable structural and cladding materials that will withstand the harsh in-reactor operating conditions. Ensuring the mechanical integrity of these materials under high temperatures, corrosive environments, cyclic loading, and high irradiation damage, is paramount to the safety, performance, and long-term success of these three DOE-NE base programs.



This project aims to assist in the understanding of the mechanical integrity of candidate cladding and structural materials under irradiation, a goal that directly addresses the DOE-NE base programs and overarching mission. This project also addresses three DOE-NE research objectives: develop sustainable fuel cycles, minimize the risks of nuclear proliferation, and develop improvements in the affordability of new reactors.



DOE-NE is also very interested in advanced characterization techniques, evidenced by their recent call for proposals. The project proposed herein is directly applicable to this topic. Because results of this project are relevant to multiple reactor systems, and can be extended to other irradiated metallic alloy systems, this project is also inherently crosscutting across multiple DOE-NE base programs. Therefore, this project addresses the DOE-NE mission of meeting the country’s energy, environmental, and security needs with nuclear power.