Preliminary Exam Seminar: Nicole Rodriguez Perez

Event Date: December 1, 2023
Time: 8:30am
Location: ARMS 3115 or via WebEx
Priority: No
School or Program: Materials Engineering
College Calendar: Show

"Effects and Mitigation of FCCI in U-Zr Fuels for Fast Reactor Applications" 

Nicole Rodriguez Perez, MSE PhD Candidate 

Advisor: Professor Maria Okuniewski

WebEx Link


Advanced fast reactors represent the next generation of nuclear reactor technology; more reliable, safe, and sustainable than the conventional thermal reactors. Fast reactors use high energy neutrons (>1MeV), and do not rely in a moderator, such as water, to slow down the neutrons used in the fission reaction. Uranium-zirconium alloys, such as U-10wt.%Zr-19wt.%Pu and U-10wt.%Zr, with ferritic-martensitic HT9 cladding demonstrated great irradiation performance achieving up to 19 at.% burn-up without fuel failure. The high performance of the fuel pin was possible as a result of experimental experience that revealed the effects of Zr alloying in reducing the fuel cladding chemical interaction (FCCI) and increasing the fuel melting point, development of new cladding materials, and design improvements. However, the U-Zr fuels are affected by temperature changes and irradiation damage during operation. Characterization of U-10wt.%Zr fuels from historical performance-based experiments has revealed distinct four phenomena in the irradiated fuel: gaseous and solid fission product swelling, constituent redistribution, fuel-cladding mechanical interaction (FCMI), and FCCI.
Particularly, the FCCI was identified as the cause for failure of many of the fuel pins. The FCCI reduces the mechanical integrity of the cladding via reduction of the cladding thickness, changes in the chemical composition of the steel, and irradiation hardening. However, the formation of the FCCI is a complex phenomenon where thermal and irradiation variables interfere. There is evidence that the FCCI is correlated with temperature, as the higher FCCI regions have been observed at the higher temperature regions of the fuel pin. On the other hand, the influence of irradiation is not established. The irradiation-induced diffusion of the different elements for the fuel, the cladding, and the fission products is neither well understood. As a result, the data collected from historical experiments needs to be revised to identify possible gaps, and propose research that can assess the existent knowledge needs. The objective of the current review is to offer an overview of the current status on FCCI research, including possible strategies to eliminate or mitigate the FCCI.

2023-12-01 08:30:00 2023-12-01 09:30:00 America/Indiana/Indianapolis Preliminary Exam Seminar: Nicole Rodriguez Perez ARMS 3115 or via WebEx