NSUF 21-4269: Pre-oxidation effect on ATF cladding performance by characterization of irradiated FeCrAl-UO2 capsules
FeCrAl alloys are selected as candidate nuclear fuel cladding materials for the design and development of accident tolerant fuel (ATF) over the current Zr-based alloys for the light water reactors (LWRs) due to its high temperature oxidation resistance, good radiation tolerance and mechanical properties. The next key step is to evaluate the coupling of ATF fuel-cladding. This requires the understanding of Fuel-Cladding Chemical and Mechanical Interactions (FCCI/FCMI) by examination of the fuel-cladding interfaces. The controlling mechanisms on the FeCrAl cladding performance involve complexities of radiation-induced microstructure (loops and α’ precipitates), chemistry, pre-oxidation, grain size (dislocation and crack along the grain boundary) and fission products (Cs, Br, Sr, etc.). The cladding attack FCCI/FCMI mechanisms are the combination of production and transport from the fuel to the clad and interfacial diffusion at the fuel-cladding interface. Furthermore, formation of loops and α’ alter the vacancy/interstitial flux at the interface or grain boundary, result in irradiation hardening and initiate cracking. A phenomenological FCCI model in UO2-FeCrAl needs to consider the microstructure, product transport and interfacial diffusion. The objective of this project is to understand the pre-oxidation effect on the FCCI/FCMI of irradiated UO2-FeCrAl capsule dependent on microstructure.
We are investigating the phenomena through a scanning electron microscopy (SEM)/scanning transmission electron microscopy (STEM) characterization and transmission Kikuchi diffraction (tKD) study of irradiated UO2-FeCrAl capsules to compare the irradiation-induced microstructure of dislocation loops, α’ precipitates and fission products at the fuel-cladding interface. These results will reveal underlying mechanisms of the FCCI and FCMI phenomena. Our work will target high-value UO2-FeCrAl capsules that were irradiated in the Advanced Test Reactor (ATR). Specifically, three capsules with C35M alloys of industry interest (GE) irradiated at ~360°C with as machined (AM)/pre-oxidized (PO) conditions will be studied. These conditions are relevant to early-life conditions of light water reactors (300~350°C). The C35M alloys have nominal compositions of Fe-13Cr-5Al-2Mo-0.1Si (in wt.%). TEM lamellae will be fabricated from each of these conditions with a focus on the interface. We are requesting access to ORNL LAMDA for 5 days of focused ion beam (FIB) time to prepare lamella, 3 days to perform SEM/tkD scan and 6 days of STEM time to conduct STEM-energy dispersive x-ray spectroscopy (EDS) on the FeCrAl-Al2O3-UO2 interface and dislocation imaging. The project outcome is a systematic and mechanistic understanding of microstructural evolution and FCCI/FCMI in irradiated UO2-FeCrAl capsules; the broader impact of this work will lead to a science-based design and selection of FeCrAl alloys to maximize microstructural, microchemical and interfacial stability under irradiation.
Additional Info
| Field | Value |
|---|---|
| Abstract | FeCrAl alloys are selected as candidate nuclear fuel cladding materials for the design and development of accident tolerant fuel (ATF) over the current Zr-based alloys for the light water reactors (LWRs) due to its high temperature oxidation resistance, good radiation tolerance and mechanical properties. The next key step is to evaluate the coupling of ATF fuel-cladding. This requires the understanding of Fuel-Cladding Chemical and Mechanical Interactions (FCCI/FCMI) by examination of the fuel-cladding interfaces. The controlling mechanisms on the FeCrAl cladding performance involve complexities of radiation-induced microstructure (loops and α’ precipitates), chemistry, pre-oxidation, grain size (dislocation and crack along the grain boundary) and fission products (Cs, Br, Sr, etc.). The cladding attack FCCI/FCMI mechanisms are the combination of production and transport from the fuel to the clad and interfacial diffusion at the fuel-cladding interface. Furthermore, formation of loops and α’ alter the vacancy/interstitial flux at the interface or grain boundary, result in irradiation hardening and initiate cracking. A phenomenological FCCI model in UO2-FeCrAl needs to consider the microstructure, product transport and interfacial diffusion. The objective of this project is to understand the pre-oxidation effect on the FCCI/FCMI of irradiated UO2-FeCrAl capsule dependent on microstructure. We are investigating the phenomena through a scanning electron microscopy (SEM)/scanning transmission electron microscopy (STEM) characterization and transmission Kikuchi diffraction (tKD) study of irradiated UO2-FeCrAl capsules to compare the irradiation-induced microstructure of dislocation loops, α’ precipitates and fission products at the fuel-cladding interface. These results will reveal underlying mechanisms of the FCCI and FCMI phenomena. Our work will target high-value UO2-FeCrAl capsules that were irradiated in the Advanced Test Reactor (ATR). Specifically, three capsules with C35M alloys of industry interest (GE) irradiated at ~360°C with as machined (AM)/pre-oxidized (PO) conditions will be studied. These conditions are relevant to early-life conditions of light water reactors (300~350°C). The C35M alloys have nominal compositions of Fe-13Cr-5Al-2Mo-0.1Si (in wt.%). TEM lamellae will be fabricated from each of these conditions with a focus on the interface. We are requesting access to ORNL LAMDA for 5 days of focused ion beam (FIB) time to prepare lamella, 3 days to perform SEM/tkD scan and 6 days of STEM time to conduct STEM-energy dispersive x-ray spectroscopy (EDS) on the FeCrAl-Al2O3-UO2 interface and dislocation imaging. The project outcome is a systematic and mechanistic understanding of microstructural evolution and FCCI/FCMI in irradiated UO2-FeCrAl capsules; the broader impact of this work will lead to a science-based design and selection of FeCrAl alloys to maximize microstructural, microchemical and interfacial stability under irradiation. |
| Award Announced Date | 2021-06-07T16:16:55.41 |
| Awarded Institution | Idaho National Laboratory |
| Facility | Advanced Test Reactor |
| Facility Tech Lead | Alina Zackrone, Kory Linton |
| Irradiation Facility | None |
| PI | Vipul Gupta |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 4269 |