NSUF 19-1635: Atom probe tomography study of the fuel cladding chemical interaction (FCCI) layer in irradiated U-10Zr fuel with HT-9 cladding

This project aims to understand the fuel cladding chemical interaction (FCCI) from a Fast Flux Test Facility (FFTF) irradiated U-10Zr fuel with ferritic/martensitic steel HT-9 clad using atom probe tomography (APT). The goal of this project is to advance the DOE Office of Nuclear Energy’s fast reactor campaign by understanding the performance of HT-9 cladding. APT will be used to quantify the microchemistry changes at lath boundaries, grain boundaries, and phase boundaries, and chemical composition of precipitates observed in the FCCI layers. The outcomes of this project will provide critical insights on the FCCI phenomena for the U-Zr/U-Pu-Zr fuel and HT9 clad system.This project is a continuing effort of several RTEs trying to fully understand the performance of the U-10Zr fuel and HT-9 cladding system using state-of-the-art characterization techniques. The material selected for this project comes from the Mechanistic Fuel Failure (MFF) series of metallic fuel irradiations performed in FFTF. The cross section from a MFF-3 pin (#193045) was taken at an axial location near the top of the fuel column length (X/L=0.98). The U-10Zr fuel slugs were cast into quartz molds to achieve 75% smeared density in relation to the cladding inner diameter and sodium bounded to the HT-9 cladding. This specific sample is readily available for PIE work and hence this project can begin immediately with no anticipated delays that could slow project completion. Preliminary scanning electron microscopy (SEM) characterization has also been performed. The fuel constituent redistribution was confirmed by roughly three distinct radial regions with differing Zr content. Non-uniform FCCI layers and regions with no apparent FCCI were also identified along the circumference of the HT-9 cladding. HT-9 is a Fe-12Cr ferritic/martensitic steel that showed exceptional swelling resistance during previous irradiation tests in FFTF, EBR-II, and HFIR. It is important to determine after FCCI with U-10Zr fuel, whether HT-9 could maintain its microstructural stability and whether the FCCI could impact the swelling resistance or induce brittle phases in the interaction zones.

추가 정보

필드	값
Abstract	This project aims to understand the fuel cladding chemical interaction (FCCI) from a Fast Flux Test Facility (FFTF) irradiated U-10Zr fuel with ferritic/martensitic steel HT-9 clad using atom probe tomography (APT). The goal of this project is to advance the DOE Office of Nuclear Energy’s fast reactor campaign by understanding the performance of HT-9 cladding. APT will be used to quantify the microchemistry changes at lath boundaries, grain boundaries, and phase boundaries, and chemical composition of precipitates observed in the FCCI layers. The outcomes of this project will provide critical insights on the FCCI phenomena for the U-Zr/U-Pu-Zr fuel and HT9 clad system.This project is a continuing effort of several RTEs trying to fully understand the performance of the U-10Zr fuel and HT-9 cladding system using state-of-the-art characterization techniques. The material selected for this project comes from the Mechanistic Fuel Failure (MFF) series of metallic fuel irradiations performed in FFTF. The cross section from a MFF-3 pin (#193045) was taken at an axial location near the top of the fuel column length (X/L=0.98). The U-10Zr fuel slugs were cast into quartz molds to achieve 75% smeared density in relation to the cladding inner diameter and sodium bounded to the HT-9 cladding. This specific sample is readily available for PIE work and hence this project can begin immediately with no anticipated delays that could slow project completion. Preliminary scanning electron microscopy (SEM) characterization has also been performed. The fuel constituent redistribution was confirmed by roughly three distinct radial regions with differing Zr content. Non-uniform FCCI layers and regions with no apparent FCCI were also identified along the circumference of the HT-9 cladding. HT-9 is a Fe-12Cr ferritic/martensitic steel that showed exceptional swelling resistance during previous irradiation tests in FFTF, EBR-II, and HFIR. It is important to determine after FCCI with U-10Zr fuel, whether HT-9 could maintain its microstructural stability and whether the FCCI could impact the swelling resistance or induce brittle phases in the interaction zones.
Award Announced Date	2019-02-08T00:00:00
Awarded Institution	None
Facility	None
Facility Tech Lead	Yaqiao Wu
Irradiation Facility	None
PI	Xiang Liu
PI Email	[email protected]
Project Type	RTE
RTE Number	1635