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Title: Microstructure studies of interdiffusion behavior of U 3Si 2/Zircaloy-4 at 800 and 1000 °C

Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U 3Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000°C up to 100 hours. The microstructure of pristine U 3Si 2 and U 3Si 2/Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800°C is ZrSi 2, with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi 2 interface and Fe-Cr-U-Si phases at ZrSi 2/U-Si interface. As a result, the primary interdiffusion products at 1000°C were Zr 2Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U 6Fe.
Authors:
ORCiD logo [1] ; ORCiD logo [1] ;  [1] ;  [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Report Number(s):
INL-JOU-16-40511
Journal ID: ISSN 0022-3115; PII: S0022311516311114
Grant/Contract Number:
AC07-05ID14517; FOA-0000712
Type:
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 486; Journal Issue: C; Journal ID: ISSN 0022-3115
Publisher:
Elsevier
Research Org:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE)
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; accident tolerant fuel; diffusion; microstructure; nuclear fuels; zirconium alloy
OSTI Identifier:
1357771
Alternate Identifier(s):
OSTI ID: 1419388