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Micro-tensile testing of neutron irradiated Al/Zr and Zr/U-Mo Diffusion Bonds

Journal Article · · Journal of Nuclear Materials
 [1];  [1];  [1];  [1]
  1. Idaho National Laboratory
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This study focuses on micro-tensile testing of neutron-irradiated Al/Zr and Zr/U-Mo diffusion bonds, which are integral to the structure and performance of nuclear fuel plates used in the fuel being developed for U.S. high-power research reactors. The fuel consists of uranium metal alloyed with 10 wt% molybdenum (U-10Mo), and diffusion bonded with zirconium and aluminum. Two distinct diffusion bonds occur—one at the Zr/U-10Mo interface via roll bonding, and the Zr/Al interface via hot isostatic pressing at 833 K. Understanding the post-irradiation mechanical behavior of these diffusion layers is critical to ensuring the fuel's stability and integrity. While previous studies have shown that U-10Mo loses ductility and becomes brittle due to irradiation-induced porosity, the mechanical properties of the diffusion layers remain largely unknown. This work utilizes micro-tensile testing specimens approximately 7 × 7 × 20 µm3, which are sufficiently large to encompass the identified diffusion zones to evaluate the failure strength and identify the failure location. For the Zr/Al interface, all failures occurred in the bulk aluminum. The average yield strength, ultimate strength, and strain at failure for the two parent samples were: 125?±?10?MPa, 139?±?24?MPa, 156?±?21?MPa, 185?±?14?MPa, 37.9?±?5.6%, and 29.7?±?3.4%, respectively. For specimens across the Zr/U-Mo interface, failures occurred either in the bulk Zr or U-Mo, with no failures observed in the diffusion region. These findings support the development of accurate thermo-mechanical models for irradiated fuel performance by identifying the mechanical limits and failure locations within the bonded regions.
Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE)
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
3013850
Report Number(s):
INL/JOU-25-87393
Journal Information:
Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: 0 Vol. 619
Country of Publication:
United States
Language:
English

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Related Subjects

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Aluminum
Micro-tensile
U-Mo
Zirconium
diffusion bond