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Title: Mechanical Properties of Irradiated U-10 wt. %Mo Alloy Degraded by Porosity Development

Journal Article · · Journal of Nuclear Engineering and Radiation Science
DOI: https://doi.org/10.1115/1.4064779 · OSTI ID:2349335
ORCiD logo [1];  [1];  [1];  [1];  [2];  [1];  [1];  [1];  [1]
  1. Idaho National Laboratory (INL), Idaho Falls, ID (United States)
  2. Idaho National Laboratory (INL), Idaho Falls, ID (United States); US Nuclear Regulatory Commission (NRC), Rockville, MD (United States)

In this study, a plate-type nuclear fuel consisting of a solid monolithic foil of U-10 wt. %Mo is under development for use in the United States' high-performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate postirradiation mechanical properties of the commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo samples to generate the fuel material properties, including the modulus of elasticity and the bending strength. Although the material behaves in a brittle manner due to the accumulated porosity, a general trend of strength and modulus reduction was found as fission density increases. The data produced was evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

Research Organization:
Idaho National Laboratory (INL), Idaho Falls, ID (United States)
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA)
Grant/Contract Number:
AC07-05ID14517
OSTI ID:
2349335
Report Number(s):
INL/JOU--23-75174-Rev000
Journal Information:
Journal of Nuclear Engineering and Radiation Science, Journal Name: Journal of Nuclear Engineering and Radiation Science Journal Issue: 3 Vol. 10; ISSN 2332-8983
Publisher:
ASMECopyright Statement
Country of Publication:
United States
Language:
English

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