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Title: The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

Abstract

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.

Authors:
; ;
Publication Date:
Research Org.:
Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1361009
Report Number(s):
PNNL-SA-121666
Journal ID: ISSN 0022-3115; DN3001010
DOE Contract Number:  
AC05-76RL01830
Resource Type:
Journal Article
Resource Relation:
Journal Name: Journal of Nuclear Materials; Journal Volume: 486
Country of Publication:
United States
Language:
English
Subject:
U-Mo; fi; irradiation; cl

Citation Formats

Burkes, Douglas E., Casella, Amanda J., and Casella, Andrew M. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel. United States: N. p., 2017. Web. doi:10.1016/j.jnucmat.2017.01.016.
Burkes, Douglas E., Casella, Amanda J., & Casella, Andrew M. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel. United States. doi:10.1016/j.jnucmat.2017.01.016.
Burkes, Douglas E., Casella, Amanda J., and Casella, Andrew M. Sat . "The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel". United States. doi:10.1016/j.jnucmat.2017.01.016.
@article{osti_1361009,
title = {The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel},
author = {Burkes, Douglas E. and Casella, Amanda J. and Casella, Andrew M.},
abstractNote = {The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.},
doi = {10.1016/j.jnucmat.2017.01.016},
journal = {Journal of Nuclear Materials},
number = ,
volume = 486,
place = {United States},
year = {Sat Apr 01 00:00:00 EDT 2017},
month = {Sat Apr 01 00:00:00 EDT 2017}
}