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Title: A model to predict failure of irradiated U–Mo dispersion fuel

Abstract

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials of interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused moremore » rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

Authors:
; ;
Publication Date:
Research Org.:
Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1343178
Report Number(s):
PNNL-SA-115734
Journal ID: ISSN 0029-5493; DN3001010
DOE Contract Number:  
AC05-76RL01830
Resource Type:
Journal Article
Resource Relation:
Journal Name: Nuclear Engineering and Design; Journal Volume: 310; Journal Issue: C
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; composite materials; nuclear reactor materials; dispersion fuel; failure analysis

Citation Formats

Burkes, Douglas E., Senor, David J., and Casella, Andrew M.. A model to predict failure of irradiated U–Mo dispersion fuel. United States: N. p., 2016. Web. doi:10.1016/j.nucengdes.2016.09.032.
Burkes, Douglas E., Senor, David J., & Casella, Andrew M.. A model to predict failure of irradiated U–Mo dispersion fuel. United States. doi:10.1016/j.nucengdes.2016.09.032.
Burkes, Douglas E., Senor, David J., and Casella, Andrew M.. Thu . "A model to predict failure of irradiated U–Mo dispersion fuel". United States. doi:10.1016/j.nucengdes.2016.09.032.
@article{osti_1343178,
title = {A model to predict failure of irradiated U–Mo dispersion fuel},
author = {Burkes, Douglas E. and Senor, David J. and Casella, Andrew M.},
abstractNote = {Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials of interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.},
doi = {10.1016/j.nucengdes.2016.09.032},
journal = {Nuclear Engineering and Design},
number = C,
volume = 310,
place = {United States},
year = {Thu Dec 01 00:00:00 EST 2016},
month = {Thu Dec 01 00:00:00 EST 2016}
}