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

Abstract

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop 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. The program is focused on assisting with the 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. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of themore » modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

Authors:
; ;
Publication Date:
Research Org.:
Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1253830
Report Number(s):
PNNL-SA-113336
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: 473
Country of Publication:
United States
Language:
English
Subject:
nuclear fuel; uranium-molybdenum; thermal conductivity; dispersion

Citation Formats

Burkes, Douglas E., Huber, Tanja K., and Casella, Andrew M.. A model to predict thermal conductivity of irradiated U–Mo dispersion fuel. United States: N. p., 2016. Web. doi:10.1016/j.jnucmat.2016.01.012.
Burkes, Douglas E., Huber, Tanja K., & Casella, Andrew M.. A model to predict thermal conductivity of irradiated U–Mo dispersion fuel. United States. doi:10.1016/j.jnucmat.2016.01.012.
Burkes, Douglas E., Huber, Tanja K., and Casella, Andrew M.. Sun . "A model to predict thermal conductivity of irradiated U–Mo dispersion fuel". United States. doi:10.1016/j.jnucmat.2016.01.012.
@article{osti_1253830,
title = {A model to predict thermal conductivity of irradiated U–Mo dispersion fuel},
author = {Burkes, Douglas E. and Huber, Tanja K. and Casella, Andrew M.},
abstractNote = {The Office of Materials Management and Minimization Reactor Conversion Program continues to develop 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. The program is focused on assisting with the 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. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.},
doi = {10.1016/j.jnucmat.2016.01.012},
journal = {Journal of Nuclear Materials},
number = ,
volume = 473,
place = {United States},
year = {Sun May 01 00:00:00 EDT 2016},
month = {Sun May 01 00:00:00 EDT 2016}
}