Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing
Abstract
The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.
- Authors:
-
- Colorado School of Mines, Golden, CO (United States)
- Oregon State Univ., Corvallis, OR (United States)
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Publication Date:
- Research Org.:
- Idaho National Laboratory (INL), Idaho Falls, ID (United States)
- Sponsoring Org.:
- USDOE
- OSTI Identifier:
- 1294498
- Alternate Identifier(s):
- OSTI ID: 1341165
- Report Number(s):
- INL/JOU-15-35414
Journal ID: ISSN 0022-3115; PII: S0022311516301015
- Grant/Contract Number:
- AC07-05ID14517
- Resource Type:
- Journal Article: Accepted Manuscript
- Journal Name:
- Journal of Nuclear Materials
- Additional Journal Information:
- Journal Volume: 475; Journal Issue: C; Journal ID: ISSN 0022-3115
- Publisher:
- Elsevier
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 36 MATERIALS SCIENCE; 97 MATHEMATICS AND COMPUTING; nuclear fuel; MATLAB; automated image analysis; fission bubbles; fission density; porosity
Citation Formats
Collette, R., King, J., Buesch, C., Keiser, Jr., D. D., Williams, W., Miller, B. D., and Schulthess, J. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing. United States: N. p., 2016.
Web. doi:10.1016/j.jnucmat.2016.03.028.
Collette, R., King, J., Buesch, C., Keiser, Jr., D. D., Williams, W., Miller, B. D., & Schulthess, J. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing. United States. https://doi.org/10.1016/j.jnucmat.2016.03.028
Collette, R., King, J., Buesch, C., Keiser, Jr., D. D., Williams, W., Miller, B. D., and Schulthess, J. 2016.
"Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing". United States. https://doi.org/10.1016/j.jnucmat.2016.03.028. https://www.osti.gov/servlets/purl/1294498.
@article{osti_1294498,
title = {Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing},
author = {Collette, R. and King, J. and Buesch, C. and Keiser, Jr., D. D. and Williams, W. and Miller, B. D. and Schulthess, J.},
abstractNote = {The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.},
doi = {10.1016/j.jnucmat.2016.03.028},
url = {https://www.osti.gov/biblio/1294498},
journal = {Journal of Nuclear Materials},
issn = {0022-3115},
number = C,
volume = 475,
place = {United States},
year = {Fri Apr 01 00:00:00 EDT 2016},
month = {Fri Apr 01 00:00:00 EDT 2016}
}
Web of Science
Works referenced in this record:
Fission gas bubble identification using MATLAB's image processing toolbox
journal, August 2016
- Collette, R.; King, J.; Keiser, D.
- Materials Characterization, Vol. 118
Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements
journal, August 2009
- Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing
- Journal of Nuclear Materials, Vol. 392, Issue 3
Microstructural characterization of irradiated U–7Mo/Al–5Si dispersion fuel to high fission density
journal, November 2014
- Gan, J.; Miller, B. D.; Keiser, D. D.
- Journal of Nuclear Materials, Vol. 454, Issue 1-3
TEM characterization of U–7Mo/Al–2Si dispersion fuel irradiated to intermediate and high fission densities
journal, May 2012
- Gan, J.; Keiser, D. D.; Miller, B. D.
- Journal of Nuclear Materials, Vol. 424, Issue 1-3
Comparative Analysis of Structural Changes in u-mo Dispersed fuel of Full-Size fuel Elements and Mini-Rods Irradiated in the mir Reactor
journal, December 2013
- Izhutov, Aleksey. L.; Iakovlev, Valeriy. V.; Novoselov, Andrey. E.
- Nuclear Engineering and Technology, Vol. 45, Issue 7
Effects of irradiation on the microstructure of U–7Mo dispersion fuel with Al–2Si matrix
journal, June 2012
- Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.
- Journal of Nuclear Materials, Vol. 425, Issue 1-3
Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix
journal, June 2012
- Kim, Yeon Soo; Hofman, G. L.
- Journal of Nuclear Materials, Vol. 425, Issue 1-3
IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL
journal, April 2014
- Meyer, M. K.; Gan, J.; Jue, J. F.
- Nuclear Engineering and Technology, Vol. 46, Issue 2
Advantages and disadvantages of using a focused ion beam to prepare TEM samples from irradiated U–10Mo monolithic nuclear fuel
journal, May 2012
- Miller, B. D.; Gan, J.; Madden, J.
- Journal of Nuclear Materials, Vol. 424, Issue 1-3
Conjugate heat transfer simulations of advanced research reactor fuel
journal, July 2014
- Piro, M. H. A.; Leitch, B. W.
- Nuclear Engineering and Design, Vol. 274
Adaptive document image binarization
journal, February 2000
- Sauvola, J.; Pietikäinen, M.
- Pattern Recognition, Vol. 33, Issue 2
Works referencing / citing this record:
Determination of the degree of grain refinement in irradiated U-Mo fuels
journal, December 2018
- Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.
- Heliyon, Vol. 4, Issue 12