TY - JOUR
T1 - Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation
AU - Keiser, Dennis D.
AU - Jue, Jan Fong
AU - Woolstenhulme, Nicolas E.
AU - Ewh, Ashley
N1 - Funding Information:
This work was supported by the US Department of Energy, Office of Nuclear Materials Threat Reduction (NA-212), National Nuclear Security Administration, and this manuscript authored by a contractor of the US Government under DOE-NE Idaho Operations Office Contract DE-AC07-05ID14517. Accordingly, The US Government retains and the publisher, by accepting the article for publication, acknowledges that the US Government retains a nonexclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US Government purposes.
PY - 2011/12
Y1 - 2011/12
N2 - Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525°C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475°C.
AB - Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525°C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475°C.
UR - https://www.scopus.com/pages/publications/80053492986
U2 - 10.1016/j.jnucmat.2011.08.039
DO - 10.1016/j.jnucmat.2011.08.039
M3 - Article
AN - SCOPUS:80053492986
SN - 0022-3115
VL - 419
SP - 226
EP - 234
JO - Journal of Nuclear Materials
JF - Journal of Nuclear Materials
IS - 1-3
ER -