Stable in-reactor performances at low temperature of U-10wt.%Mo dispersion fuel containing centrifugally atomized powder. Page: 2 of 10
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The two key issues are the reaction of the fuel alloy with the aluminum matrix and the
irradiation behavior of the dispersion. The former issue is important because excessive reaction
will consume the matrix aluminum and possibly a significant amount of the aluminum-alloy
cladding. The latter issue relates principally to the behavior of the fission gas in the fuel. If the
mobility of the fission gas is small enough, it will be contained in small bubbles that do not
interlink, as shown for U3Si2 [4-5]. Such a fuel will exhibit a steady but stable increase in
volume during irradiation. On the other hand, if the fission gas is very mobile, some bubbles
grow preferentially, becoming large and interlinking with adjacent bubbles, as shown for U3Si
[10-11]. If the volume loading of such fuel particles is high enough that a significant number of
particles are touching, the fission gas bubbles can interlink across many particles and lead to
unstable, rapid (breakaway) swelling.
Early irradiation experiments with uranium alloys showed the promise of acceptable irradiation
behavior, if these alloys could be maintained in their cubic y-U crystal structure [12]. If
centrifugally atomized U-Mo powder can retain this gamma uranium phase during fuel element
fabrication and irradiation, and if it is compatible with aluminum which forms the matrix of
dispersion fuels, the uranium alloy would be a prime candidate for dispersion fuel for research
reactors. Very-high-density atomized U-IOwt.%Mo powder prepared by centrifugal atomization
retained the isotropic high temperature y-U phase [13]. Moreover, the gamma phase did not
decompose into the equilibrium y-U and U2Mo two phase structure in U-IOwt.%Mo alloy after
annealing of up to 100 hours at 400 C. In addition, the U-IOwt.%Mo particles dispersed in
aluminum did not show significant dimensional changes after annealing up to 2,000 hours at
400 C, and interdiffusion between U-IOwt.%Mo and aluminum was found to be minimal [14].
In this study, in order to characterize the in-reactor performances of the atomized U-IOwt.%Mo
dispersion fuels, the U-IOwt.%Mo microplates have been irradiated to approximately 40at.% and
70at.% burn-up at low temperature.
2. Experimental procedure
Low enriched uranium lumps (99.9 pct pure) and molybdenum buttons (99.7 % pct pure) were
used for the preparation of the U-IOwt.%Mo powders by rotating-disk centrifugal atomization
[13]. Dispersion fuel meats with a nominal volume fraction of 25% were prepared by blending
the U-IOwt.%Mo and aluminum powder and by rolling the blended powders at a working
temperature of about 485 C.
The microplate fuel samples were fabricated with external dimensions of 76 mm x 22 mm x
1.3 mm in aluminum cladding. The fuel zone is elliptical in shape with major and minor axes of
approximately 51 mm and 9.5 mm, respectively; the fuel zone thickness is nominally 0.5 mm.
U-IOwt.%Mo microplates, which were irradiated at an average fuel center temperature of 65 C,
were discharged after 94 effective full-power days (EFPDs) of irradiation, and then discharged
after 232 EFPDs of irradiation, achieving (calculated) 235U burnups of 40at.% and 70at.%,
independently. Thereafter, post-irradiation examinations of the microplates are performed,
primarily using a scanning electron microscope.
3. Experimental results
1999 International Meeting on Reduced Enrichment for Research and Test Reactors, Budapest, Hungary, October 3-8, 1999
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Kim, K. H.; Park, J. M.; Kim, C. K.; Hofman, G. L.; Meyer, M. K. & Snelgrove, J. L. Stable in-reactor performances at low temperature of U-10wt.%Mo dispersion fuel containing centrifugally atomized powder., article, September 11, 2001; Illinois. (https://digital.library.unt.edu/ark:/67531/metadc723912/m1/2/: accessed April 19, 2024), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.