Measurement of fission gas release from irradiated U Mo dispersion fuel samples

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

doi:10.1016/j.jnucmat.2016.05.039

Cited by 7 publications

(1 citation statement)

References 16 publications

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“…Additionally, results of fission gas release studies have been reported for irradiated U-Mo dispersion fuel samples with AA6061 cladding, but no microstructural characterization was reported for the samples that were tested to 1,000˚C. [25] The samples were tested at a constant heating rate and the temperatures where fission gas was released was determined. Burkes and between 500 and 670˚C another 57.0 % of the fission gas was released.…”

Section: Discussionmentioning

confidence: 99%

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

Keiser

Jue

Gan

et al. 2017

Journal of Nuclear Materials

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The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500˚C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

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Section: Discussionmentioning

confidence: 99%