journal article·2017-05Open access

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

Dennis D. Keiser, Jan-Fong Jue, Jian Gan, Brandon D. Miller, Adam B. Robinson, James W. Madden, M. Ross Finlay, Glenn Moore, Pavel Medvedev, Mitch Meyer

Abstract

Abstract 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 3 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.
Source: openaire · Quality: candidate unverified · Source record ↗

Research context

OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Fusion materials and technologies

SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.

References (28)

Show source-provided references
1. Irradiation performance of U-Mo monolithic fuel10.5516/NET.07.2014.706 ↗
2. Microstructure characterization of as-fabricated and 475°C annealed U–7 wt.% Mo dispersion fuel in Al–Si alloy matrix10.1016/j.jallcom.2011.07.048 ↗
3. Microstructural characterization of U–7Mo/Al–Si alloy matrix dispersion fuel plates fabricated at 500°C10.1016/j.jnucmat.2011.02.027 ↗
4. Electron microscopy characterization of an as-fabricated research reactor fuel plate comprised of U–7Mo particles dispersed in an Al–2Si alloy matrix10.1016/j.matchar.2010.07.010 ↗
5. Effects of irradiation on the microstructure of U–7Mo dispersion fuel with Al–2Si matrix10.1016/j.jnucmat.2012.01.013 ↗
6. “Microstructural development in irradiated U-7Mo/6061 Al alloy matrix dispersion fuel10.1016/j.jnucmat.2009.06.018 ↗
7. TEM characterization of U–7Mo/Al–2Si dispersion fuel irradiated to intermediate and high fission densities10.1016/j.jnucmat.2012.02.001 ↗
8. Scanning electron micrscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices10.5516/NET.07.2014.704 ↗
9. Phase changes in pile-irradiated uranium-base alloys10.1063/1.1722250 ↗
10. Effect of fission rate and lamella spacing upon the irradiation-induced phase transformation of U-9wt.% Mo alloy10.1016/0022-3115(59)90051-0 ↗
11. Scanning electron micrscopy analysis of fuel/matrix interaction layers in highly-irradiated U-Mo dispersion fuel plates with Al and Al-Si alloy matrices10.5516/NET.07.2014.704 ↗
12. Microstructural characterization of irradiated U–7Mo/Al–5Si dispersion fuel to high fission density10.1016/j.jnucmat.2014.08.052 ↗
13. Transmission electron microscopy characterization of irradiated U–7Mo/Al–2Si dispersion fuel10.1016/j.jnucmat.2009.11.015 ↗
14. Development and irradiation performance of uranium aluminide fuels in test reactors10.13182/NT80-A32515 ↗
15. Perez · 2011
16. Hayes · 2003
17. G.S. Chang, Idaho National Laboratory, 2008, (personal communication).
18. Kumar · 2005 · 11A3 · 1
19. Olander · 1976
20. Fission gas induced swelling in uranium at high temperatures and pressures10.1016/0022-3115(69)90246-3 ↗
21. Reversibility of fission gas swelling in uranium10.1016/0022-3115(69)90224-4 ↗
22. Formation and growth of intragranular fission gas bubbles in UO2 fuel with burnup of 6-83 Gwd/t10.1016/0022-3115(93)90229-R ↗
23. Fission gas release during postirradiation annealing of UO2-2 wt% Gd2O3 fuels10.1016/0022-3115(92)90534-R ↗
24. Golosov, O.A., Lyutikova, M.S., Semerikov V.B., and Teplykh, A.Y., The Effect of Fuel Burn-up Followed by Annealing on Changes in Structure and Structural Parameters of U-9% Mo Dispersion Fuel.” RRFM 2008 Transactions e Poster Session, Hamburg, Germany.
25. Measurement of fission gas release from irradiated U-Mo dispersion fuel samples10.1016/j.jnucmat.2016.05.039 ↗
26. Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation-Non-destructive analysis of the AFIP-1 fuel plates10.1016/j.jnucmat.2016.04.048 ↗
27. Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel10.1016/j.jnucmat.2014.05.052 ↗
28. Thermal stability of fission gas bubble superlattice in irradiated U–10Mo fuel10.1016/j.jnucmat.2015.04.023 ↗