TEM characterization of irradiated U-7Mo/Mg dispersion fuel

Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J. F.; Robinson, A. B.; Madden, J.

doi:10.1016/j.jnucmat.2017.07.030

Title: TEM characterization of irradiated U-7Mo/Mg dispersion fuel

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Abstract

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10²¹ f/cm³, 7.4 × 10¹⁴ f/cm³/s and 123 °C, and 5.5 × 10²¹ f/cm3, 11.0 × 10¹⁴ f/cm³/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al₃Mg₂ and Al₁₂Mg₁₇ along with precipitates of MgO, Mg₂Si and FeAl_5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubblemore »« less

Authors:

^[1]; Keiser, D. D. ^[1]; Miller, B. D. ^[1];

^[1]; Robinson, A. B. ^[1]; Madden, J. ^[1]

Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced PIE and Characterization Division

Publication Date:: Sat Jul 15 00:00:00 EDT 2017

Research Org.:: Idaho National Lab. (INL), Idaho Falls, ID (United States)

Sponsoring Org.:: USDOE National Nuclear Security Administration (NNSA)

OSTI Identifier:: 1374736

Alternate Identifier(s):: OSTI ID: 1495573

Report Number(s):: INL/JOU-17-41167
Journal ID: ISSN 0022-3115; PII: S0022311517304105

Grant/Contract Number:: AC07-05ID14517; NA-212

Resource Type:: Accepted Manuscript

Journal Name:: Journal of Nuclear Materials

Additional Journal Information:: Journal Volume: 494; 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; Fission; Interface; Mg matrix; Microstructure; TEM; U-7Mo

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                    Gan, J., Keiser, D. D., Miller, B. D., Jue, J. F., Robinson, A. B., and Madden, J. TEM characterization of irradiated U-7Mo/Mg dispersion fuel.  United States: N. p., 2017. 
Web.  doi:10.1016/j.jnucmat.2017.07.030.

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                    Gan, J., Keiser, D. D., Miller, B. D., Jue, J. F., Robinson, A. B., & Madden, J. TEM characterization of irradiated U-7Mo/Mg dispersion fuel.  United States.  https://doi.org/10.1016/j.jnucmat.2017.07.030

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                    Gan, J., Keiser, D. D., Miller, B. D., Jue, J. F., Robinson, A. B., and Madden, J. Sat .  
"TEM characterization of irradiated U-7Mo/Mg dispersion fuel".  United States.  https://doi.org/10.1016/j.jnucmat.2017.07.030.  https://www.osti.gov/servlets/purl/1374736.

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@article{osti_1374736,

  title        = {TEM characterization of irradiated U-7Mo/Mg dispersion fuel},

  author       = {Gan, J. and Keiser, D. D. and Miller, B. D. and Jue, J. F. and Robinson, A. B. and Madden, J.},

  abstractNote = {This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 1021 f/cm3, 7.4 × 1014 f/cm3/s and 123 °C, and 5.5 × 1021 f/cm3, 11.0 × 1014 f/cm3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al3Mg2 and Al12Mg17 along with precipitates of MgO, Mg2Si and FeAl5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.},

  doi          = {10.1016/j.jnucmat.2017.07.030},

  journal      = {Journal of Nuclear Materials},

  number       = C,

  volume       = 494,

  place        = {United States},

  year         = {Sat Jul 15 00:00:00 EDT 2017},

  month        = {Sat Jul 15 00:00:00 EDT 2017}

}

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