Scanning electron microscopy examination of a Fast Flux Test Facility irradiated U-10Zr fuel cross section clad with HT-9

Harp, Jason; Porter, Douglas L.; Miller, Brandon; Trowbridge, Tammy L.; Carmack, Jon

doi:10.1016/j.jnucmat.2017.07.040

Cited by 63 publications

(28 citation statements)

References 21 publications

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“…This was also observed in a previous diffusion couple [19]. This is not the case during irradiation, though, as a recent post-irradiation examination of U-10Zr shows [20]. Figure 13 in Ref.…”

Section: Ce-fe Diffusion Couplesupporting

confidence: 67%

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Thermodynamic stability studies of Ce-Sb compounds with Fe

Xie

Zhang

Benson

et al. 2018

Journal of Nuclear Materials

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Lanthanide fission products can migrate to the fuel periphery and react with cladding, causing fuelcladding chemical interaction (FCCI). Adding a fuel additive dopant, such as Sb, can bind lanthanide, such as Ce, into metallic compounds and thus prevent migration. The present study focuses on the thermodynamic stability of Ce-Sb compounds when in contact with the major cladding constituent Fe by conducting diffusion couple tests. Ce-Sb compounds have shown high thermodynamic stability as they did not react with Fe. When Fe-Sb compounds contacted with Ce, Sb was separated out of Fe-Sb compounds and formed the more stable Ce-Sb compounds.

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Section: Ce-fe Diffusion Couplesupporting

confidence: 67%

“…Figure 13 in Ref. [20] clearly shows lanthanide diffusion into the cladding, as well as Fe diffusion into the fuel. This difference between irradiation and out-of-pile testing is likely caused by the temperature gradient present during irradiation, with the hot zone in the center of the fuel.…”

Section: Ce-fe Diffusion Couplementioning

confidence: 98%

Thermodynamic stability studies of Ce-Sb compounds with Fe

Xie

Zhang

Benson

et al. 2018

Journal of Nuclear Materials

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“…In many ways this experiment is similar to AFC-2E [40], but the irradiation conditions are better known. The FCCI in Figure 69 is very similar to that seen in high temperature areas of steady state irradiations [37,41].…”

Section: Metallographysupporting

confidence: 62%

“…Palladium has been proposed as an additive to metallic fuel to help prevent lanthanide-initiated FCCI in U-Zr alloys at high burnup [35,36]. This concept was tested in out-of-pile tests, and lanthanide-palladium precipitates have been observed in bot U-5Fs and irradiated U-10Zr [37]. During fabrication, Zr and Pd form a stable intermetallic compound, and Pd was expected to release Zr and form stable intermetallic compounds with lanthanides.…”

Section: Pd Bearing Fuelmentioning

confidence: 99%

Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments

Harp

Capriotti

Cappia

2018

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The AFC-3C, AFC-3D, and AFC-4A capsule irradiations were irradiated at the Idaho National Laboratory Advanced Test Reactor. These irradiations were planned to test several different fast reactor fuels that could be used to facilitate ultra-high burnup applications in sodium fast reactors. Several different alloys, fuel geometries, bonding materials, and the use of additives were tested in ferritic-martensitic HT-9 cladding. The AFC-3C and 3D experiment took various different U-Mo, U-Zr, U-Pd-Zr, U-Mo-Ti-Zr, and U-Mo-Ti-Zr-Pd alloys to burnups of 2.1% to 4.5% FIMA (fissions per initial heavy metal atoms) and fission densities of 6.13x10 20 to 1.27x10 21 fissions / cm 3 . These alloys were evaluated against the historical fuel performance of previously irradiated fuel from literature. Fuel performance was irradiated through a suite of postirradiation examination techniques including neutron radiography, gamma spectrometry, dimensional inspection, fission gas release measurements, chemical burnup analysis, and optical microscopy. The irradiations were performed at relatively aggressive temperatures that resulted in significant interaction between the cladding and the fuel in many cases. In spite of this, there were no in-pile cladding breaches in these tests. The U-Zr alloys performed better than the U-Mo or U-Mo-Ti-Zr alloys. Some of the fuel-cladding interaction issues seen in helium bonded annular fuel that was not machined to tight tolerances appear to have been resolved by machining the outer diameter of annular fuel in this irradiation. Higher Zr concentrations in additive bearing fuel appears to have improved the performance of Pd additive fuel.Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments iv

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“…This point emphasizes the need for an accurate and thorough pedigree on any property measurements of this type. When it comes to properties of irradiated U-Zr and U-Pu-Zr alloys, there are few publications that describe the irradiated microstructure of U-Zr phases using both neutrons and He+ ions [71][72][73] and even fewer describing irradiated U-Pu-Zr alloys. 74 As for thermal properties, INL conducted first-ofits-kind bulk thermal diffusivity and mesoscale thermal conductivity measurements on irradiated U-10 wt% Zr using laser flash analysis and thermoreflectance methods in 2019.…”

Section: Iiif Fuel Thermal Conductivitymentioning

confidence: 99%

Integrating Advanced Modeling and Accelerated Testing for a Modernized Fuel Qualification Paradigm

et al. 2021

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With the increasing interest in sodium fast reactor technology, as seen by applications to the U.S. Nuclear Regulatory Commission for the OKLO Aurora plant, fuel testing for the TerraPower Traveling Wave Reactor, and the impending construction and startup of the versatile test reactor (VTR), a modernized, accelerated approach to fuel qualification is needed. To guide this effort, a Phenomena Identification Ranking Table-styled analysis was performed for a U-Pu-Zr sodium-free annular fuel system. This analysis evaluated a series of fuel design properties and parameters against their contributions to key fuel performance phenomena. The resulting priority parameters were then reviewed against existing modeling and experimental capabilities to support investigation of the highest-priority parameters. A pathway for qualification was then established using highthroughput, high-volume experiments from MiniFuel and FAST in parallel with advanced physics-based model development. This effort outlines how the first stages of qualification can be reduced from the typical 20+-year development cycle to 5 to 7 years by deploying accelerated irradiation testing platforms. As with any accelerated test, these methods are prototypic in some aspects and less so in others; however, by coupling with advanced fuel performance modeling and simulation capabilities, the larger space of irradiation parameters and material response provided offers advantages for the validation of physics-based models supporting the deployment of novel fuel designs. As a test case, this paper utilizes a proposed Mark II fuel system for the upcoming VTR. Thus, an accelerated qualification method can be tested for the development of MARK II driver fuel so that by the time of VTR startup, lead test assemblies for a Mark II fuel can be initiated.

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Scanning electron microscopy examination of a Fast Flux Test Facility irradiated U-10Zr fuel cross section clad with HT-9

Cited by 63 publications

References 21 publications

Thermodynamic stability studies of Ce-Sb compounds with Fe

Thermodynamic stability studies of Ce-Sb compounds with Fe

Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments

Integrating Advanced Modeling and Accelerated Testing for a Modernized Fuel Qualification Paradigm

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