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

Abstract

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

Thermochemically-Informed Mass Transport Model for Zr in U-Zr Fuel

“…2 is a resulting computed phase diagram. Also indicated on the diagram is the typical fuel alloy composition, U-10wt% Zr, or U-22at% Zr, that was utilized in our simulations as well as in reported experimental irradiations in both EBR-II [16], and most recently included in rodlets irradiated in the Advanced Test Reactor (ATR) [14]. While in our previous work involving oxide fuels, the UO 2±x fluorite phase is the dominant phase in all regions of the fuel element throughout the simulation, the case for metallic fuel is not as simple.…”

Developmental Objectives for Advanced Reactor Fuels

“…For these reasons, the pursuit of sodium-bonded solid fuel designs with low smear density (e.g., less than 75 percent) will be discontinued. 12 Compared with the U-10Zr alloy, the fuel pin with U-10Mo shows large amounts of fuel/cladding chemical interaction and possibly centerline melting of the fuel. The U-10Zr pin may have avoided melting because the migration of zirconium towards the pin center (and the corresponding migration of uranium away from the pin center) has the effect of increasing the solidus temperature near the center of the pin and moving power towards the fuel/cladding interface.…”

“…Based on these results, the testing of U-Mo alloys for fast reactor applications will be continued only after evaluating designs that employ liners. ... 12 Ironically, the pin with the palladium appears to have more fuel/cladding chemical interaction than the standard U-10Zr alloy. Preliminary work 14 indicates that most of the zirconium and palladium migrated to the center of the pin.…”

“…Figure 8 shows neutron radiographs of low smear density (55 percent) pins taken to about 2-4 at.% burnup. 12 Note that the sodium-bonded, solid fuel designs exhibited liftoff, extensive axial deformation, and ended irradiation with unusual, asymmetric distributions of fuel within the cladding; in contrast, the annular fuel design shows little evidence of axial extension of the fuel and the distribution of fuel within the pin appears to be relatively uniform. The features associated with the low smear density, solid fuel designs will result in substantial increases in uncertainties from the reactor physics point of view, and the uneven distribution of fuel raises concerns about fuel slumping that will be difficult to address.…”

“…Reduced smear density (55 percent) annular fuel pins that used either U-10Zr, U-4Pd-10Zr, or U-10Mo were tested as part of the AFC-3A and AFC-3B experiments 12 (note, however, that only low burnups results are currently available). In the case of the U-10Zr alloys, the overall performance of the fuel elements appears to be satisfactory, although there are some concerns associated with increased porosity and FCCI at the fuel/cladding interface in these experiments (see Figure 9).…”