Jan-Fong Jue

Irradiation performance of U-Mo monolithic fuel

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

Abstract The Reduced Enrichment for Research and Test Reactors (RERTR) program develops advanced nuclear fuels for high-power test reactors. Monolithic fuel design provides higher uranium loading than that of the traditional dispersion fuel design. Hot isostatic pressing is a promising process for low-cost batch fabrication of monolithic RERTR fuel plates for these high-power reactors. Bonding U-Mo fuel foil and 6061-Al cladding by hot isostatic press bonding was successfully developed at Idaho National Laboratory. Because of the relatively high processing temperature, the interaction between fuel meat and aluminum cladding is a concern. Two different methods were employed to mitigate this effect: a diffusion barrier and a doping addition to the interface. Both types of fuel plates have been fabricated by hot isostatic press bonding. Preliminary results show that the direct fuel/cladding interaction during the bonding process was eliminated by introducing a thin zirconium diffusion barrier layer between the fuel and the cladding. Fuel plates were also produced and characterized with a silicon-rich interlayer between fuel and cladding. This paper reports the recent progress of this developmental effort and identifies the areas that need further attention.

SCANNING ELECTRON MICROSCOPY ANALYSIS OF FUEL/MATRIX INTERACTION LAYERS IN HIGHLY-IRRADIATED U-Mo DISPERSION FUEL PLATES WITH Al AND Al–Si ALLOY MATRICES

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U?7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U?7Mo dispersion fuel elements with pure Al, Al?2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U?7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U?7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al?Si matrices.

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements.

Microstructural Characterization of the Irradiated Nuclear Fuels

The microstructural characterization using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) for the irradiated fuels played an important role to the understanding of fuel performance. Significant progresses have been made in recent years on SEM and TEM work for fuel development in reduced-enrichment for research and test reactors (RERTR) program [1, 2]. It is extremely challenging to prepare the samples from the highly radioactive irradiated fuel for high resolution microscopy analysis. For the complex microstructure of irradiated fuels, the traditional mechanical polishing tends to produce a smeared and disturbed surface making it difficult to reveal the original microstructure in SEM while the traditional TEM sample preparation often limits the ability to access the areas of interest for detailed analysis. The new development using the focused-ion-beam (FIB) lift-out and polishing technical at the Idaho National Laboratory (INL) demonstrated the great advantage in microstructural characterization for the irradiated nuclear fuels.