• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.213-218
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• 1996

Characteristics of high density U-Mo alloy powder prepared by centrifugal atomization have been examined. The results indicate that the majority of the atomized U-Mo alloy particles has a smooth surface and frequently near-perfect spheroidal shape with few satellites attached. The size distribution of atomized U-Mo alloy powder shows the mono-modal size distribution seen in ligament disintegration mechanism. All phases of atomized alloy powder below 150$\mu\textrm{m}$ irrespectively to particle size are found to be ${\gamma}$-U (cubic structure) phases with isotropic structure and not to be U$_2$Mo phase at all. The microstructure of atomized U-Mo alloy particulates has micro-crystalline structure with non-dendritic gram supersaturated with Mo element. Also the grain size of ${\gamma}$ -U tends to decrease with the decrease of the powder diameter.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.859-870
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• 2013

The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ${\sim}60%^{235}U$; the mini-rods were irradiated to an average burnup of ${\sim}85%^{235}U$. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ~ 40% up to ~ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ~ 40% up to ~ 85%.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.495-500
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• 2019

When a new fuel is developed, various mechanical properties are absolutely necessary for a safety analysis of the fuel for the licensing and prediction of its mechanical behavior during operation and accident conditions. In this paper, a mechanically equivalent surrogate plate of U-Mo dispersion fuel is presented using tungsten, substitute material of U-Mo particle. A surrogate plate, composed of tungsten/aluminum dispersion meat and aluminum alloy cladding, is manufactured with the same fabrication process with that of fuel plate except that a tungsten powder is used instead of U-Mo powder. A modal test showed that the surrogate plate and fuel plate have similar dynamic characteristics, and a tensile test demonstrated the similarity of the material property up to the yield strength range. The conducted tests proved that the surrogate tungsten plate has equivalent mechanical behaviors with that of a fuel plate, which leads to the acceptable use of a surrogate fuel assembly using tungsten/aluminum dispersion meat in various mechanical tests. The surrogate fuel assembly can be utilized for various out-of-pile characteristic tests, which are necessary for the licensing achievement of a research reactor that uses U-Mo dispersion fuel as a driver.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.147-158
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• 2014

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.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.226-233
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• 2022

In the present work, a phase-field model was developed to investigate the influence of recrystallization on bubble evolution during irradiation. Considering the interaction between bubbles and grain boundary (GB), a set of modified Cahn-Hilliard and Allen-Cahn equations, with field variables and order parameters evolving in space and time, was used in this model. Both the kinetics of recrystallization characterized in experiments and point defects generated during cascade were incorporated in the model. The bubble evolution in recrystallized polycrystalline of U-Mo alloy was also investigated. The simulation results showed that GB with a large area fraction generated by recrystallization accelerates the formation and growth of bubbles. With the formation of new grains, gas atoms are swept and collected by GBs. The simulation results of bubble size and distribution are consistent with the experimental results.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.228-228
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• 1998

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.645-650
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• 2017

This technical note demonstrates the feasibility of using laser ablation inductively coupled plasma mass spectrometry for the characterization of U-7Mo/Ale5Si dispersion fuel. Our measurements show 5.0% Relative Standard Deviation (RSD) for the reproducibility of measured $^{98}Mo/^{238}U$ ratios in fuel particles from spot analysis, and 3.4% RSD for $^{98}Mo/^{238}U$ ratios in a NIST-SRM 612 glass standard. Line scanning allows for the distinction of U-7Mo fuel particles from the Al-5Si matrix. Each mass spectrum peak indicates the presence of U-7Mo fuel particles, and the time width of each peak corresponds to the size of that fuel particle. The size of the fuel particles is estimated from the time width of the mass spectrum peak for $^{98}Mo$ by considering the scan rate used during the line scan. This preliminary application clearly demonstrates that laser ablation inductively coupled plasma mass spectrometry can directly identify isotope ratios and sizes of the fuel particles in U-Mo/Al dispersion fuel. Once optimized further, this instrument will be a powerful tool for investigating irradiated dispersion fuels in terms of fission product distributions in fuel matrices, and the changes in fuel particle size or shape after irradiation.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.827-838
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• 2013

Interaction layer growth between U-Mo alloy fuel particles and Al in a dispersion fuel is a concern due to the volume expansion and other unfavorable irradiation behavior of the interaction product. To reduce interaction layer (IL) growth, a small amount of Si is added to the Al. As a result, IL growth is affected by the Si content in the Al matrix. In order to predict IL growth during fabrication and irradiation, empirical models were developed. For IL growth prediction during fabrication and any follow-on heating process before irradiation, out-of-pile heating test data were used to develop kinetic correlations. Two out-of-pile correlations, one for the pure Al matrix and the other for the Al matrix with Si addition, respectively, were developed, which are Arrhenius equations that include temperature and time. For IL growth predictions during irradiation, the out-of-pile correlations were modified to include a fission-rate term to consider fission enhanced diffusion, and multiplication factors to incorporate the Si addition effect and the effect of the Mo content. The in-pile correlation is applicable for a pure Al matrix and an Al matrix with the Si content up to 8 wt%, for fuel temperatures up to $200^{\circ}C$, and for Mo content in the range of 6 - 10wt%. In order to cover these ranges, in-pile data were included in modeling from various tests, such as the US RERTR-4, -5, -6, -7 and -9 tests and Korea's KOMO-4 test, that were designed to systematically examine the effects of the fission rate, temperature, Si content in Al matrix, and Mo content in U-Mo particles. A model converting the IL thickness to the IL volume fraction in the meat was also developed.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.169-182
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• 2014

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. U-Mo 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.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09a
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• pp.587-588
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• 2006

The stress corrosion cracking (SCC) susceptibility of Alloy 600 MA, Alloy 600 TT, Alloy 800, and Alloy 690 TT were investigated in a deaerated 0.01 M solution of sodium tetrathionate using reverse u-bend test samples at $340^{\circ}C$. The results showed that SCC occurred in all alloys, excluding Alloy 690 TT. The SCC susceptibility decreased with an increase in the chromium content of the alloys. The results of the deposits and spectra taken from an energy dispersive X-ray system confirmed the existence of a reduced sulfur causing SCC.