• Analytical Science and Technology
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• v.13 no.3
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• pp.291-296
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• 2000

The SIMFUEL which composition is similar to PWR nuclear spent fuels was dissolved with a acid digestion bomb. An analytical conditions of ICP-AES for the direct determination of molybdenum in the uranium matrices without separation process were investigated. Based on the effect of uranium on molybdenum intensity. the most optimum wavelengths of molybdenum were found to be 202.030 and 203.844 nm. However, the method of standard additions is applied to overcome the effects of changing background caused by analyzing the sample solutions containing high concentration of uranium and the standard calibration solutions. The relative error of two methods, direct and indirect measurements with cation exchange resin separation procedures, was less than 5%. Therefore it was possible for this procedure to directly measure molybdenum in uranium matrices without separation. And this method was also applied to the determination of several percent of molybdenum in a U-Mo alloy.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.608-609
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• 2018

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1855-1862
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• 2023

Graphite crucibles are used for melting uranium and its alloys in VIM furnace. Various coating materials namely Al₂O₃, ZrO₂, MgO etc. are applied on the inner surface of the crucibles using paint brush or thermal spray technique to mitigate U-C interaction. These leads to significant amount of carbon pick-up in uranium. In this study, the attempts are made to develop multilayer coatings comprising of SiC/Y₂O₃ and Mo/Y₂O₃ on graphite to study the feasibility of minimizing U-C interaction. The parameters are optimized to prepare SiC coating of about 70㎛ thickness using CVD technique on graphite coupons and subsequently Y₂O₃ coating of about 250㎛ thickness using plasma spray technique. Molybdenum and Y₂O₃ layers were deposited using plasma spray technique with 70㎛ and 250㎛ thickness, respectively. Interaction studies of the coated graphite with molten uranium at 1450℃ for 20 min revealed that Y₂O₃ coating with SiC interlayer provides physical barrier for uranium-graphite interaction, however, this led to the physical separation of coating layer. Y₂O₃ coating with Mo interlayer provided superior barrier effect showing no degradation and the coatings remained intact after interaction tests. Therefore, the Mo/Y₂O₃ coating was found to be a promising solution for minimizing carbon pick-up during uranium/uranium alloy melting.

• 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.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.839-846
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• 2013

This paper describes a finite element model of the microstructure of dispersion type nuclear fuels, which can be used to determine the effective thermal conductivity of the fuels during irradiation. The model simulates a representative region of the fuel as a prism shaped unit cell made of brick elements. The elements within the unit cell are assigned material properties of either the fuel or the matrix depending on position, in such a way as to represent randomly distributed fuel particles with a size distribution similar to that of the as manufactured fuel. By applying an appropriate heat flux across the unit cell it is possible to determine the effective thermal conductivity of the unit cell as a function of the volume fraction of the fuel particles. The presence of a fuel/matrix interaction layer is simulated by the addition of a third set of material properties that are assigned to the finite elements that surround each fuel particle. In this way the effective thermal conductivity of the material may also be determined as a function of the volume fraction of the interaction layer. Work is on going to add fission gas bubbles in the fuel as a fourth phase to the model.

• 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.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.