Relevant bibliographies by topics / Uranium-molybdenum alloy

Academic literature on the topic 'Uranium-molybdenum alloy'

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Published: 10 December 2022
Last updated: 28 January 2023

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Journal articles on the topic "Uranium-molybdenum alloy"

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Visser, A. E., R. A. Pierce, and J. E. Laurinat. "Purification of Uranium from a Uranium/Molybdenum Alloy." Separation Science and Technology 43, no. 9-10 (July 18, 2008): 2775–85. http://dx.doi.org/10.1080/01496390802121701.

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Chmela, T., and P. Krupička. "The oxidation kinetics of depleted uranium and its low-alloy molybdenum alloys in moist air." Koroze a ochrana materialu 63, no. 3 (November 1, 2019): 100–104. http://dx.doi.org/10.2478/kom-2019-0013.

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Abstract The oxidation kinetics of depleted uranium and its low-alloy molybdenum alloys (U-2wt.%Mo, U-5wt.%Mo) were measured in a moist air (75% relative humidity) at 60 and 75 ° C. Coefficients of reaction rate equations were determined for linear oxidation kinetics. In the oxidation of depleted uranium at 75 ° C, a change in reaction kinetics from linear to exponential behaviour was observed after about 2500 hours.
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Meyer, M. K., G. L. Hofman, S. L. Hayes, C. R. Clark, T. C. Wiencek, J. L. Snelgrove, R. V. Strain, and K. H. Kim. "Low-temperature irradiation behavior of uranium–molybdenum alloy dispersion fuel." Journal of Nuclear Materials 304, no. 2-3 (August 2002): 221–36. http://dx.doi.org/10.1016/s0022-3115(02)00850-4.

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Orlov, V. K., V. M. Teplinskaya, and N. T. Chebotarev. "Decomposition of a metastable solid solution in uranium-molybdenum alloy." Atomic Energy 88, no. 1 (January 2000): 42–47. http://dx.doi.org/10.1007/bf02673318.

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Adamska, A. M., R. Springell, and T. B. Scott. "Characterization of poly- and single-crystal uranium–molybdenum alloy thin films." Thin Solid Films 550 (January 2014): 319–25. http://dx.doi.org/10.1016/j.tsf.2013.11.087.

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Kautz, Elizabeth J., Sten V. Lambeets, Jacqueline Royer, Daniel E. Perea, Sivanandan S. Harilal, and Arun Devaraj. "Compositional partitioning during early stages of oxidation of a uranium-molybdenum alloy." Scripta Materialia 212 (April 2022): 114528. http://dx.doi.org/10.1016/j.scriptamat.2022.114528.

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Camarano, D. M., F. A. Mansur, A. M. M. Santos, W. B. Ferraz, and T. A. Pedrosa. "Effects of heat treatments on the thermal diffusivity of Uranium-Molybdenum alloy." Journal of Physics: Conference Series 733 (July 2016): 012014. http://dx.doi.org/10.1088/1742-6596/733/1/012014.

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Naymushin, Artem G., Yuri B. Chertkov, Vasily V. Kurganov, Ivan I. Lebedev, Svetlana A. Mongush, and Natalya V. Daneikina. "Feasibility Study of Using New Fuel Composition in IRT-T Research Reactor." Advanced Materials Research 1084 (January 2015): 306–8. http://dx.doi.org/10.4028/www.scientific.net/amr.1084.306.

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The results of simulation of IRT-T reactor conversion from highly enriched fuel to new perspective low enriched fuel based on uranium-molybdenum alloy are given. Main characteristics of reacting core with the use of highly enriched and low enriched fuel are calculated. It is shown that impact of using new materials in fuel composition remains neutronic and thermal hydraulic characteristics of the core at an acceptable level.
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Karpyuk, L. A., A. M. Savchenko, Yu V. Konovalov, G. A. Kulakov, S. V. Maranchak, S. A. Ershov, E. V. Maynikov, et al. "Features of the behavior of the dispersion fuel METMET under irradiation." Voprosy Materialovedeniya, no. 3(111) (November 1, 2022): 148–55. http://dx.doi.org/10.22349/1994-6716-2022-111-3-148-155.

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The paper considers the behavior under irradiation of the METMET fuel composition, which consists of particles of uranium-molybdenum alloy in a matrix of zirconium alloys. Post-reactor investigations confirmed the satisfactory performance of pilot fuel elements irradiated in the MIR reactor to a burnup of 61 MW day/kgU under significant thermal loads. The structural stability of the fuel under irradiation, good compatibility of the fuel rod components with each other could be noted. Fuel rods with METMET fuel composition have good prospects for use in reactors of floating nuclear power units and low-capacity nuclear plants, as well as a tolerant fuel.
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Kolotova, Lada, and Ilia Gordeev. "Structure and Phase Transition Features of Monoclinic and Tetragonal Phases in U–Mo Alloys." Crystals 10, no. 6 (June 16, 2020): 515. http://dx.doi.org/10.3390/cryst10060515.

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Using molecular dynamics simulations, we studied the structural properties of orthorhombic, monoclinic, and body-centered tetragonal (bct) phases of U–Mo alloys. A sequence of shear transformations between metastable phases takes place upon doping of uranium with molybdenum from pure α -U: orthorhombic α ′ → monoclinic α ″ → bct γ 0 → body-centered cubic (bcc) with doubled lattice constant γ s → bcc γ . The effects of alloy content on the structure of these phases have been investigated. It has been shown that increase in molybdenum concentration leads to an increase in the monoclinic angle and is more similar to the γ 0 -phase. In turn, tetragonal distortion of the γ 0 -phase lattice with displacement of a central atom in the basic cell along the <001> direction makes it more like the α ″ -phase. Both of these effects reduce the necessary shift in atomic positions for the α ″ → γ 0 -phase transition.
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Dissertations / Theses on the topic "Uranium-molybdenum alloy"

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Gardner, Levi D. "Low-Temperature Aqueous Corrosion Behavior of Uranium Molybdenum Alloys." DigitalCommons@USU, 2016. https://digitalcommons.usu.edu/etd/4755.

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Nuclear fuel characterization requires understanding of the various conditions to which materials are exposed in-reactor. One of these important conditions is corrosion, particularly that of fuel constituents. Therefore, corrosion behavior is of special interest and an essential part of nuclear materials characterization efforts. In support of the Office of Material Management and Minimization’s Reactor Conversion Program, monolithic uranium-10 wt% molybdenum alloy (UMo) is being investigated as a low enriched uranium alternative to highly enriched uranium dispersion fuel currently used in domestic high performance research reactors. The aqueous corrosion behavior of U-Mo is being examined at Pacific Northwest National Laboratory (PNNL) as part of U-Mo fuel fabrication capability activity. No prior study adequately represents this behavior given the current state of alloy composition and thermomechanical processing methods, and research reactor water chemistry. Two main measurement techniques were employed to evaluate U-Mo corrosion behavior. Low-temperature corrosion rate values were determined by means of U-Mo immersion testing and subsequent mass-loss measurements. The electrochemical behavior of each processing condition was also qualitatively examined using the techniques of corrosion potential and anodic potentiodynamic polarization. Scanning electron microscopy (SEM) and optical metallography (OM) imagery and hardness measurements provided supplemental corrosion analysis in an effort to relate material corrosion behavior to processing. The processing effects investigated as part of this were those of homogenization heat treatment (employed to mitigate the effects of coring in castings) and sub-eutectoid heat treatment, meant to represent additional steps in fabrication (such as hot isostatic pressing) performed at similar temperatures. Immersion mass loss measurements and electrochemical results both showed very little appreciable difference between specimens of different process parameters. Comparative results were presented as linear corrosion rates and temperature-dependent Arrhenius equations, which were then correlated with electrochemical and metallographic findings for each condition under investigation. This thesis was prepared in the monograph style using the ASME reference format.
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OLIVEIRA, FABIO B. V. de. "Desenvolvimento de um combustível de alta densidade à base da liga urânio-molibdênio com alta compatibilidade em altas temperaturas." reponame:Repositório Institucional do IPEN, 2008. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11621.

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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
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Perez, Emmanuel. "Interdiffusion behavior of U-Mo alloys in contact with Al and Al-Si alloys." Doctoral diss., University of Central Florida, 2011. http://digital.library.ucf.edu/cdm/ref/collection/ETD/id/5007.

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U-Mo dispersion and monolithic fuels embedded in Al-alloy matrix are under development to fulfill the requirements of research reactors to use low-enriched molybdenum stabilized uranium alloys as fuels. The system under consideration in this study consisted of body centered cubic gamma] U-Mo alloys embedded in an Al structural matrix. Significant interaction has been observed to take place between the U-Mo fuel and the Al matrix during manufacturing of the fuel-plate system assembly and during irradiation in reactors. These interactions produce Al-rich phases with physical and thermal properties that adversely affect the performance of the fuel system and can lead to premature failure. In this study, interdiffusion and microstructural development in the U-Mo vs. Al system was examined using solid-to-solid diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo and U-12wt.%Mo vs. pure Al, annealed at 600[degrees]C for 24 hours. The influence of Si alloying addition (up to 5 wt.%) in Al on the interdiffusion microstructural development was also examined using solid-to-solid diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo and U-12wt.%Mo vs. pure Al, Al-2wt.%Si, and Al-5wt.%Si annealed at 550??C for 1, 5 and 20 hours. To further clarify the diffusional behavior in the U-Mo-Al and U-Mo-Al-Si systems, Al-rich 85.7Al-11.44U-2.86Mo, 87.5Al-10U-2.5Mo, 56.1Al-18.9Si-21.9U-3.1Mo and 69.3Al-11.9Si-18.8U (at.%) alloys were cast and homogenized at 500[degrees]C to determine the equilibrium phases of the system. Scanning electron microscopy (SEM), transmission electron microscopy (TEM) and electron probe microanalysis (EPMA) and X-ray diffraction (XRD) were employed to examine the phase development in the diffusion couples and the cast alloys. In ternary U-Mo-Al diffusion couples annealed at 600[degrees]C for 24 hours, the interdiffusion microstructure consisted of finely dispersed UAl[sub3], UAl[sub4], U[sub6]Mo[sub2]Al[sub20], and UMo[sub2]Al[sub20] phases while the average composition throughout the interdiffusion zone remained constant at approximately 80 at.% Al. The interdiffusion microstructures observed by EPMA, SEM and TEM analyses were correlated to explain the observed morphological development in the interdiffusion zones. The concept of thermodynamic degrees of freedom was used to justify that, although deviations are apparent, the interdiffusion zones did not significantly deviate from an equilibrium condition in order for the observed microstructures to develop. Selected diffusion couples developed periodic bands within the interdiffusion zone as sub-layers in the three-phase regions. Observation of periodic banding was utilized to augment the hypothesis that internal stresses play a significant role in the phase development and evolution of U-Mo vs. pure Al diffusion couples. The addition of Si (up to 5 wt.%) to the Al significantly reduced the growth rate of the interdiffusion zone. The constituent phases and composition within the interdiffusion zone were also modified. When Si was present in the Al terminal alloys, the interdiffusion zones developed layered morphologies with fine distributions of the (U,Mo)(Al,Si)[sub3] and UMo[sub2]Al[sub20] phases. The U[sub6]Mo[sub4]Al[sub43] phase was observed scarcely in Si depleted regions within the interdiffusion zone. The phase development and evolution of the interdiffusion zone was described in terms of thermodynamic degrees of freedom with minimal deviations from equilibrium.
ID: 029809410; System requirements: World Wide Web browser and PDF reader.; Mode of access: World Wide Web.; Thesis (Ph.D.)--University of Central Florida, 2011.; Includes bibliographical references (p. 111-115).
Ph.D.
Doctorate
Mechanical, Materials, and Aerospace Engineering
Engineering and Computer Science
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ALMEIDA, CIRILA T. de. "Desempenho sob irradiação de elementos combustíveis do tipo U-Mo." reponame:Repositório Institucional do IPEN, 2005. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11286.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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MUNIZ, RAFAEL O. R. "Análise neutrônica e especificação técnica para o combustível a dispersão UMo-Al com adição de veneno queimável." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/25671.

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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Creasy, John Thomas. "Thermal Properties of Uranium-Molybdenum Alloys: Phase Decomposition Effects of Heat Treatments." Thesis, 2011. http://hdl.handle.net/1969.1/ETD-TAMU-2011-12-10625.

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Uranium-Molybdenum (U-Mo) alloys are of interest to the nuclear engineering community for their potential use as reactor fuel. The addition of molybdenum serves to stabilize the gamma phase of uranium, as well as increasing the melting point of the fuel. Thermal properties of U-Mo alloys have not been fully characterized, especially within the area of partial phase decomposition of the gamma phase of the alloy. Additional data was acquired through this research to expand the characterization data set for U-Mo alloys. The U-Mo alloys used for this research were acquired from the Idaho National Laboratory and consisted of three alloys of nominal 7, 10, and 13 percent molybdenum by weight. The sample pins were formed by vacuum induction melt casting. Once the three sample pins were fabricated and sent to the Fuel Cycle and Materials Laboratory at Texas A&M University, the pins were homogenized and sectioned for heat treatment. Several heat treatments were performed on the samples to induce varying degrees of phase decomposition, and the samples were subsequently sectioned for phase verification and thermal analysis. An Electron Probe Microanalyzer with wavelength dispersive spectroscopy was used to observe the phases in the samples as well as to characterize each phase. The density of each sample was determined using Archimedes method. Finally, a light flash analyzer was used to determine thermal diffusivity of the samples up to 300 degrees C as well as to estimate the thermal conductivity. For U-10Mo, thermal diffusivity increased with increasing phase decomposition from gamma to alpha +U2Mo while U-7Mo saw a flattening of the thermal diffusivity curve with increased phase decomposition.
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Book chapters on the topic "Uranium-molybdenum alloy"

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Pasqualini, Enrique E. "Gamma Uranium Molybdenum Alloy: Its Hydride and Performance." In Nuclear Material Performance. InTech, 2016. http://dx.doi.org/10.5772/63652.

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Conference papers on the topic "Uranium-molybdenum alloy"

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Hollis, K., N. Mara, R. Field, T. Wynn, and P. Dickerson. "Bond Strength Characterization of Plasma Sprayed Zirconium on Uranium Alloy by Microcantilever Testing." In ITSC 2012, edited by R. S. Lima, A. Agarwal, M. M. Hyland, Y. C. Lau, C. J. Li, A. McDonald, and F. L. Toma. ASM International, 2012. http://dx.doi.org/10.31399/asm.cp.itsc2012p0070.

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Abstract The future production of low enriched uranium nuclear fuel for test reactors requires a well-adhered diffusion barrier coating of zirconium (Zr) on the uranium/molybdenum (U-Mo) alloy fuel. In this study, the interfacial bond between plasma sprayed Zr coatings and U-Mo fuel has been characterized for localized bond strength by microcantilever beam testing. Test results have revealed the effect of specific flaws such as cracks and pores on the bond strength of interfaces with a sampling area of approximately 20 µm2. TEM examination has shown the Zr/U-Mo interface to contain rows of very fine grains (5-30 nm) with the Zr in contact with UO2. Bond strengths of plasma sprayed samples have been measured that are similar to those of diffusion-bonded samples showing the potential for plasma sprayed Zr coatings to have high bond strength.
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Hollis, K. J., and M. I. Pena. "Plasma Sprayed and Electrospark Deposited Zirconium Metal Diffusion Barrier Coatings." In ITSC2010, edited by B. R. Marple, A. Agarwal, M. M. Hyland, Y. C. Lau, C. J. Li, R. S. Lima, and G. Montavon. DVS Media GmbH, 2010. http://dx.doi.org/10.31399/asm.cp.itsc2010p0439.

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Abstract Zirconium metal coatings applied by plasma spraying and electrospark deposition (ESD) have been investigated for use as diffusion barrier coatings on low enrichment uranium fuel for research nuclear reactors. The coatings have been applied to both stainless steel as a surrogate and to simulated nuclear fuel uranium-molybdenum alloy substrates. Deposition parameter development accompanied by coating characterization has been performed. The structure of the plasma sprayed coating was shown to vary with transferred arc current during deposition. The structure of ESD coatings was shown to vary with the capacitance of the deposition equipment.
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Hollis, Kendall J., Dustin R. Cummins, Sven C. Vogel, Donald W. Brown, and David E. Dombrowski. "Characterization of Plasma Sprayed Zirconium Coatings on Uranium Alloy Using Neutron Diffraction." In ITSC2018, edited by F. Azarmi, K. Balani, H. Li, T. Eden, K. Shinoda, T. Hussain, F. L. Toma, Y. C. Lau, and J. Veilleux. ASM International, 2018. http://dx.doi.org/10.31399/asm.cp.itsc2018p0299.

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Abstract Plasma sprayed zirconium (Zr) metal coatings onto uranium-molybdenum (U-Mo) alloy nuclear reactor fuel foils act as a diffusion barrier between the fuel and the aluminum fuel cladding. Neutron diffraction was performed to investigate the crystallographic phase composition, crystal orientations and lattice parameters of the plasma sprayed Zr and the U-Mo substrate. The neutron diffraction results show that the plasma sprayed Zr coating is crystalline, phase pure (alpha-Zr) and has preferred crystalline orientation likely due to directional solidification. Also, there is a slight (~0.01 Å for a direction and ~0.016 Å for c direction) increase in the plasma sprayed Zr lattice parameter indicating oxygen in the lattice and some residual thermo-mechanical strain. There is little or no modification of the underlying U-Mo following plasma spraying. In particular, there is no detectable allotropic transformation of the starting gamma-U (body-centered cubic) to alpha-U (orthorhombic). The unique neutron diffraction capabilities at LANL are well suited for nuclear fuel characterization offering distinct advantages over conventional X-ray diffraction and destructive metallography.
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Hawkes, Grant L., Warren F. Jones, Wade Marcum, Aaron Weiss, and Trevor Howard. "Flow Testing and Analysis of the FSP-1 Experiment." In ASME 2017 Nuclear Forum collocated with the ASME 2017 Power Conference Joint With ICOPE-17, the ASME 2017 11th International Conference on Energy Sustainability, and the ASME 2017 15th International Conference on Fuel Cell Science, Engineering and Technology. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/nuclrf2017-3639.

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The U.S. High Performance Research Reactor conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Size Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water channel velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A pressure differential versus flow rate curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported.
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Hawkes, Grant L., and Nicolas E. Woolstenhulme. "Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60752.

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The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.
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Turner, Kyler K., Gary L. Solbrekken, and Charlie W. Allen. "Thermal-Mechanical Analysis of Annular Target Design for High Volume Production of Molybdenum-99 Using Low-Enriched Uranium." In ASME 2009 International Mechanical Engineering Congress and Exposition. ASMEDC, 2009. http://dx.doi.org/10.1115/imece2009-13238.

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Techenetium-99m is a diagnostic radioactive medical isotope that is currently used 30,000 times a day in the United States. All supplies of techenetium-99m’s parent isotope molybdenum-99 currently originate from nuclear reactor facilities located in foreign countries and use highly enriched uranium (HEU). In accordance with the Global Threat Reduction Initiative all uranium used in future molybdenum-99 production will use low enriched uranium (LEU). A design approach to using LEU in a cost-effective manner is to use a target that is based on LEU foil. A potential failure mode for the LEU foil based target is temperature excursion during irradiation due to poor thermal contact between the foil and the target cladding. The purpose of this study is to establish the theoretical basis for experimentally measuring the thermal contact resistance. Replicating in service heating conditions is nearly impossible when testing the thermal contact resistance as part of a study to establish LEU foil warpage tolerances, thus it is necessary to establish an alternate heating configuration that will allow a conservative estimate of the contact resistance. Thermal and mechanical analysis suggests that external heating of an annular target will place the interface into a state that will over-estimate the contact resistance relative to use conditions. Further, the magnitude of the heat load used for testing can be adjusted to control the degree of overestimation.
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Zhong, Haoliang, Yu Si, Xiaofei Chen, Yading Zhang, and Ran Su. "Patent Perspective on the Development of Metal Nuclear Fuel Technology." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-93138.

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Abstract Irradiated growth and irradiated swelling have been the key factors limiting the technological development of metal fuels. Compared with other types of nuclear fuels, metal fuels have the advantages of high thermal conductivity, high fission atomic density and easy processing, so the technological development of metal fuels has been of great interest to researchers, and there has been uninterrupted research and development over the years to overcome the technical problems of irradiation swelling of metal fuels. In this paper, we take a patent perspective and conduct a global patent search on uranium-zirconium alloys, uranium-molybdenum alloys, uranium-plutonium-zirconium alloys and metal fuels of other compositions in the technical fields of composition ratio, preparation process, preparation devices, fuel performance, fuel testing and back-end treatment of fuels, and analyse in depth the global development history of metal fuels, the patent situation of key applicants, the global sources of metal fuel technologies, the The analysis of the legal status, the analysis of key patent technologies, the technical means to solve the problems of irradiated growth and swelling, and the R&D direction of metal fuels at the present stage, and explore the future technological development path of metal fuels. Through in-depth analysis and research, this paper puts forward reasonable suggestions on how to break through the bottleneck of metal fuel technology and the follow-up development of metal fuel related technologies by designing and manufacturing fuel with special structure, improving the composition of metal fuel, modernizing and controlling processing methods, and improving processing technology.
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Bojanowski, C., G. L. Solbrekken, G. H. Schnieders, J. D. Rivers, E. H. Wilson, and L. P. Foyto. "Deflections of Plates in Research Reactor Fuel by Disparities in Thicknesses of Flow Channels." In ASME 2019 Power Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/power2019-1928.

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Abstract Low-enriched uranium (LEU) fuel element designs for the U.S. high performance research reactors LEU conversion cores have been optimized by each reactor facility to allow the reactors to meet mission, operational, and safety basis requirements using monolithic uranium-molybdenum fuel. As a part of work supporting the Preliminary Safety Analysis Report (PSAR) submitted to the NRC by the University of Missouri Research Reactor (MURR), the impact of thinner 1.12 mm LEU curved fuel plates, compared to the currently used highly-enriched uranium (HEU) curved plate thickness of 1.27 mm, has been assessed for hydro-mechanical performance. Plate deflection can be induced by the hydrodynamic pressure differential caused by differences in the thicknesses of surrounding coolant flow channels. An experimental study was conducted on relevant Materials Test Reactor-type (MTR-type) reactor plate geometries in a water flow test loop to validate computational models simulating flow-induced plate deflection. Three-dimensional fluid-structure interaction (FSI) simulations of the experiments were performed using several commercially available multi-physics simulation codes. Inclusion of as-built geometry of the plates and channels in the simulations was key to achieving good agreement with measured deflections. The validated computational methodology was applied to a model of the prototypic MURR LEU plate. For the nominal flow conditions, a small deflection of the plate on the order of 5 micrometers was predicted. That deflection is significantly less than the allowances in the PSAR for change in coolant channel thickness. The experimental model validation of plate deflection is important since conventional figures of merit for the robustness of MTR-type fuel plates under flow, such as the Miller critical velocity, often show a weak correlation with the prediction of stability. Subsequent to this work, irradiation qualification of the MURR LEU fuel element design has begun and will conclude with a full-size demonstration element test.
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Kozlov, E. A. "Comment to a Problem of Two-Wave Configurations Existing in Unalloyed Depleted Uranium and its Alloys with Molybdenum in a Region of 50.5–57.0 GPa." In ZABABAKHIN SCIENTIFIC TALKS - 2005: International Conference on High Energy Density Physics. AIP, 2006. http://dx.doi.org/10.1063/1.2337237.

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Reports on the topic "Uranium-molybdenum alloy"

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Visser, A., and R. Robert Pierce. SOLVENT EXTRACTION FOR URANIUM MOLYBDENUM ALLOY DISSOLUTION FLOWSHEET. Office of Scientific and Technical Information (OSTI), June 2007. http://dx.doi.org/10.2172/908930.

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Bhattacharya, Sumit, Yinbin Miao, Kun Mo, Laura Jamison, Heather Connaway, and Abdellatif Yacout. Temperature Effect over Gas Bubble Evolution in Uranium-10 wt.% Molybdenum alloy Irradiated by Swift Xe Ions. Office of Scientific and Technical Information (OSTI), June 2021. http://dx.doi.org/10.2172/1807665.

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3

Hill, Mary Ann, Samantha Kay Lawrence, and Charles H. Henager, Jr. Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys. Office of Scientific and Technical Information (OSTI), November 2018. http://dx.doi.org/10.2172/1481958.

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4

Hill, Mary Ann, and Samantha Kay Lawrence. Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys. Office of Scientific and Technical Information (OSTI), July 2018. http://dx.doi.org/10.2172/1463514.

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