Uranium-molybdenum Alloy Research Articles

Characterization of γ-U–10 wt.%Mo alloy powders obtained by hydride-milling-dehydride process

Uranium–molybdenum alloy powders are raw material of uranium alloy fuels prepared by powder metallurgic process in nuclear reactor. U–10 wt.%Mo alloy powders were prepared directly by hydride-milling-dehydride process without γ phase to α phase transformation under long time heat treatment, the powders were studied by XRD and SEM. During absorb–desorb hydrogen cycle, the absorbed hydrogen quantity of U–10 wt.%Mo alloy increased with the cycle number increasing, after eight cycles, the absorbed hydrogen quantity reached a stable value, and absorbed hydrogen was saturated in alloy. U–10 wt.%Mo alloy hydride was milled in argon atmosphere, then, alloy hydride were heated to desorb the absorbed hydrogen at 600 °C, after these processes, U–10 wt.%Mo alloy powders were obtained. XRD results show that U–10 wt.%Mo alloy powders was still γ phase, its indicated that during preparation, the alloy was always body centered cubic structure, phase transformation process did not exist. SEM observation showed that the particle shape of U–10 wt.%Mo alloy powders was nonuniform, but particle size was less than 50 μm, some were below 10 μm.

Microstructural Characterization of U-Nb-Zr, U-Mo-Nb, and U-Mo-Ti Alloys via Electron Microscopy

Ternary uranium molybdenum alloys are being examined for use as dispersion and monolithic nuclear fuels in research and test reactors. In this study, three such ternary alloys, with compositions U-10Nb-4Zr, U-8Mo-3Nb, and U-7Mo-3Ti in wt.%, were examined using scanning electron microscopy (SEM), x-ray diffraction (XRD), and transmission electron microscopy (TEM) with high angle annular dark field (HAADF) imaging via scanning transmission electron microscopy (STEM). These alloys were homogenized at 950 °C for 96 h and were expected to be single-phase bcc-γ-U. However, upon examination, it was determined that despite homogenization, each of the alloys contained a small volume fraction precipitate phase. Through SEM and XRD, it was confirmed that the matrix retained the bcc-γ-U phase. TEM specimens were prepared using site-specific focused ion beam (FIB) in situ lift out (INLO) technique to include at least one precipitate from each alloy. By electron diffraction, the precipitate phases for the U-10Nb-4Zr, U-8Mo-3Nb, and U-7Mo-3Ti alloys were identified as bcc-(Nb,Zr), bcc-(Mo,Nb), and bcc-(Mo,Ti) solid solutions, respectively. The composition and phase information collected in this study was then used to construct ternary isotherms for each of these alloys at 950 °C.

Mechanical Properties of DU-xMo Alloys with x = 7 to 12 Weight Percent

Mechanical properties of six depleted uranium-molybdenum (U-Mo) alloys have been obtained using microhardness, quasistatic tensile tests, and scanning electron microscopy (SEM) failure analysis. U-Mo alloy foils are currently under investigation for potential conversion of high power research reactors to low enriched uranium fuel. Although mechanical properties take on a secondary effect during irradiation, an understanding of the alloy behavior during fabrication and the effects of irradiation on the integrity of the fuel is essential. In general, the microhardness, yield strength, Young’s modulus, and ultimate tensile strength improved with increasing Mo content. Microhardness measurements were very sensitive to local composition, while the failure mode was significantly controlled by the impurity concentration of the alloy, especially carbon. Values obtained from literature are also provided with reasonable agreement, even though processing conditions and applications were quite different in some instances.

Microstructural characteristics of DU– xMo alloys with x = 7–12 wt%

Microstructural, phase, and impurity analyses of six depleted uranium–molybdenum alloys were obtained using optical metallography, X-ray diffraction, and carbon/nitrogen/oxygen determination. Uranium–molybdenum alloy foils are currently under investigation for the conversion of high-power research reactors using high-enriched uranium fuel to accommodate the use of low-enriched uranium fuel. Understanding basic microstructural behavior of these foils is an important consideration in determining the impact of fabrication processes and in anticipating performance of the foils in a reactor. Average grain diameter decreased with increasing molybdenum content. Lattice parameter decreased with increasing molybdenum content, and no significant degree of phase decomposition or crystallographic ordering was caused by processing and post-processing conditions employed in this study. Impurity concentration, specifically carbon, inhibited the degree of microstructural recrystallization but did not appear to impact other microstructural traits, such as γ-phase retention or lattice parameter.

On dislocation mechanism of dynamic deformation of uranium

In this work, based on results of the tests with study of dynamic diagrams of compression of uranium–molybdenum alloy, we made an attempt to determine dislocation velocity, length of free run of dislocations, and increase of dislocation density during plastic deformation of alloy at dynamic strain rates.

Hydrogen Absorption-Desorption and Gamma-UMo Nuclear Fuel Powder Production

Gamma uranium-molybdenum alloys has been considered as the fuel phase in plate type fuel elements for MTR reactors due to its performance under irradiation and metallurgical processing. To its usage as dispersion phase in aluminum matrix, a necessary step is the conversion of the as cast structure into powder, and the technique considered at IPEN / CNEN - Brazil was HDH (hydration-dehydration). This work has the aim to study the hydrogen incorporation by gamma-UMo alloys with 8% weight molybdenum. The samples were thermally treated under constant flow of hydrogen, for temperatures varying from 500oC up to 600oC and times of 1 to 4 hours. Some of the curves relating mass incorporation and time for the above temperatures were obtained, and the results related to its microstructures and ease of fragmentation.

Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R&D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U–Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U–Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

Thermophysical Properties of U–Mo/Al Alloy Dispersion Fuel Meats

Uranium–molybdenum alloy dispersion fuel meats are being studied for utilization as a research reactor fuel. Thermophysical properties of U–Mo/Al dispersion fuel, where U–Mo was dispersed in aluminum in research reactor fuel for the study, were determined by computing the thermal conductivity through measurements of the specific heat capacity and thermal diffusivity. Uranium molybdenum powder was first fabricated and utilized as U–Mo/Al dispersion fuel; the molybdenum-to-uranium ratios were 6, 8, and 10 mass% to produce the initial powder, which was then combined with aluminum (Al 1060). The volume fractions of U–Mo powder to aluminum were 10, 30, 40, and 50 vol.% to fabricate the dispersion fuel. The thermal diffusivity and specific heat capacity were measured by the laser-flash and differential scanning calorimetry (DSC) methods, respectively. Although the thermal diffusivity showed a decreasing trend with the U–Mo volume fraction when the dispersion quantity was insignificant, the trend reversed with a higher dispersion level. The specific heat capacity increases monotonically with temperature; its value is larger for a smaller dispersion level. Additionally, the overall thermal conductivity increases with temperature. Finally, the thermal conductivity decreases with an increase in the amount of U–Mo powder in the dispersion fuel.

Irradiation-enhanced interdiffusion in the diffusion zone of U-Mo dispersion fuel in Al

Uranium-molybdenum (U-Mo) alloy fuel particles dispersed in an aluminum (Al) matrix, designated as U-Mo/Al dispersion fuel, is in the development stage in the worldwide RERTR (Reduced Enrichment for Research and Test Reactors) program. The main issue in developing U-Mo/Al dispersion fuel is the diffusion reaction occurring at the interface between the fuel particles and matrix. To accurately analyze fuel performance, a model to predict the diffusion kinetics is necessary. For this purpose, the authors developed a diffusion layer growth rate correlation for out-of-pile annealing tests and a similar correlation for in-reactor tests. The correlation for in-reactor tests is considerably different from that of out-of-pile tests because it contains factors that amplify diffusion kinetics by fission damage in the diffusion reaction zone. This irradiation enhancement was formulated by a combination of the fission rate in the fuel and fission fragment damage distribution in the diffusion reaction zone. Using a computer code, fission damage factors were obtained as a function of diffusion reaction layer thickness and composition. The model correlation was established and fitted to the in-reactor data. As a result of this data fitting, the interaction layer growth rate is found to be proportional to the square root of the fission fragment damage rate and to have a temperature dependence characterized by the effective activation energy of 46 to 76 kJ/mole, which is smaller by a factor of 4 to 7 than that of out-of-pile tests.

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium—molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium—molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used.

Use of ion beams to simulate reaction of reactor fuels with their cladding

Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium–molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium–molybdenum coated with a 100nm thin film of aluminum was irradiated with 3MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27Al(p,γ)28Si reaction which occurs at a proton energy of 991.9keV. During irradiation at 150°C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30nm2/dpa. At a dose of 90dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery.

Variants of Aperiodic Pulsed Reactors with Boosted Pulse Parameters

Single (one-section) reactors, containing a neptunium–gallium or uranium–molybdenum alloy core, and coupled two-section systems, consisting of a single pulsed reactor and a driven subcritical assembly, consisting of a uranium–molybdenum alloy or dispersed uranium–graphite material, are studied. Calculations of the neutron and dynamical characteristics of these systems are performed. The information obtained, together with data from previous calculations, made it possible to draw a conclusion about the structure of reactor systems characterized by a short radiation pulse and large-volume cores and cavities for holding specimens. It is shown that a reactor with a single neptunium–gallium alloy core and a coupled cascade-type system consisting of this single driven reactor together with a driven subcritical uranium–graphite assembly have the best pulse parameters.

Production of uranium–molybdenum particles by spark-erosion

With the spark-erosion method we have produced spheroidal particles of an uranium–molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO 2 with a distinctive size distribution with peaks centered at 70 and 10 μm were obtained. The particles have central inclusions of U and Mo compounds.

Irradiation-Induced Recrystallization of Cellular Dislocation Networks in Uranium-Molybdenum Alloys

ABSTRACTWe developed a rate-theory-based model to investigate the nucleation and growth of interstitial loops and cavities during low-temperature in-reactor irradiation of uranium-molybdenum alloys. Consolidation of the dislocation structure takes into account the generation of forest dislocations and capture of interstitial dislocation loops. The theoretical description includes stress-induced glide of dislocation loops and accumulation of dislocations on cell walls. The loops accumulate and ultimately evolve into a low-energy cellular dislocation structure. Calculations indicate that nanometer-size bubbles are associated with the walls of the cellular dislocation structure. The accumulation of interstitial loops within the cells and of dislocations on the cell walls leads to increasing values for the rotation (misfit) of the cell wall into a subgrain boundary and a change in the lattice parameter as a function of dose. Subsequently, increasing values for the stored energy in the material are shown to be sufficient for the material to undergo recrystallization. Results of the calculations are compared with SEM photomicrographs of irradiated U- 10Mo, as well as with data from irradiated UO2.

Effect of Formation and Growth of Dislocation Loops and Cavities on Low-Temperature Swelling of Irradiated Uranium-Molybdenum Alloys

AbstractScanning electron photomicrographs of U–10 wt.% Mo irradiated at low temperature in the Advanced Test Reactor (ATR) to about 40 at.% burnup show the presence of cavities. We have used a rate-theory-based model to investigate the nucleation and growth of cavities during low-temperature irradiation of uranium-molybdenum alloys in the presence of irradiation-induced interstitial-loop formation and growth. Our calculations indicate that the swelling mechanism in the U–10 wt.% Mo alloy at low irradiation temperatures is fission-gas driven. The calculations also indicate that the observed bubbles must be associated with a subgrain structure. Calculated bubble-size-distributions are compared with irradiation data.

Fabrication of Uranium-Molybdenum Alloy Fuels for Pulsed Reactors

Fabrication of Uranium-Molybdenum Alloy Fuels for Pulsed Reactors

Isotope Effect in Superconductingγ−UraniumAlloys

A new determination of the isotope effect in $\ensuremath{\gamma}$-stabilized uranium-molybdenum alloys has been made. We find ${T}_{c}\ensuremath{\propto}{M}^{\ensuremath{-}0.40}$, rather than the previously reported ${T}_{c}\ensuremath{\propto}{M}^{\ensuremath{-}0.53}$.

Dynamic Neutronic and Mechanical Measurements with a Pulse Reactor of Uranium-Molybdenum Alloy

The time-dependent behavior of the neutron population in an unreflected, unmoderated cylindrical assembly of 90 wt% uranium (93.2 wt% 235U), 10 wt% molybdenum alloy following rapid establishment of a superprompt criticality with negligible initial neutron population has been studied. Reactivity increases up to 14¢ above prompt criticality resulted in pulses yielding as many as 3.72 × 1017 fissions with reactor periods as short as 12.4 µsec and temperature increases as large as 880°C. In these experiments the reactor produced pulses of 2 × 1017 fissions without any damage. A pulse of 2.37 × 1017 fissions resulted in permanent elongation of the bolts holding the core together, and a pulse of 2.66 × 1017 fissions caused cracks in some of the core parts. Stresses obtained from measurements of the mechanical vibration of the reactor parts were consistent with the observed damage.

A study of the shear transformations from the gamma-phase in uranium-molybdenum alloys containing 6.0–12.5 at % molybdenum

A dilatometric investigation has been made of the shear transformations from the γ phase of uranium-molybdenum alloys in the composition range 6.0 to 12.5 at % molybdenum. It has been found that there are two shear transformations occurring successively over most of this composition range. The one at the higher temperature is a transformation from the γ 0 to γ 0 d phase (a tetragonal variant of the γ phase) with c/a < 0.5, while that at the lower temperature is from γ 0 d with c/a < 0.5 to α (a monoclinic variant of the α-phase of uranium). The enthalpies of these two transformations have also been measured as a function of composition and cooling rate. It is found that the enthalpy depends on the magnitude of the martensitic shear which is in turn controlled by the cooling rate because of the effect this has on the degree of order.

Superconductivity ofα-Phase U-Mo Alloys at Zero and High Pressures

Heat-capacity measurements and determinations of the superconducting transition temperature as a function of applied pressure up to 10 kbar have been made on a number of $\ensuremath{\alpha}$-phase uranium-molybdenum alloys. Small additions of molybdenum rapidly change the superconducting properties of the $\ensuremath{\alpha}$ phase. The strong initial pressure dependence of ${T}_{c}$ for pure $\ensuremath{\alpha}$-U is destroyed, and bulk superconducting behavior at zero pressure develops.