Uranium Alloys Research Articles

The microstructure and hydriding characteristics of high temperature aged U–13 at.%Nb alloy

Niobium as alloying element significantly improves physical and chemical properties of metallic uranium, exhibiting great application potential in uranium alloy materials. The corrosion resistance performance as well as the internal alloy phase structure of uranium–niobium alloy is closely related to aging processes.Microstructure and hydriding characteristics of the 400°C/9h+500°C/2h aged uranium–13at.% niobium alloys (U–13at.%Nb) were investigated from the point of view of relationship between the microstructure and growth of the hydriding areas. The microstructure, morphology and composition of the alloy phases before and after the hydriding were well characterized by the laser scanning confocal microscopy (LSCM), scanning electron microscopy (SEM) and X-ray diffraction (XRD), respectively. Experimental results indicated that the hydrogen preferentially reacted with the Nb-depleted phase α-like-U to form monolithic β-UH3Nbx, and the alloy microstructure played an important role in hydride growth.

Enhancement of the cyclotron effective mass in U0.03Th0.97Ru2Si2

Electronic states of a dilute uranium alloy U0.03Th0.97Ru2Si2 have been investigated by using de Haas-van Alphen (dHvA) measurements on single crystal samples. Quantum oscillations were successfully observed for the field along the principal axes. The dHvA frequency of the observed branches roughly agrees with those of the reference compound ThRu2Si2, indicating the change of Fermi surface volume is not significant. On the other hand, the dHvA amplitude is strongly diminished compared to ThRu2Si2. Furthermore, cyclotron effective masses for corresponding branches are strongly enhanced. The latter effects are indicative of the strong scattering as well as the mass renormalization due to 5f moments.

Metallic Uranium: Protective Material with Enhanced γ-Ray Absorption

Research performed at the Bochvar All-Russia Research Institute for Inorganic Materials (VNIINM) and the All-Russia Research Institute of Experimental Physics (VNIIEF) on the development of the TUK-117 multipurpose transport packaging is presented. The TUK-117 packaging possesses biological protection based on depleted uranium and intended for transporting and storing spent nuclear fuel. The higher capacity of the metallic uranium alloy for absorbing of γ-rays and the alloy’s high mechanical characteristics have made it possible to develop a design meeting IAEA specifications. The technology for fabricating safety-engineered articles from uranium and its alloys by pouring the alloy into a prepared mold made of corrosion-resistant steel is patented.

Characterization of Respirable Uranium Aerosols from Various Uranium Alloys in Fire Events

Aerosols dispersed from the oxidation of various uranium alloys exposed to air and direct flame impingement from combustible substrates are characterized. An apparatus was designed to incorporate desired characteristics of previous experiments on uranium to sample aerosol on a kilogram scale in a laboratory environment. Previous studies involving β-phase stabilized uranium (99.25 wt% U:0.75 wt% Ti) were benchmarked using an identical alloy with identical characteristics of the original specimens. Other studies involving a-phase uranium (100 wt% U) were also benchmarked in this experiment. Unique to this study is the use of γ-phase stabilized uranium (94 wt% U:6 wt% Nb). These three alloys represent the crystallographic range of typical uranium metals, providing a complete spectrum of potential uranium aerosolization. Oxidation rates and extents observed in these experiments were directly comparable to existing data and provided correlation between previous studies. These experiments indicate a distinct or...

Progress in physics design of fusion-fission hybrid energy reactor

In this paper, we propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER), based on current fusion science and technology and well-developed fission technology. Design rules are listed and a primary concept blanket with uranium alloy as fuel and water as coolant is put forward. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. The interaction between the fusion neutron and the uranium fuel with the aim of achieving greater energy multiplication and tritium sustainability is studied. Other concept hybrid reactor designs are also reviewed. Integral neutron experiments were carried out to verify the credibility of our proposed physical design. The combination of the physical design with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology basis for future development of the FFHER concept in China.

Integral neutronics experiments in analytical mockups for blanket of a hybrid reactor

The paper describes recent progress in integral neutronics experiments in the analytical mockups for the blanket in a fusion-fission hybrid energy reactor. A conceptual blanket of the hybrid reactor is mainly loaded with natural uranium and lithium material. In the fission fuel region, uranium material and light water are arranged alternately. The mockups of the conceptual blanket are designed and used for checking neutron property of the blanket by integral experiments. Based on materials available, the spherical fission mockup for fission research and plutonium production consists of three layers of depleted uranium shells and several layers of polyethylene and graphite shells. The spherical lithium mockup for tritium production consists of depleted uranium and LiPb alloy shells. The cubic mockup consists of natural uranium and polyethylene and its structure is basically consistent with one of the fuel region. In the mockups with the D-T neutron source, the plutonium production rates, uranium fission rates and tritium production rates are measured, separately. The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data.

Defect structures induced by high-energy displacement cascades in γ uranium

Displacement cascade simulations were conducted for the γ uranium system based on molecular dynamics. A recently developed modified embedded atom method (MEAM) potential was employed to replicate the atomic interactions while an embedded atom method (EAM) potential was adopted to help characterize the defect structures induced by the displacement cascades. The atomic displacement process was studied by providing primary knock-on atoms (PKAs) with kinetic energies from 1keV to 50keV. The influence of the PKA incident direction was examined. The defect structures were analyzed after the systems were fully relaxed. The states of the self-interstitial atoms (SIAs) were categorized into various types of dumbbells, the crowdion, and the octahedral interstitial. The voids were determined to have a polyhedral shape with {110} facets. The size distribution of the voids was also obtained. The results of this study not only expand the knowledge of the microstructural evolution in irradiated γ uranium, but also provide valuable references for the radiation-induced defects in uranium alloy fuels.

The Research Status of the Preparation and Properties of the Tungsten Heavy Alloy

Due to the excellent physical and chemical characteristics and mechanical properties of the tungsten heavy alloy (WHA), the WHAs have been widely applied in many areas of the national economy. In this paper, the concept of the WHA is introduced briefly, and the preparation process and properties of the WHAs are summarized. Compared with depleted uranium alloy, the WHA shows broad application prospects in the field of kinetic energy projectile.

Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

Metal fuel slugs of U–Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U–10 wt% Zr and U–10 wt% Zr–5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions.

Interaction phenomena between ceramic coatings and liquid uranium alloys

Owing to the high chemical reactivity of molten uranium alloys, the use of traditional graphite crucibles for casting fuel slugs for a sodium-cooled fast reactor (SFR) is problematic. Moreover, rare earth (RE) elements retained in the fuel slugs for an SFR, which are extracted from the spent fuel by pyro-processing, are more reactive than uranium melt. Therefore, in this study, Y2O3 single-layer coatings with thicknesses of approximately 50, 70, and 120 μm and double-layer coatings of TaC/Y2O3 and Y2O3/TaC were plasma-sprayed onto niobium substrates and tested for thermal shock resistance and compatibility against U–10 wt% Zr and U–10 wt% Zr–5 wt% RE melt. The Y2O3 single-layer coating, regardless of coating thickness, and the TaC/Y2O3 double-layer coating showed good contact at the interface between the coating and the niobium substrate, with no deterioration after the thermal cycling test. In the interaction studies, the single- and double-layer coatings showed good compatibility with the U–Zr melt. However, the Y2O3 coatings with thicknesses of approximately 50 and 70 μm showed severe penetration of the U–Zr–RE melt and reacted with the niobium substrate. The single-layer Y2O3 coating with a thickness of 120 μm and the double-layer TaC/Y2O3 coating exhibited the most promising performance among the candidate coatings.

次临界能源堆物理设计进展

In this paper, we propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER), based on current fusion science and technology and well-developed fission technology. Design rules are listed and a primary concept blanket with uranium alloy as fuel and water as coolant is put forward. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. The interaction between the fusion neutron and the uranium fuel with the aim of achieving greater energy multiplication and tritium sustainability is studied. Other concept hybrid reactor designs are also reviewed. Integral neutron experiments were carried out to verify the credibility of our proposed physical design. The combination of the physical design with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology basis for future development of the FFHER concept in China.

Thermomechanical process optimization of U-10 wt% Mo – Part 1: high-temperature compressive properties and microstructure

Nuclear power research facilities require alternatives to existing highly enriched uranium alloy fuel. One option for a high density metal fuel is uranium alloyed with 10wt% molybdenum (U–10Mo). Fuel fabrication process development requires specific mechanical property data that, to date has been unavailable. In this work, as-cast samples were compression tested at three strain rates over a temperature range of 400–800°C to provide data for hot rolling and extrusion modeling. The results indicate that with increasing test temperature the U–10Mo flow stress decreases and becomes more sensitive to strain rate. In addition, above the eutectoid transformation temperature, the drop in material flow stress is prominent and shows a strain-softening behavior, especially at lower strain rates. Room temperature X-ray diffraction and scanning electron microscopy combined with energy dispersive spectroscopy analysis of the as-cast and compression tested samples were conducted. The analysis revealed that the as-cast samples and the samples tested below the eutectoid transformation temperature were predominantly γ phase with varying concentration of molybdenum, whereas the ones tested above the eutectoid transformation temperature underwent significant homogenization.

Phase transformation studies in U–Nb–Zr alloy

Phase transformation diagrams provide fundamental informations for designing thermomechanical processes being a must regarding uranium alloys nuclear fuels. The work shows the evaluation of a kinetic transformation diagram for U–7.5Nb–2.5Zr (wt.%) based on both calorimetry experiments and dilatometry allied to X-ray diffraction analysis. Calorimetry measurements in scanning and drop modes can detect enthalpies of heating and transformation onset points from ambient up to select isotherms while the dilatometer is used to scan for sample volume changes related to phase transformations. The resulted kinetic diagram shows the gamma phase is stable for this alloy, guiding the rolling deformation process to temperature ranges where this phase remains for longer periods. Comparing to the literature results, the low temperature transformation (300–400°C) is shifted to longer times accordingly to the disclosed TTT kinetic diagram. Therefore, two forming process windows can be proposed at 200°C and 400°C neighborhood where gamma-phase remains for enough time to accomplish total reduction.

Formation of uranium based nanoparticles via gamma-irradiation

The ability to fabricate nuclear fuels at low temperatures allows for the production of complex Uranium metal and alloys with minimum volatility of alloy components in the process. Gamma irradiation is a valuable method for the synthesis of a wide range of metal-based nanoparticles. We report on the synthesis via room temperature radiolysis and characterization of uranium (depleted, d-U) metal and uranium–lathanide (d-ULn, Ln=lanthanide surrogates) alloy nanoparticles from aqueous acidic salt solutions. The lanthanide surrogates chosen include La and Eu due to their similarity in ionic size and charge in solution. Detailed characterization results including UV–vis, TEM/HR-TEM, and single particle EDX (elemental analyses) are presented for the room temperature formed nanoparticle products.

Mechanical and thermal behaviour of U–Mo and U–Nb–Zr Alloys

Abstract Nuclear fuels composed of uranium alloys in monolithic and dispersed forms are being considered for research and compact power reactors due to their density properties (greater than 15 g-U/cm 3 ) and fast heat transfer. U–Nb–Zr and U–Mo alloys are the most promising systems for plate fuel elements owing to their broad γ-phase stability field, which shows higher ductility and isotropic behaviour, allowing extensive fabrication capability. In the present work, γ-phase stabilized U–7.5Nb–2.5Zr and U–10Mo alloys were characterized by mechanical and thermal analyses for comparison of their behaviour under deformation and heat-treatment. The results demonstrate that the alloys have substantially different properties regarding deformation, kinetics phase transformation and recovery/recrystallization. The main results show that U–Nb–Zr is superior regarding fabrication capabilities although the γ-phase is less stable than U–Mo alloys.

Bond Characterization of Plasma Sprayed Zirconium on Uranium Alloy by Microcantilever Testing

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 was characterized by microcantilever beam testing. Test results revealed the effect of specific flaws such as cracks and pores on the bonding of interfaces with a sampling area of approximately 20 μm2. TEM examination showed the Zr/U-Mo interface to contain rows of very fine grains (5-30 nm) with the Zr in contact with UO2. Bond characteristics of plasma sprayed samples were measured that are similar to those of roll bonded samples showing the potential for plasma sprayed Zr coatings to have high bond strength.

Gibbs energies of formation of the intermetallic compounds of U–Sn system

Gibbs energies of formation of the intermetallic compounds, USn3, U3Sn7, USn2, and USn were determined by using high temperature molten salt galvanic cells: U(s)//UCl3 in LiCl–KCl (eutectic)//U–Sn alloy, where pure uranium was used as one of the electrodes and a two phase alloy of uranium and tin as the other. The two phase alloys used in cells I, II, III and IV as the electrodes were 〈USn3〉+{Sn}, 〈U3Sn7+USn3〉,〈USn2+U3Sn7〉 and 〈USn+USn2〉, respectively. The Gibbs energies of formation (ΔfG0) of USn3, U3Sn7, USn2 and USn with respect to α-U and liquid Sn, in the temperature range 682–905K, were determined to be given by ΔfG0(USn3)(kJmol-1)=-172.8+0.061T(K),ΔfG0(U3Sn7)(kJmol-1)=-527.8+0.200T(K),ΔfG0(USn2)(kJmol-1)=-174.9+0.073T(K),ΔfG0(USn)(kJmol-1)=-176.9+0.064T(K) The Gibbs energy of formation for U3Sn7, USn2 and USn have been obtained for the first time.

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys

Extended abstract of a paper presented at Microscopy and Microanalysis 2012 in Phoenix, Arizona, USA, July 29 – August 2, 2012.

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.