Uranium-molybdenum Alloy Research Articles

Formation of uranium nitride nanoparticles via mechanical alloying of uranium-molybdenum alloy fuels in gaseous nitrogen

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and fuel loading capability. However, U-Mo systems are susceptible to irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interactions. To address these shortcomings, this research investigated the formation of uranium mononitride (UN) nanoparticles within a 90 wt% U/ 10 wt% Mo (U-10Mo) matrix to act as a prospective defect sink for fission products at nanometric hetero-interfaces. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen atmosphere. Variations of the milling process investigated included media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. Characterization of the fuel microstructure was completed using light element analysis, X-ray diffraction, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. UN nanoparticles measuring 1–5 nm in radius were observed in the U-Mo matrix as early as 1 h into the mechanical alloying process. Milling time in excess of 10 h was found to lead to deleterious effects induced by the stainless-steel milling media.

Hot Hydrogen Testing of W-Coated UN Kernels in a Mo30W Matrix

Ceramic uranium mononitride (UN) is being considered as a reactor fuel for nuclear thermal propulsion. To avoid or reduce the dissociation of UN at the high temperatures needed, embedding it in a metallic matrix (cermet) has been proposed. To assess the viability of this concept, hot hydrogen testing of tungsten-coated UN kernels embedded in a Mo-30 wt% W (Mo30W) alloy matrix has been performed at temperatures from 1800°C to 2300°C. Both the isolated kernels and kernels consolidated by spark plasma sintering in the Mo30W matrix were tested. In addition to direct observations and mass loss measurements, the samples were analyzed by X-ray diffraction (XRD) and scanning electron microscopy (SEM)/energy dispersive X-ray spectroscopy (EDS) after each run. The decomposition of UN started at 1800°C despite the coating and matrix, and increased at 2000°C. Uranium seeped through the tungsten grain boundaries of the coating at all temperatures. The consolidated sample expanded irregularly at 2000°C through the formation of voids, and SEM/EDS analysis showed uranium-containing veins in the matrix consisting of U2Mo according to the XRD data. The observed pore generation at 2000°C was explained by the formation of water vapor from residual oxides and diffused hydrogen. At 2200°C and above, both the kernels and the consolidated samples melted through the formation of uranium or low–melting point uranium-molybdenum alloys.

A European manufacturing line for monolithic U–Mo bare foil production

The nuclear research reactor Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II), operated by the Technical University of Munich (TUM) in Germany, provides intense thermal neutron beams for the international scientific community. The uranium fuel currently used for the compact core is an assembly of fuel plates composed of dispersed U3Si2 powder with an enrichment up to 93%, dispersed in aluminium (Al) powder and cladded with an Al alloy. Within the international efforts to minimise the usage of HEU (Highly Enriched Uranium), the FRM II aims to convert its compact core to LEU (Low Enriched Uranium), i.e., a fuel with an enrichment in 235U lower than 20%. The most promising candidate is the metallic monolithic uranium–molybdenum alloy (U–Mo) with zirconium (Zr) coating and Al-based cladding. Here, we show the first European pilot line for U–Mo bare foil manufacturing, implemented in collaboration with the European fuel manufacturer Framatome-CERCATM in Romans-sur-Isère, France. This line involves innovative manufacturing processes, including casting, laser welding, hot rolling, and laser cutting. An overview of the bare foil development with the first results is presented, including visual inspections of foils produced from the hot rolling, the laser cutting and the manufacturing tools used. These first depleted uranium (DU) bare foils are analysed to extract the best manufacturing parameters for future industrialisation. This European manufacturing process gives an alternative solution to the US development for U–Mo monolithic bare foil manufacturing. It represents the first step for European monolithic fuel manufacturing, which could push other European research reactors to use the monolithic U–Mo for their conversion.

First principles study of H<sub>2</sub> dissociation, H atom and O atom diffusion on Mo doped γ-U (100) surface

<sec>As an important uranium alloy, U-Mo alloy has excellent mechanical properties, structural stability and thermal conductivity, which is an important nuclear reactor fuel and tank armor. However, there exists a serious of fundamental problems of U-Mo alloy which need solving for practical applications. U-Mo alloy is easily subjected to surface corrosion of small molecules including the H<sub>2</sub>, O<sub>2</sub>, H<sub>2</sub>O, and CO<sub>2</sub>. The hydrogen corrosion and oxidation will have significant influence on it. In order to further investigate the reaction mechanism, based on the density functional theory and the transition state algorithm, the first principles calculation of γ-U (100) with Mo atom doping and Mo coating is carried out.</sec><sec>Firstly, the minimum energy path of H<sub>2</sub> molecule is calculated for the dissociation adsorption on Mo-U and 4Mo-U surface. Secondly, the transition states of H and O atoms are studied during surface diffusing between adjacent most stable adsorption sites. Thirdly, the bulk phase diffusion of H and O atoms are investigated and the relationship is analyzed between adsorption energy and adsorption height in the bulk phase diffusion.</sec><sec>The results show that when H<sub>2</sub> molecule is adsorbed at the configuration of top horizontal position, the H atom needs to overcome a barrier to triggering off the H—H bond-broken and then is adsorbed on surface bridge site by the neighboring atoms. The energy barrier for H<sub>2</sub> dissociation on 4Mo-U is higher than that of Mo-U. Meanwhile, the lower energy barrier is required for O atom to diffuse in Mo-U, so that it can be adsorbed, dissociated and diffused quickly, and then forming an oxidation film on the surface. Furthermore, both H and O atoms need to cross the energy barrier to diffuse into the body phase, forming chemical bonds with the atoms and staying in the body phase stably finally.</sec><sec>In this paper, we comprehensively analyze the dissociation and diffusion of the initial stage for hydrogen corrosion and oxidation on uranium-molybdenum alloy by theoretical studies. The results lay a foundation for theoretically exploring the surface corrosion mechanism of U-Mo alloy. Meanwhile, They provide theoretical support for investigating burn-in and corrosion of uranium-molybdenum alloy, predicting material properties under extreme and special environment, and providing a reference for further research on corrosion resistance of uranium-molybdenum alloy.</sec>

Features of the behavior of the dispersion fuel METMET under irradiation

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.

Compositional partitioning during early stages of oxidation of a uranium-molybdenum alloy

Compositional partitioning during uranium alloy oxidation was studied via complementary ex situ - in situ atom probe tomography. Nanoscopic volumes of uranium- 22 at. % molybdenum were exposed to air at room temperature/atmospheric pressure for 30–60 minutes (ex situ), and 300 ∘C - 10−5 mbar O2 gas for 2–5 minutes in a chemical reaction chamber attached to an atom probe system (in situ). For all environmental conditions, a hypostoichiometric uranium oxide is formed. Reaction fronts are observed at oxide/metal, oxide/hydride, and outer oxide/environment interfaces. Results reveal Mo redistributes across the oxide/metal interface, with a tendency for enrichment in the outer oxide. The formation of a hydrogen-rich subsurface layer between the oxide and base alloy accompanies oxidation in both air and oxygen gas environments. Carbon and silicon impurity elements also redistribute to the outer oxide, contributing to oxide film composition.

Effect of Concentration of Mo on the Mechanical behavior of γ UMo: an Atomistic Study

We performed molecular dynamics simulation on nanoindentation ofγ phase Uranium Molybdenum alloys using spherical indenter. A ternary potential developed for UMoXe was utilized. We calculated the updated values for hardness and reduced elastic modulus at different concentrations of Mo. The whole process of deformation and dislocation analysis was visualized using OVITO. We found an increase in deformation withincreasein stress while dislocations are not found anyhow induced defects have been distributed throughout the simulation cell randomly. The increase in concentration affected the hardness and reduced elastic modulus significantly. This study provides insights into the structure and mechanical characteristics of γ UMo under deformation.

Benchmark of the Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Component Critical Configurations

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) was a prototype for the U.S. National Aeronautics and Space Administration’s Kilopower Program. KRUSTY has a highly enriched uranium–molybdenum alloy (with 7.65 wt% molybdenum) annular core reflected by beryllium oxide with an outer stainless steel shield. Five configurations from the experimental campaign were chosen to be evaluated as benchmark cases. Uncertainties were evaluated in five categories: (1) criticality measurement, (2) mass and density, (3) dimensions, (4) material compositions, and (5) positioning. The largest contribution to the overall uncertainty in each case was from the radial alignment of the movable platen. A simplified model was created to increase computational efficiency, and an average bias of –16 pcm was calculated due to the simplifications. Sample calculations were completed for each case using MCNP6.2, COG, and MC21, all with ENDF/B-VIII.0 nuclear data. For MCNP6.2, the average difference (absolute value) between the calculated and experimental keff for the five configurations was 14 pcm for both the detailed and the simplified models. The keff results from all three codes are within 1σ of the benchmark values. KRUSTY’s value as a benchmark is due to its sensitivity to beryllium and molybdenum. For beryllium, KRUSTY adds an 18th benchmark with a total cross-section sensitivity greater than 0.05%/%/(unit lethargy). For molybdenum, KRUSTY adds a 9th benchmark with a total cross-section sensitivity greater than 0.004%/%/(unit lethargy).

Microstructural Changes and Chemical Analysis of Fission Products in Irradiated Uranium-7 wt.% Molybdenum Metallic Fuel Using Atom Probe Tomography

Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.

Atomic Simulations of U-Mo under Irradiation: A New Angular Dependent Potential

Uranium-Molybdenum alloy has been a promising option in the production of metallic nuclear fuels, where the introduction of Molybdenum enhances mechanical properties, corrosion resistance, and dimensional stability of fuel components. Meanwhile, few potential options for molecular dynamics simulations of U and its alloys have been reported due to the difficulty in the description of the directional effects within atomic interactions, mainly induced by itinerant f-electron behaviors. In the present study, a new angular dependent potential formalism proposed by the author’s group has been further applied to the description of the U-Mo systems, which has achieved a moderately well reproduction of macroscopic properties such as lattice constants and elastic constants of reference phases. Moreover, the potential has been further improved to more accurately describe the threshold displacement energy surface at intermediate and short atomic distances. Simulations of primary radiation damage in solid solutions of the U-Mo system have also been carried out and an uplift in the residual defect population has been observed when the Mo content decreases to around 5 wt.%, which corroborates the negative role of local Mo depletion in mitigation of irradiation damage and consequent swelling behavior.

An atomistic study of defect energetics and diffusion with respect to composition and temperature in γU and γU-Mo alloys

• Vacancy and self-interstitial formation energies were calculated in γ U and γ U-Mo. • Self-diffusion and interdiffusion coefficients in γ U- x Mo ( x = 7 , 10, 12 wt.%) were calculated. • Results can be used as input parameters in mesoscale and engineering scale fuel modeling. Uranium-molybdenum (U-Mo) alloys are promising candidates for high-performance research and test reactors, as well as fast reactors. The metastable γ phase, which shows acceptable irradiation performance, is retained by alloying U with Mo with specific quenching conditions. Point defects contribute to the atomic diffusion process, defect clustering, creep, irradiation hardening, and swelling of nuclear fuels, all of which play a role in fuel performance. In this work, properties of point defects in γ U and γ U- x Mo ( x = 7, 10, 12 wt. % ) were investigated. Vacancy and self-interstitial formation energies in γ U and γ U- x Mo were calculated with molecular dynamics (MD) simulations using an embedded atom method interatomic potential for the U-Mo system. Formation energies of point defects were calculated in the temperature range between 400 K and 1200 K. The vacancy formation energy was higher than the self-interstitial formation energy in both γ U and γ U- x Mo in the evaluated temperature range, which supports the previous results obtained via first-principles calculations and MD simulations. In γ U- x Mo, the vacancy formation energy decreased with increasing Mo content, whereas the self-interstitial formation energy increased with increasing Mo content in the temperature range of 400 K to 1200 K. The self-diffusion and interdiffusion coefficients were also determined in γ U- x Mo as a function of temperature. Diffusion of U and Mo atoms in γ U- x Mo were negligible below 800 K. The self-diffusion and interdiffusion coefficients decreased with increasing Mo concentration, which qualitatively agreed with the previous experimental observations. Point defect formation energies, self-diffusion coefficients, and interdiffusion coefficients in γ U- x Mo calculated in the present work can be used as input parameters in mesoscale and engineering scale fuel performance modeling.

Crystallographic texture of hot rolled uranium-molybdenum alloys

The uranium molybdenum (U-Mo) alloys have potential to be used as low enriched uranium nuclear fuel in research, test and power nuclear reactors. U-Mo alloy with composition between 7 and 10 wt% molybdenum shows excellent body centered cubic phase (γ phase) stabilization and presents a good nuclear fuel testing performance. Hot rolling is commonly utilized to produce parallel fuel plate where it promotes the cladding and the fuel alloy bonding. The mechanical deformation generates crystallographic preferential orientation, the texture, which influences the material properties. This work studied the texture evolution in hot rolled U-Mo alloys. The U7.4Mo and U9.5Mo alloys were melted in a vacuum induction furnace, homogenized at 1000°C for 5 h and then hot rolled at 650°C in three height reductions: 50, 65 and 80%. The crystalline phases and the texture were evaluated by X-ray diffraction (XRD). The as-cast and processed alloys microstructures were characterized by optical and electronic microscopies. The as-cast, homogenized and deformed alloys have γ phase. It was found microstructural differences between the U7.4Mo and U9.5Mo alloys. The homogenized treatment showed effective for microsegregation reduction and were not observed substantial grain size increasing. The deformed uranium molybdenum alloys presented α, γ, θ texture fibers. The intensity of these texture fibers changes with deformation step.

Methodology for ultrasonic inspection of dispersion type U–Mo fuel plates

Nuclear fuel based on uranium-molybdenum alloys (U–Mo) showed a promising performance under irradiation tests, except for an undesirable interaction between the U–Mo fuel particles and the aluminum matrix. This development requires new methods and evaluation criteria for inspection of this fuel. Non-destructive analysis techniques, such as Ultrasonic Test (UT), can play a key role in the characterization and in-pile monitoring of test fuel plates, in particular, for the anisotropic fuel core.This paper proposes the establishment of a relationship between attenuation and transmission effects for U–Mo fuel plates with different uranium loads, based on mathematical processing of UT signals. For this purpose, the Chilean Nuclear Energy Commission (CCHEN) manufactured a set of 8 dispersion type test plates, with uranium loads from 1 to 8 gU/cm3. The U–Mo fuel plates were evaluated in transmission mode with a Panametrics PR 5800 pulser-receiver ultrasonic system, using frequencies of 20 and 25 MHz. The data set collected for each plate was processed using the Fourier Function transform of the data function in MATLAB®. The obtained results showed that the attenuation coefficient of the ultrasonic waves increased linearly as a function of the uranium load. From 12.3 dB/mm for 1 gU/cm3 to 74.2 dB/mm, for 8 gU/cm3. The transmission percentage of the signal reduces linearly as a function of uranium load and the grain boundaries present in the Al cladding, decreasing from 66.7% for 1 gU/cm3 to 9.1% for 8 gU/cm3. In conclusion, the uranium load and the metal forming processes strongly affects the response of the fuel plates subjected to UT scanning. These results established an acceptance criterion for fuel plates based on this response, and the transformations experienced by the fissile U–Mo material.

Accident tolerant fuels to replace uranium dioxide

The thermal conductivities for the traditional fuel UO2 and for the Accident Tolerant Fuel (uranium Molybdenum alloy, U-10Mo and uranium silicide, U3Si2) are calculated and compared. Then the temperature distributions on these core reactor, namely on the surface of the rod and in the center line of the rod were determined. The calculation is carried out using MATHCAD program. The calculations showed enhancement of thermal conductivity of the Accident Tolerant Fuel as they increased linearly with increasing temperature, and reduction of their center line temperatures of the rods. Beside, their steep thermal gradient, which may reduce the induced heat stresses in the core of the reactor.

Evaluation of the Possibility of Fabricating Uranium-Molybdenum Fuel for VVER by Powder Metallurgy Methods

It is shown that the technological scheme for fabricating uranium dioxide pellets is suitable for fabricating uranium-molybdenum pellets. It permits using the standard powder metallurgy methods to fabricate pellets from uranium-molybdenum powder obtained by thermocentrifugal atomization of ingots of uranium-molybdenum alloy. The proposed technology uses standard equipment operating in easily reproducible regimes.

Atomistic Simulation of the Fission-Fragment-Induced Formation of Defects in a Uranium–Molybdenum Alloy

The formation of defects in a uranium–molybdenum alloy induced by xenon and zirconium ion irradiation with energies typical of fission fragments has been simulated within the two-temperature atomistic model and Monte Carlo method. This has allowed estimating the total number of primary radiation defects formed at the “thermal spike” stage and in collision cascades for different temperatures. The calculated threshold stopping powers of ions responsible for the formation of defects at the thermal spike stage are in good agreement with the available theoretical and experimental data.

Preparation of γ-Uranium-Molybdenum Alloys by Electrochemical Reduction of Solid Oxides in LiCl Molten Salt

Uranium-molybdenum (U-Mo) alloys especially the γ-U-Mo phase, with excellent irradiation performance at high temperatures, have been extensively studied as low-enriched-uranium candidate fuels for strengthening nuclear security and nonproliferation worldwide. Herein, we provide a facile method to successfully prepare the γ-U-Mo alloys (U-10Mo and U-8Mo) directly from oxide precursors via electro-reduction by constant cell voltage electrolysis in LiCl molten salt at 986 K. The reduction mechanism was elucidated through cyclic voltammetry using the metallic cavity electrode (MCE) combined with XRD, SEM and EDS characterizations toward the electrolysis products under different conditions. It was found that the reduction of mixed MoO3-UO2 oxides may take place in three steps. Firstly, MoO3 was reduced to Li4Mo5O12, and then Li4Mo5O12 was converted into Mo. Finally, the U-Mo alloys were formed due to the underpotential deposition of U on the pre-formed Mo. In addition, the effects of cell voltage, sintering temperature and mixture composition on the electrochemical reduction were investigated. The results found that the reduction rate increased with the augmentation of applied potential and the decrease of sintering temperature. Moreover, the favorable composition of the mixed precursor varied from 24 at% to 21at%. The reduction conversion rate was determined to be 93.7% by ICP and TGA.

PRODUCTION OF URANIUM-MOLYBDENUM ALLOY AS A CANDIDATE FOR NUCLEAR RESEARCH REACTOR FUEL

PRODUCTION OF URANIUM−MOLYBDENUM ALLOY AS A CANDIDATE FOR NUCLEAR RESEARCH REACTOR FUEL. Research and development on high density uranium for nuclear research reactor fuel is still in progress. Uranium-molybdenum alloy is one of the strongest candidates of nuclear research reactor fuel material. The properties and characteristics of U-Mo alloy is of important consideration for the selction of the fabrication techniques for the production of the fuel. In this work, uranium-molybdenum (U-Mo) alloys with varied molybdenum content have been produced succesfully by arc melting technique. The molybdenum content variations were 7 %wt, 8 %wt, 9 %wt and 10 %wt Mo. The melting process was done 5 times to achieve homogenization. Metallographic micrograph shows the presence of dendritic structure. XRD examination result affirms the presence of 2 phases of γ-U phase and d-U2Mo phase. Microhardness Vickers test shows higher hardness value for Uranium-molybdenum alloy with higher molybdenum content.Keywords: U−Mo alloy, research reactor, fuel.

Surface Characterization of Depleted Uranium–Molybdenum to Determine Surface Coating Compatibility

Scanning electron microscopy, transmission electron microscopy, and X-ray photoelectron spectroscopy have been used to characterize the surface of depleted uranium molybdenum (DU-Mo) alloys as a chemical surrogate to determine potential challenges with the surfaces of manufactured and stored U-Mo foils and powders. Even when stored and shipped in an inert atmosphere, U-Mo has a tenacious surface contamination of oxygen and carbon. The 8 at. % molybdenum (DU-8Mo) powder and 10 at. % molybdenum (DU-10Mo) foil samples have surface contamination of oxygen and carbon in different ratios that is hundreds to thousands of nanometers thick. The DU-8Mo powder sample has been stored in an inert atmosphere and as a result has a lower carbon-to-oxygen ratio at the surface than the DU-10Mo foil sample that was stored in air. This surface contamination has not been removed by up to 20 min of argon ion sputtering nor with 5% hydrogen in argon heat treatment for up to 96 h at 950°C.

Characterization of Plasma-Sprayed Zirconium Coatings on Uranium Alloy Using Neutron Diffraction

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, is phase pure (alpha-Zr), and has preferred crystalline orientation due to directional solidification influenced by the substrate crystalline orientation. Also, there is a slight (~ 0.01 A for a direction and ~ 0.016 A 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.