Uranium Alloys Research Articles

Study on the modification and mechanism of N-Mo co-doping for U-2.0Nb alloy surface by pulsed laser

The surface of a U-2.0Nb alloy pre-coated with a Mo layer was successfully modified by pulsed laser in N2 gas, resulting in an N-Mo co-doped layer. Evaluation of this layer at 363 K in a water vapor environment demonstrated its exceptional corrosion resistance for the U-2.0Nb substrate. Analysis using XRD and XPS revealed that the layer comprises uranium nitrides (UN and UO2-xNy) and uranium alloys (U, Nb, Mo), whose chemical stability and distribution are crucial for significantly enhancing corrosion resistance. This study also confirms that high-energy pulsed laser modification induces the sequential formation of reactants, favoring those with higher melting points. This technology shows promise for developing high melting point, ablation-resistant layers and enhancing the corrosion resistance of advanced nuclear fuels like UN/UC/UC1-xNx, thereby broadening the application scope of pulsed laser modifications.

Optimization of conceptual design on the lead-based modular nuclear power reactor core loaded with U-10Zr alloy fuel

As one of the forth-generation nuclear energy system reactor types, lead fast reactor has good safety and economical properties due to the stable chemical properties of the coolant and the proliferation characteristics of the fuel, and modular nuclear power faster reactor designed for nuclear plant can further improve the economics of the reactor. In this article, the conceptual design of the lead-based modular power reactors with different power levels loaded with uranium alloy fuel is found to be found that when reactor core size increased to a certain level, the proliferation performance is too high due to the increase of the reactor core size under a specific core life such as 2000EFPD, so at the end of core life, the reactor core still has a large remaining reactivity. The proliferation advantage of the core cannot be fully released during the current core life time. Based on this phenomenon, in this article, we optimized the conceptual design of lead-based modular nuclear power reactor core loaded with uranium alloy fuel, and proposed to choose the appropriate rod to diameter ratio and effective density of fuel based on the power level and life time of the core. By adjusting the amount of uranium and 235U per unit volume, the proliferation performance of the core can be changed to match the power level and life time of the core. So the reactivity of core during the life period does not change, which not only reduce the difficulty of the reactivity control, but also make full use of the proliferation performance of the core. And at the same time, the reasonable rod to diameter ratio can provide safety and design margin for the analysis of thermal and hydraulic safety, and effectively improve the economy and safety of the core.

A Review on the Adiabatic Shear Banding Mechanism in Metals and Alloys Considering Microstructural Characteristics, Morphology and Fracture

The current review work studies the adiabatic shear banding (ASB) mechanism in metals and alloys, focusing on its microstructural characteristics, dominant evolution mechanisms and final fracture. An ASB reflects a thermomechanical deformation instability developed under high strain and strain rates, finally leading to dynamic fracture. An ASB initially occurs under severe shear localization, followed by a significant rise in temperature due to high strain rate adiabatic conditions. That temperature increase activates thermal softening and mechanical degradation mechanisms, reacting to strain instability and facilitating micro-voiding, which, through its coalescence, results in cracking failure. This work aims to summarize and review the critical characteristics of an ASB’s microstructure and morphology, evolution mechanisms, the propensity of materials against an ASB and fracture mechanisms in order to highlight their stage-by-stage evolution and attribute them a more consecutive behavior rather than an uncontrollable one. In that way, this study focuses on underlining some ASB aspects that remain fuzzy, allowing for further research, such as research on the interaction between thermal and damage softening regarding their contribution to ASB evolution, the conversion of strain energy to internal heat, which proved to be material-dependent instead of constant, and the strain rate sensitivity effect, which also concerns whether the temperature rise reflects a precursor or a result of ASB. Except for conventional metals and alloys like steels (low carbon, stainless, maraging, armox, ultra-high-strength steels, etc.), titanium alloys, aluminum alloys, magnesium alloys, nickel superalloys, uranium alloys, zirconium alloys and pure copper, the ASB propensity of nanocrystalline and ultrafine-grained materials, metallic-laminated composites, bulk metallic glasses and high-entropy alloys is also evaluated. Finally, the need to develop a micro-/macroscopic coupling during the thermomechanical approach to the ASB phenomenon is pointed out, highlighting the interaction between microstructural softening mechanisms and macroscopic mechanical behavior during ASB evolution and fracture.

Deformation strengthening of uranium and dilute uranium alloys

The strength and ductility of uranium and dilute uranium alloys can be substantially improved by deformation strengthening. The yield strength of unalloyed uranium can be nearly doubled by rolling in the vicinity of 250 °C. The embrittling effect of hydrogen is also overcome, eliminating the need for vacuum outgassing to obtain good ductility. Significant differences in longitudinal versus transverse tensile properties result, in part, from texture introduced by unidirectional rolling. Warm rolling of U-2.3 %Nb results in both traditional strain hardening and a significant Bauschinger effect, resulting in differences between tensile and compressive yield strength. A similar combination of strain hardening and Bauschinger effect also occurs in deformation strengthened U-0.75 %Ti. Warm rolling of U-0.75 %Ti results in significantly better combinations of tensile yield strength and ductility than can be obtained by conventional age hardening. Warm rolling prior to aging also reduces the loss of ductility which typically accompanies conventional age hardening. An approach is developed in which the effects of prior deformation on tensile and compressive yield strengths are analyzed in terms of texture, long-range dislocation, and Bauschinger effects. The Bauschinger effect is shown to vary in sign with the type of deformation, thus providing opportunities to tailor tensile and compressive properties.

Selection of mechanical properties of uranium and uranium alloys after corrosion based on machine learning

Uranium and uranium alloys play a vital role as service materials in strategic equipment, but their mechanical properties can be adversely affected by corrosion. Therefore, accurately predicting the mechanical properties of uranium and uranium alloys after corrosion holds significant importance. In this study, we created a database to investigate the impact of oxygen corrosion on the mechanical properties of uranium and uranium alloys. Utilizing this database, we developed a feature-guided decision tree algorithm to predict various tensile properties, including yield strength, tensile strength, elongation, and cross-section shrinkage. Our research highlights three key findings. Firstly, we established a machine learning modeling framework that effectively predicts tensile properties and exhibits potential for predicting other properties of uranium and uranium alloys. Secondly, through feature engineering, we uncovered crucial correlations involving reaction time, reaction temperature, alloy type, phase structure composition, and phase number. These correlations significantly enhanced the performance of machine learning models in predicting tensile properties after oxygen corrosion. Lastly, by employing the decision tree algorithm guided by feature engineering, we successfully predicted the mechanical properties of uranium and uranium alloys after oxygen corrosion with a prediction error of less than 5%.

Simultaneous separation and determination of Ti and Fe in uranium alloy by ion chromatography

A simple and reliable ion chromatographic (IC) method for the simultaneous separation and quantification of Ti and Fe in U-Ti alloys is proposed. The developed method does not require pre-separation of matrix uranium and hence direct injection of the dissolved sample into IC is feasible. Separation was realized on a reversed phase column dynamically modified with ion interaction reagent (IIR). Suitability of three IIRs was evaluated and Camphor-10-sulphonic acid (CSA) was selected. Separation of Ti and Fe was achieved by mandelic acid whereas the strongly retained U(VI) was eluted with tartaric acid. The developed method was calibrated for Ti and Fe in presence of U(VI). The limit of detection (LOD) for Ti and Fe is 0.2 µg/mL and 0.02 µg/mL, respectively. The precision of measurements was better than 5 % (RSD) at 1 ppm of Ti and 0.2 ppm of Fe. The real samples were analysed with the developed method.

POSSIBILIITY OF URANIUM EXTRACTION FROM SPENT NUCLEAR FUEL IN FUSED ELECTROLITES CONTAINING RARE ELEMENTS

Link for citation: Nikitin D.I., Polovov I.B., Rebrin O.I. Possibility of uranium extraction from spent nuclear fuel in fused electrolytes containing rare elements . Bulletin of the Tomsk Polytechnic University. Geo Аssets Engineering, 2023, vol. 334, no. 10, рр. 210-218. In Rus. The relevance of the research is caused by the plans of using electrolytic separation of metalized spent nuclear fuel as a stage in pyrochemical reprocessing of mixed nitride uranium-plutonium fuel. The main aim is to determine the parameters of uranium electrolytic separation from an alloy with simulators of fission products (precious metals and rare earth elements) simulating spent nuclear fuel in salt mixtures based on 3LiCl–2KCl with additives of rare earth element chlorides. Objects: model spent nuclear fuel – uranium alloy with simulators of fission product (precious metals and rare earth elements) with a mass compound of Pd:Ru:Ag:Rh=25:1:3:3, Nd:Ce:La:Pr:Sm:Y=15:10:5:5:5:1). Methods: electrorefining, X-ray fluorescence and X-ray diffraction method of analysis, inductively coupled plasma mass spectrometry, assessment of distribution of components in the system. Results. The data obtained showed that uranium deposits have dendrite formations of alpha-uranium at 550 °C in orthorhombic crystal system toward needle cathode current density. The resulting cathode deposits are free from impurities of ruthenium, rhodium, molybdenum, praseodymium and yttrium. The purification coefficient for palladium reaches 3000, and for silver 1700. Noble metals accumulate in the anode sludge, and even with the complete depletion of the anode material. The concentration of noble metals in the cathode deposit does not exceed 0,0015 wt %. Despite the high concentration of rare-earth chlorides in the electrolyte, which simulates the accumulation of rare-earth ions during repeated reprocessing of spent nuclear fuel, the concentration of rare-earth metals in the cathode product did not exceed 0,007 wt %. During uranium electrowinning from a model spent nuclear fuel in the 3LiCl–2KCl–UCl3 electrolyte (10,1 wt %) with REE chlorides, which imitate their accumulation in the electrolyte during repeated processing, at 550 °C, as well as the initial cathode current density of 0,2 A/cm2, the specific amount of electricity is 1,0 A∙h/cm2. A cathode uranium deposit is released with a current efficiency exceeding 90 % with all anode mass depletion, and purification coefficient 1800 for the sum of precious metals and for the sum of rare earth metals.

Isoplethal study of phase formation and morphology in uranium-304L steel via scanning electron microscopy

ABSTRACT Understanding the formation of uranium alloys with steel is important to advance nuclear technologies involving U metal fuels and machining U metal, and for nuclear forensics applications. No known phase diagram for the quaternary U-(M = Fe, Ni, Cr) system exists. We synthesize samples of U-304 L steel (nominal composition 70.1:18.3:10.4 at% Fe:Cr:Ni) across the U composition range 4.45—63.35 at%U by arc melting under inert conditions. Using the binary UFe phase diagram as a reference, we identify four U-steel alloy phases. We find the known U-Fe analogue phases UM2 and U6M, and two low-U composition phases with nominal compositions UM10 and U2M7. We apply a correlation length analysis to backscatter scanning electron microscopy images of sectioned and polished cross sections to quantify the domain formation length scale. We demonstrate that these depend heavily on the initial composition and range from 30 nm to 1.5 µm. This result, in particular, could be applicable to theoretical predictions of transport properties. Furthering our understanding of U alloy phase formation with important structural elements such as steel primaries is foundational in developing future nuclear technology. 1

Electrochemical extraction of uranium on the gallium and cadmium reactive electrodes in molten salt

The electrochemical extraction of uranium in ternary low melting LiCl–KCl–CsCl eutectic on inert and reactive electrodes via different electrochemical techniques was investigated. It was established that the electrochemical reduction process of U(III) ions on the inert W electrode was irreversible and proceeded in one stage. On reactive liquid Ga and liquid Cd electrodes the reduction of uranium ions took place with the considerable depolarization with the formation of UGa2, UGa3 and UCd11 intermetallic compounds. Thermodynamic characteristics of uranium compounds and alloys were calculated. The conditions for the extraction of uranium from the electrolyte in the form of alloys on both liquid reactive electrodes via potentiostatic electrolysis were found.

Removal of bond sodium from Fermi-1 blanket assemblies using a melt-drain-evaporate process

A series of runs was performed to demonstrate the removal of bond sodium from full-length unirradiated Fermi-1 radial blanket elements and an entire radial blanket assembly using a melt-drain-evaporate process. The runs were performed in an inert atmosphere glovebox, where the blanket material was breached and loaded into a retort assembly. It was then subjected to elevated temperature and reduced pressure to melt, drain, and evaporate the bond sodium, while simultaneously collecting the sodium distillate in a crucible at the base of the assembly. After each run, the depleted uranium alloy slugs in every element slid out of its cladding, providing the first indication that the blanket material was effectually devoid of sodium metal. Entire columns of depleted uranium alloy slugs and associated cladding from select treated elements were separately subjected to quantitative analyses for residual sodium metal on the element surfaces, yielding removal efficiencies of ≥99.9998%. The remaining uranium alloy slugs and cladding segments from the series of runs were piece-wise immersed in ethanol to react with any residual sodium and provide a visual means of its presence or absence through the corresponding formation of hydrogen gas bubbles or lack thereof. Accordingly, the predominant absence of sodium reactivity on the balance of treated Fermi-1 blanket material was observed. The successful outcome of this demonstration provides a path forward for the treatment and disposal of 34 metric tons heavy metal (MTHM) of irradiated sodium-bonded blanket material from the decommissioned Fermi-1 reactor.

Anodic Processes of Uranium Alloys Containing Palladium and Neodymium in 3LiCl–2KCl–UCl3 Melts

Anodic Processes of Uranium Alloys Containing Palladium and Neodymium in 3LiCl–2KCl–UCl3 Melts

Effect of Niobium on Sintering Process of Gamma Phase Uranium Alloys Fuel Pellets

Abstract As a candidate material for metallic fuel, U-Mo metal fuel pellets are the most promising. U-Mo and U-Mo-Nb alloy pellets with a certain porosity were successfully prepared by the process of hydrogenation/dehydrogenation—compression molding—argon liquid-phase sintering. In order to study the effect of Nb addition on γ phase uranium alloy fuel pellets, microstructure and thermophysical properties of the samples were investigated using scanning electron microscope (SEM), in situ synchrotron X-ray diffraction (XRD), and differential scanning calorimeter (DSC) measurements. Results showed that with the increase of Nb content in the pellets from the nonadd to micro-adding, Nb can facilitate the diffusion of Mo into the U matrix, resulting in the formation of a metastable γ-U phase. Meanwhile, during the same liquid phase sintering process of U-Mo fuel pellets, with the increase of Nb content, the number of secondary phases in U-Mo fuel pellets gradually decreased, while the size and number of voids of the secondary phases decreased. The specific heat capacity and thermal diffusivity of porous γ phase uranium alloys fuel pellets with different density were measured and thermal conductivity from 100 °C to 600 °C were calculated according to the experiment results. It is suggested that the thermal conductivity will increase with the density of pellets increasing.

Development and experimental verification of the new WHA sinters intended for kinetic energy projectiles

AbstractDepleted uranium (DU) and tungsten heavy alloys (WHA) are commonly used as kinetic energy projectiles penetrators due to their excellent properties such as remarkably high density and strength which significantly affect their armor penetration capabilities. This article presents the results of laboratory and field tests of the new WHA sinter which summarize the results of a development project intended to increase the strength of the above-mentioned sinters used in the production of kinetic ammunition. The alloy with the composition W91–6Ni–3Co was used for the tests. The obtained alloy was subjected to cold swaging deformation with reduction of 25%. The parameters of individual technological processes were determined based on previous research. These studies concerned the impact of the degree of cold swaging deformation on the mechanical properties of the 91W–6Ni–3Co alloy. The above-mentioned work presents the results of UTS test, Charpy impact strength measurement, hardness and microhardness measurement, and the results of WHA microscopic observations in two states: after heat treatment and the final state after cold swaging deformation with reduction of: 25%. The next purpose of the research was to check the strength and functioning of the new WHA rods used in 120 mm sub-caliber projectiles. This examination was crucial for the completion of this research project, and its results enabled a synthesis of laboratory results and experimental tests.

ELECTROREFINING OF URANIUM ALLOYS CONTAINING PALLADIUM AND NEODYMIUM IN 3LiCl–2KCl–UCl<sub>3</sub> MELTS

The technology of pyrochemical processing of mixed nitride uranium-plutonium spent fuel, realizable at the experimental and demonstration energy complex of the site of the Siberian Chemical Plant, includes several operations with the ultimate goal of isolating the target fission products. It’s planned to use the electrofining of the products of the previous stage, metallized spent nuclear fuel, аs the penultimate stage of processing. It’s necessary to determine the processes and technological modes of electrolytic refining of alloys modeling the product of this stage of the processing module to implement electrolytic refining. This paper presents the results of electrofining of model alloys (simulating the raw materials of the stage of electrofining processing) on an enlarged laboratory electrolyzer. The initial parameters of uranium refining processes in melts based on 3LiCl–2KCl–UCl3 were determined earlier. The basic parameters of refining were the use of electrolyte 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) and conducting experiments at 550°C. Uranium alloys containing palladium and neodymium were prepared by direct fusion of uranium metal, PdAP-1 grade palladium metal powders and neodymium metal (99.99%) in a medium of high-purity argon (99.998%). The data obtained showed that at a temperature of 550°C, cathode precipitates are typical dendritic forms of alpha-uranium in rhombic syngony with a tendency to needle formation with an increase in cathode current density. An increase in the company time and cathode current density leads to a decrease in the current output due to short-circuiting of the electrodes with cathode sediment needles or metal shedding from the cathode. The modes of the cathode process have been experimentally refined as a result of electrofining. When electrofining alloys U–Pd(1.59 wt %), U–Pd(1.62 wt %), U–Pd(1.54 wt %), U–Pd(1.58 wt %)–Nd(5.64 wt %), U–Pd(1.84 wt %)–Nd(6.49 wt %), U–Pd(1.79 wt %)–Nd(6.54 wt %), uranium cathode precipitates were obtained, which were subjected to chemical analysis, which showed the high purity of the resulting metallic uranium, as well as the absence of metallic palladium and molybdenum in it. The palladium purification coefficient exceeds 5000, the neodymium purification coefficient exceeds 1000, which meets the requirements for purification from fission products at this stage of pyrochemical processing of spent fuel. Palladium accumulates in anode slime, while the bulk of neodymium passes into the molten electrolyte.

ANODIC PROCESSES OF URANIUM ALLOYS CONTAINING PALLADIUM AND NEODYMIUM IN 3LiCl–2KCl-UCl<sub>3</sub> MELTS

At the reprocessing module of the pilot demonstration power complex site of the Siberian Chemical Combine, a combined technological scheme for the reprocessing of mixed nitride uranium-plutonium spent fuel consisting of pyrochemical operations, hydrometallurgical refining of uranium, plutonium and neptunium is implemented step by step. According to this scheme, the target pyrochemical reprocessing products, purified from the main mass of fission products with actinoid content not less than 99.9%, are sent for hydrometallurgical reprocessing. For pyrochemical reprocessing it is necessary to develop a technology of electrorefining of metallised spent nuclear fuel. To carry out electrolytic refining it is necessary to define processes and regimes of anodic dissolution of alloys simulating product of this head operation “metallization”. In the present work the results of investigations of processes of anodic dissolution of U–Pd and U–Pd–Nd model alloys with different concentrations of palladium and neodymium in melts based on 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) at 550°C using different methods are presented. Uranium alloys containing palladium and neodymium were prepared by direct alloying of uranium metal and palladium metal powders of PdAP-1 grade, and neodymium metal (99.99%) in high-purity argon medium (99.998%). Electrochemical measurements were performed using an Autolab 302N potentiostat/halvanostat equipped with a Booster 20A high-current module. The anodic polarisation curves consist of only one oxidation wave which was attributed to the dissolution of uranium metal. Increasing the palladium content in the alloy from 1.5 to 10.0 wt %, does not affect the shape of the polarisation curves. The increase of neodymium content in the alloy from 1.0 to 10.0 wt % also does not affect the shape of polarization curves. Electrorefining parameters of uranium alloys containing palladium and neodymium were determined. The limiting current density of uranium evolution from uranium alloys containing palladium and neodymium in the electrolyte 3LiCl–2KCl–UCl3 (10.1 wt % UCl3) at 550°C was 0.4 A/cm2. It was shown that palladium does not diffuse into the melt as a result of anodic dissolution and neodymium accumulates in the electrolyte only when the alloy is refined with 10.0 wt % neodymium, which is much higher than the future real concentrations of electrotreated uranium alloy components in the technological chain of spent nuclear fuel processing.

Mechanism of Metallic Uranium and Bimetallic U-Ga, U-Cd Alloys Electrodeposition in Molten LiCl–KCl–CsCl Eutectic

Speciation and behavior of uranium (III) chloride in the ternary low melting LiCl–KCl–CsCl eutectic was studied by electrochemistry and spectroscopy techniques. Cathodic reduction of U(III) ions on inert (tungsten) and reactive (gallium, cadmium) electrodes was investigated at 623–923 K using cyclic and square wave voltammetry. The potential scan rate was changed from 0.075 to 0.5 V s−1 in all experiments. It was established that the electrochemical reduction process on the inert electrode was irreversible, proceeded in one stage, and was controlled by the charge transfer. Formation of uranium alloys with gallium and cadmium was studied using active liquid Ga and Cd electrodes. Reduction of uranium ions of the reactive electrodes proceeded with considerable depolarization. The effect of current density on the composition of the cathodic product was considered. Conditions for the electrochemical production of alloys of a given composition were determined.

Finite temperature properties of uranium mononitride

Uranium mononitride (UN) is a promising nuclear fuel that combines the advantageous properties of readily used UO2 and uranium alloys, such as high melting temperature and high uranium density, and thermal conductivity, respectively. A better understanding of UN behavior at operating temperatures can be obtained from finite temperature data, such as elastic properties. To get this information, ab initio molecular dynamics (AIMD) simulations were performed at five different temperatures using constant volume (NVT) and constant pressure (NPT) ensembles. Initially, the performance of PBE functional in reproducing experimental crystallographic properties and magnetic ordering is assessed. The finite temperature phonon dispersions are calculated using NVT simulation results, which show a softening of the phonon modes with increasing temperature. The NPT results are used to obtain the thermal expansion of UN and finite temperature electronic properties. The calculated thermal expansion is compared with our measurements using neutron diffraction. Additionally, the temperature dependent elastic properties of UN are evaluated using the strain-stress method in AIMD simulations, indicating that UN becomes softer and more compressible with increasing temperature. Also, the calculated Young’s modulus slope is in very good agreement with the experiment. The finite temperature heat capacity and electronic thermal conductivity are calculated from AIMD simulations, which are in better agreement with the experiment than the heat capacity and thermal conductivity calculated using the structures relaxed at 0 K. Lastly, the thermal diffusivity from AIMD has opposite temperature dependence compared to experimental results, which we argued comes from the underestimated electronic thermal conductivity.

A Machine Learning Method for Predicting Corrosion Weight Gain of Uranium and Uranium Alloys.

As an irreplaceable structural and functional material in strategic equipment, uranium and uranium alloys are generally susceptible to corrosion reactions during service, and predicting corrosion behavior has important research significance. There have been substantial studies conducted on metal corrosion research. Accelerated experiments can shorten the test time, but there are still differences in real corrosion processes. Numerical simulation methods can avoid radioactive experiments, but it is difficult to fully simulate a real corrosion environment. The modeling of real corrosion data using machine learning methods allows for effective corrosion prediction. This research used machine learning methods to study the corrosion of uranium and uranium alloys in air and established a corrosion weight gain prediction model. Eleven classic machine learning algorithms for regression were compared and a ten-fold cross validation method was used to choose the highest accuracy algorithm, which was the extra trees algorithm. Feature selection methods, including the extra trees and Pearson correlation analysis methods, were used to select the most important four factors in corrosion weight gain. As a result, the prediction accuracy of the corrosion weight gain prediction model was 96.8%, which could determine a good prediction of corrosion for uranium and uranium alloys.

Dynamic Compression Mechanical Properties of TiZrHfX0.3 High‐Entropy Alloys

High‐entropy alloys (HEAs) are gradually replacing armor‐piercing core materials such as depleted uranium and tungsten alloys in high strain rate, high temperature, and high pressure service environments due to their excellent mechanical properties (high strength, high strain rate strengthening effect, and strong shear self‐sharpening characteristics). The dynamic compression performance of TiZrHf‐based HEAs with different elements, different strain rates, and multiple loadings is tested by the self‐constructed separate Hopkinson compression bar experimental system. The results show that the TiZrHfCu0.3 alloy has the largest elastic modulus and the strongest deformation resistance. TiZrHfAl0.3 and TiZrHfCu0.3 show a certain strain rate effect. The elastic modulus increases with the increase of loading times. The ultimate strength of TiZrHfAl0.3 remains unchanged after multiple loading, while the ultimate strength of TiZrHfCu0.3 and TiZrHfNi0.3 decreases gradually.

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>