Uranium Corrosion Research Articles

The role of CO on initial oxidation behavior of α-U(001) surface: A first principles study

A first-principles approach based on density-functional theory has been used to investigate the corrosion resistance of alpha-U in the CO environment. Calculations show that O2 molecules spontaneously dissociate on the uranium surface, and the two O atoms formed by dissociation tend to adsorb on the hole sites and bind to the surface in the form of a U–O bond to emit a large amount of heat. The CO molecules occur on the surface of uranium as a non-dissociative chemical. The mechanism of CO inhibiting the adsorption of O2 molecules stems from the fact that CO molecules occupy the optimal adsorption sites. Another inhibition mechanism, the combination of C atoms and O atoms to form bonds and consume oxygen atoms, has little effect on uranium corrosion.

Uranium corrosion characterization by handheld laser-induced breakdown spectroscopy

A commercial-off-the-shelf handheld laser-induced breakdown spectroscopy system was modified, calibrated and tested for discrimination between uranium oxide and uranium hydride when assessing uranium corrosion products. Operating parameters, system modifications, and onboard data analysis were developed to accomplish the analytical task. This work represents the first laser-induced breakdown spectroscopy analysis of bulk uranium hydride, a pyrophoric material, within an inert glovebox. Finally, the work shown here illustrates a rapid identification technique for uranium corrosion types from pure UO2 or U3O8 to pure UH3.

A review of the mechanisms, reaction products and parameters affecting uranium corrosion in water

• Database on the reported mechanisms of uranium corrosion in water. • Investigation of the reaction processes governing early uranium oxidation. • Report on the controlling influences of the uranium–water corrosion reaction. • Physical and chemical characteristics of the corrosion products. • Conditions under which UH 3 formation occurs. This work conducts a review on reported mechanisms of uranium corrosion in water for studies carried out over the past 65 years. Close examination of the processes occurring in the distinct stages of the reaction between uranium and water was performed, including the initial parabolic and then ‘linear’ stages of reaction. Parameters that affect the kinetics and mechanism of the reaction such as pressure, temperature, gas and sample impurities, pH, thermal pre-treatment and alloying are all discussed separately. Electrochemistry results of uranium corrosion tests are also evaluated. The physical and chemical characteristics of the arising solid reaction products, primarily UO 2 and UH 3 , are extensively reported for existing literature and discussed. The review uses the available literature to highlight the conditions under which bulk-hydride formation is favoured and more likely to occur, thereby helping to delineate conditions or storage scenarios that are best avoided to help prevent hydride formation in nuclear wastes containing metallic uranium.

Early age hydration and application of blended magnesium potassium phosphate cements for reduced corrosion of reactive metals

Magnesium potassium phosphate cements (MKPC) were investigated to determine their efficacy towards retardation of reactive uranium metal corrosion. Optimised low-water content, fly ash (FA) and blast furnace slag (BFS) blended MKPC formulations were developed and their fluidity, hydration behaviour, strength and phase assemblage investigated. In-situ time resolved synchrotron powder X-ray diffraction was used to detail the early age (~60 h) phase assemblage development and hydration kinetics, where the inclusion of BFS was observed to delay the formation of struvite-K by ~14 h compared to FA addition (~2 h). All samples set within this period, suggesting the possible formation of a poorly crystalline binding phase prior to struvite-K crystallisation. Long-term corrosion trials using metallic uranium indicated that MKPC systems are capable of limiting uranium corrosion rates (reduced by half), when compared to a UK nuclear industry grout, which highlights their potential application radioactive waste immobilisation.

Monitoring uranium corrosion in Magnox sludge using X-ray computed tomography: A direct analogue to “legacy” fuel storage ponds

During the early years of nuclear industry within the UK, the typical route followed to handle Magnox (Mg-Al alloy) casing after separation from the nuclear fuel has been immersion in water temporarily in specifically-designed water ponds. Over the decades of storage, Magnox has been corroding in water, resulting in the formation of sludge within the ponds; known as Corroded Magnox Sludge. Uranium, which has been also contained within the ponds as a result of faults occurring during fuel-casing separation, is susceptible to corrosion in water-rich environments. Under certain circumstances, uranium corrosion products can be highly pyrophoric during oxidation in air. This raises serious concerns regarding the safety margins of nuclear pond decommissioning. The present study attempts to shed light on the behaviour of uranium in a Magnox sludge-surrogate environment, using X-ray computed tomography at distinct time intervals over an 18-month period. Focus was given onto the uranium-sludge interface to determine metallic corrosion onset and evolution.

Investigating the solution and diffusion properties of hydrogen in [formula omitted]-Uranium by first-principles calculations

The stability, solution and diffusion properties of an interstitial hydrogen atom in uranium metal have been firstly investigated by first-principles calculations. In energy, the octahedral site is more favorable for hydrogen to occupy than tetrahedra site with neglectable anisotropic perturbation. Besides, the effects of temperature on solution energy are quantified, which demonstrate the solution energy decreases fast with temperature. The calculated density of states and electronic charge re-distribution are analyzed. It is found the conductivity of metal uranium remains well after hydrogen occupied the interstitial position with lower concentration. The minimum migration pathways of interstitial hydrogen in uranium lattice are characterized by the climbing image nudged elastic band (CINEB) method. The obtained energy barriers are 0.239 eV, 0.298 eV and 0.313 eV with respect to O⟷T, O⟷O and T⟷T pathways with feeble structural deterioration. We believe our results for hydrogen diffusion in such a complex f-electron system not only provide en evidence for uranium corrosion but also supports the future experiments on measuring the hydriding rate and their interpretations.

Silicate stabilisation of colloidal UO2 produced by uranium metal corrosion

U(IV) mobility can be significantly enhanced by colloids in both engineered and natural environments. This is particularly relevant in decommissioning and clean-up of nuclear facilities, such as legacy fuel ponds and silos at the Sellafield site, UK, and in long-term radioactive waste geodisposal. In this study, the product of metallic uranium (U) corrosion under anaerobic, alkaline conditions was characterised, and the interaction of this product with silicate solutions was investigated. The U metal corrosion product consisted of crystalline UO2 nanoparticles (5–10 nm) that aggregated to form clusters larger than 20 nm. Sequential ultrafiltration indicated that a small fraction of the U metal corrosion product was colloidal. When the uranium corrosion product was reacted with silicate solutions under anaerobic conditions, ultrafiltration indicated a stable colloidal uranium fraction was formed. Extended X-ray absorption fine structure (EXAFS) spectroscopy and high resolution TEM confirmed that the majority of U was still present as UO2 after several months of exposure to silicate solutions, but an amorphous silica coating was present on the UO2 surface. This silica coating is believed to be responsible for formation of the UO2 colloid fraction. Atomic-resolution scanning TEM (STEM) indicated some migration of U into the silica-coating of the UO2 particles as non-crystalline U(IV)-silicate, suggesting alteration of UO2 at the UO2-silica interface had occurred. This alteration at the UO2-silica interface is a potential pathway to the formation of U-silicates (e.g. coffinite, USiO4).

In-situ, time resolved monitoring of uranium in BFS:OPC grout. Part 2: Corrosion in water

To reflect potential conditions in a geological disposal facility, uranium was encapsulated in grout and submersed in de-ionised water for time periods between 2–47 weeks. Synchrotron X-ray Powder Diffraction and X-ray Tomography were used to identify the dominant corrosion products and measure their dimensions. Uranium dioxide was observed as the dominant corrosion product and time dependent thickness measurements were used to calculate oxidation rates. The effectiveness of physical and chemical grout properties to uranium corrosion and mobilisation is discussed and Inductively Coupled Plasma Mass Spectrometry was used to measure 238U(aq) content in the residual water of several samples.

ZrO<sub>2</sub> Ceramic Composite<sub> </sub>Added with MgAl<sub>2</sub>O<sub>4</sub>: Thermal Shock Resistance and Corrosion Resistance in Molten Uranium

A kind of ceramic composite consisted of Mg-Y co-doped ZrO2 and MgAl2O4 was prepared as candidate material to replace the high destiny graphite crucible which was often used in nuclear industry. When the addition of MgAl2O4 was 7 wt%, samples with good thermal shock resistance were obtained. The obtained samples were tested up to 12 h in molten uranium, and no apparent cracks were found on the corrosion surface. Meanwhile, the corrosion rate decreased significantly from 30 mg·cm-2·h-1 to 3 mg·cm-2·h-1 as the corrosion time prolonged from 2 h to 12 h due to the formation of a protection layer consisted of UO2, Al3Zr2 and ZrC during the uranium corrosion test. These results suggested that the ceramic composite prepared in this study owned both good corrosion resistance to the molten uranium and thermal shock resistance at high temperatures,so it’s a high potential material to replace the high density graphite crucible in nuclear industry.

Plasma-Sprayed Ceramic Coatings for Barrier Applications Against Molten Uranium Corrosion

Ceramic coatings are applied on engineering components for protecting them from large thermal load and hot corrosion. Choices of coating material for protection against hot corrosion by uranium are few, because of its high reactivity. Yttrium oxide has a high melting temperature and is inert towards uranium. Therefore, yttrium oxide coatings are effective as a barrier against hot corrosion by uranium and its alloys. This paper gives a summary of the developmental work on plasma-sprayed yttria coatings for corrosion barrier applications against molten uranium. Results show that plasma-sprayed yttria coatings offer a long-term solution to hot corrosion problems.

Plasma spray-deposited lanthanum phosphate coatings for protection against molten uranium corrosion

Lanthanum phosphate is recommended as an effective barrier against molten uranium and its alloys. LaPO4 has a high melting point, is inert towards reactive metals and has low thermal conductivity, which make it an ideal candidate for corrosion barrier coatings in molten metal environment. Studies on plasma deposition of lanthanum phosphate, its characterization and interaction with molten uranium are reported in this paper. Lanthanum phosphate was synthesized by chemical route and the as-synthesized powder was converted to free flowing powder for plasma spray deposition. Lanthanum phosphate coatings were prepared on stainless steel substrates and the deposits were characterized for phase composition using X-ray diffraction (XRD), Raman spectroscopy and Fourier transformed infrared spectroscopy (FTIR). Results showed that the coatings retained the monazite structure of LaPO4 precursor powder. Microstructure of the coatings and interface by scanning electron microscope (SEM) revealed good bonding between the phosphate layer and substrate. Porosity of the coatings was observed to decrease with the increase in deposition power from 10kW to 20kW. Adhesion strength of the coatings was found to increase with deposition power. Corrosion studies of coated LaPO4 in molten uranium showed that lanthanum phosphate does not chemically react with molten uranium and LaPO4 coating offers adequate resistance to the substrate against corrosion attack by molten uranium.

Nuclear waste viewed in a new light; a synchrotron study of uranium encapsulated in grout.

How do you characterise the contents of a sealed nuclear waste package without breaking it open? This question is important when the contained corrosion products are potentially reactive with air and radioactive. Synchrotron X-rays have been used to perform micro-scale in-situ observation and characterisation of uranium encapsulated in grout; a simulation for a typical intermediate level waste storage packet. X-ray tomography and X-ray powder diffraction generated both qualitative and quantitative data from a grout-encapsulated uranium sample before, and after, deliberately constrained H2 corrosion. Tomographic reconstructions provided a means of assessing the extent, rates and character of the corrosion reactions by comparing the relative densities between the materials and the volume of reaction products. The oxidation of uranium in grout was found to follow the anoxic U+H2O oxidation regime, and the pore network within the grout was observed to influence the growth of uranium hydride sites across the metal surface. Powder diffraction analysis identified the corrosion products as UO2 and UH3, and permitted measurement of corrosion-induced strain. Together, X-ray tomography and diffraction provide means of accurately determining the types and extent of uranium corrosion occurring, thereby offering a future tool for isolating and studying the reactions occurring in real full-scale waste package systems.

Numerical Modeling of Galvanic Corrosion Behaviors on Uranium Surface

The Uranium and Titanium corrosion rates are described by a Tafel’s relationship, and the cathodic protection of Uranium is a function of a Wagner number. A numerical simulation of galvanic corrosion of Uranium surface under thin layer electrolyte is presented. The model considered that the effect of a circular defect and oxygen reduction and corrosion in the Uranium surface, the effect of electrolyte thickness and conductivity and defect radius on corrosion current distribution of Uranium with is investigated. The results shows that the corrosion rate at the center is non-uniform, and it could lead to the formation of a hemispherical-shaped pit. And the effect of radius is to increase the importance of the electrode kinetics relative to ohmic resistance, and to increase the potential difference between the center and edge of the Uranium surface, resulting in non-uniform corrosion current distribution.

Electrochemistry of Actinides and Selected Fission Products in the Head End of Spent Nuclear Fuel Reprocessing

The data on the uranium metal corrosion rate in the solutions of nitric acid (0,1 – 4M) and effect of complex forming agents on uranium corrosion properties are presented. The increase of HNO3 concentration caused the shift of corrosion potential from 38mV to 446mV and the increase of the corrosion rate from 0,02 to 0,62mg.cm-2h-1. Transpassivation potential of U metal was found weakly effected by HNO3 concentration varying from 448 to 470mV/Ag/AgCl. The addition of HCOOH to the electrolytes containing less than 3M HNO3 found to shift the values of corrosion potentials about 500mV towards negative direction reducing the passivation of U metal. The data on the kinetics of oxidative dissolution of PuO2 using Ag(II) and Am(VI,V) as mediators and the effect of the mediator generation techniques are discussed. The electrochemical properties of UC in the solutions 2 – 4M HNO3, results of the quantitative determination of “oxidizable carbon” in dissolver solutions are presented. The results of corrosion and dissolution studies of Tc metal and Tc - Ru alloys containing from 19 to 70 at.% Ru in 0.5 0– 6M HNO3 indicate the formation of passive films of Tc(IV) – Ru(III,IV) hydroxides at the electrode surface in the solutions containing less than 2M HNO3 at the potentials less than 650mV/Ag/AgCl. The increase of HNO3 concentration to values exceeding 3M and the shift of the electrode potential towards positive direction causes the transition of the Tc and Tc-Ru alloys to transpassive state. The values of transpassivation potentials increase with the increasing with HNO3 concentration. Quantitative dissolution of Tc metal without application of oxidation potential becomes possible in the electrolytes, containing more than 4M HNO3. The rate of Tc – Ru alloys dissolution is noticed to slow down with the increase of Ru content in the alloy.

Corrosion of uranium and its low content Zr, Nb, and Ru alloys in aqueous solutions

Corrosion of uranium and its alloys with low content (0.5–5.0 at %) of Zr, Nb, and Ru in water and bicarbonate aqueous solutions is studied; the effect of hydrogen peroxide, the main product of radiation processes, on the corrosion rate is elucidated. The rate of the primary corrosion process U +(2 +n)H2O=UO2·nH2O+ 2H2↑ is measured by electrochemical methods in anaerobic and aerobic conditions for uranium metal and its alloys containing 0.5 to 5.0 at % of Zr, Nb, and Ru. It is shown that the corrosion rates for the alloys are lower than that of reactor-grade uranium; however, the difference is rather close to the measurement error. The corrosion mechanism is studied; U(III) is shown to be rather unstable in neutral solutions when uranium(III) hydroxide is precipitated and no significant amount of U(III) and UH3 is present among the products of the metallic uranium corrosion in water. The kinetics of the second corrosion stage U(IV) + O2→U(VI) is studied by spectrophotometric method. It is shown that the reaction of U(IV) oxidation by atmospheric oxygen is similar in weakly acid solutions (pH 1.5–4.0) and in bicarbonate media: in particular, it has an induction period for uranium (IV) accumulation, after which the reaction accelerates; it is formally a first-order reaction with respect to uranium. The reaction mechanisms differ in the two media: in weakly acid solutions, after the appearance of U(VI), the reproportionation reaction proceeds; thus formed U(V) interacts with O2 faster than U(IV). In the bicarbonate medium, the acceleration of the reaction is due to the formation of a [U(IV)ΣU(VI)] complex whose reactivity is higher than that of uranium (IV). In the absence of bicarbonate, of great importance is the formation of a copolymer of U(IV) and U(VI), which at pH≥4 prevents formation of U(V). It is shown that on the introduction of hydrogen peroxide to aqueous solutions, the metallic uranium surface becomes transpassive, which increases the rate of corrosion process by at least an order of magnitude,. The introducing of oxidants and platinum mesh lowers the hydrogen accumulation at 120–150°C and, hence, the hydrogen-explosion danger of the uranium-water-corrosion-products system. Methods of deposition of metal oxide (Tc, Ru, Mo, W) films onto uranium surfaces by immersing uranium metal into Tc(VII), Ru(VI), or Mo and W heteropoly compound solutions are studied.

Variations in Depleted Uranium Sorption and Solubility with Depth in Arid Soils

This study examined the ability of alkaline desert soils to sorb depleted uranium (DU) as a function of soil horizon and assessed the solubility of corrosion and migration products from two DU kinetic energy penetrators exposed on the desert surface for a 22-y period. Both uranium corrosion products on the surface, and subsurface uranium originating from the dissolution of surface corrosion products followed by reprecipitation or sorption, were examined. A four-step sequential extraction method was used to classify uranium solubility at each site. Results show that distribution coefficient for uranium is highly variable, but can be correlated with the clay content (r = 0.55) and soil pH (r = 0.73) of the soil horizon considered. Surface corrosion products and near-surface uranium easily dissolve in a weak acid solution (25% acetic acid for two hours), suggesting a uranyl hydroxide form. As uranium migrates beyond several centimeters in depth, it forms insoluble aggregates with silicate minerals and requires strong acids to leach it. The formation of uranium-silicate mineral aggregates appears to be the limiting factor in vertical vadose zone transport of DU at the site investigated.

Corrosion of Depleted Uranium in an Arid Environment: Soil-Geomorphology, SEM/EDS, XRD, and Electron Microprobe Analyses

Corrosion of anthropogenic uranium in natural environments is not well understood, but is important for determining potential health risks and mobility in the environment. A site in the southwestern United States contains depleted uranium that has been weathering for approximately 22 years. Soil-geomorphic, SEM/EDS, XRD, and electron microprobe analyses were conducted to determine the processes controlling the uranium corrosion. Schoepite and metaschoepite are the primary products of corrosion, and occur as silica-cemented, mixed schoepite-metaschoepite/clay/silt aggregates, as schoepite/metaschoepite-only aggregates, or rarely as coatings upon soil grains. Current extraction procedures do not adequately explain the behavior of uranium in alkaline soils when amorphous silica and clay coatings are present. Soil geomorphology and chemistry at this site limit uranium mobility and decreases potential health risks. However, if land-use and/or regional climate changes occur, uranium mobility could increase.

Dissolution of metallic uranium and its alloys

This review focuses on dissolution/reaction systems capable of treating uranium metal waste to remove its pyrophoric properties. The primary emphasis is the review of literature describing analytical and production-scale dissolution methods applied to either uranium metal or uranium alloys. A brief summary of uranium's corrosion behavior is included since the corrosion resistance of metals and alloys affects their dissolution behavior. Based on this review, dissolution systems were recommended for subsequent screening studies designed to identify the best system to treat depleted uranium metal wastes at Lawrence Livermore National Laboratory (LLNL).

Characterization of uranium corrosion products involved in a uranium hydride pyrophoric event

Uranium metal corrosion products involved in a recent pyrophoric event were characterized using thermo-gravimetric analysis, X-ray diffraction, and BET gas sorption techniques to determine the effects of passivation treatment and long-term storage on chemical reactivity. Characterization was performed on corrosion products in three different conditions: immediately after separation from the source metal, after low-temperature passivation, and after passivation and extended vault storage. The hydride fraction and ignition temperature of the corrosion products were found to be strongly dependent on the corrosion extent of the source metal. There was little change in corrosion product properties resulting from low-temperature passivation or vault storage. The results indicate that the energy source for the pyrophoric event was a considerable quantity of uranium hydride present in the corrosion products, but the specific ignition mechanism could not be identified.

Improving adhesion between a segmented poly(ether-urethane) and a fluorocarbon copolymer coating

A moisture barrier coating of Kel-F 800, developed at LLNL to reduce uranium corrosion, had to be bonded to a porous ceramic. The adhesive could not bond too strongly or react with the coating and jeopardize its barrier properties. Methods of improving adhesion to the Kel-F coating were studied. Silane and titanate coupling agents and a fluorocarbon surfactant were somewhat effective at increasing adhesion depending on the application procedure. X-ray photoelectron spectroscopy (XPS) was used to demonstrate the presence of fluorosurfactant at the fracture interface. Postcuring at elevated temperatures (85/sup 0/C) also significantly improved adhesive strength to the fluorocarbon coating. This was attributed to thermal acceleration to interfacial diffusion of the urethane adhesive into the fluoropolymer surface.