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Convenient Syntheses of Trivalent Uranium Halide Starting Materials without Uranium Metal

Low-valent uranium coordination chemistry continues to rely heavily on access to trivalent starting materials, but these reagents are typically prepared from uranium turnings, which are becoming increasingly difficult to acquire. Here we report convenient syntheses of UI 3 (THF) 4 (THF = tetrahydrofuran) and UBr 3 (THF) 4 from UCl 4 , a more accessible uranium starting material that can be prepared from commercially available uranium oxides. UCl 3 (THF) 2 (1), UBr 3 (THF) 4 (2), and UI 3 (THF) 4 (3) were prepared by single-pot reductions from UCl 4 using KH and KC 8 and converted to 2 or 3 by halide exchange with the corresponding Me 3 SiX (where X = Br or I). Reduction of UI 4 (Et 2 O) 2 (4; Et 2 O = diethyl ether) and UI 4 (1,4-dioxane) 2 (5) was also shown to cleanly yield 3. Complex 1 was also synthesized separately by the addition of anhydrous HCl to U(BH 4 ) 3 (THF) 2 , which was prepared by thermal reduction of U(BH 4 ) 4 . All three trivalent uranium halide complexes were isolated in high crystalline yields (typically 85–99%) and their formulations were confirmed by single-crystal X-ray diffraction, elemental analysis, and 1 H NMR and IR spectroscopy. Elemental analysis conducted on triplicate samples of 1–3 exposed to vacuum for different time intervals revealed significant THF loss for all three complexes in as little as 15 min. Altogether, these results offer expedient entry into low-valent uranium chemistry for researchers lacking access to uranium turnings.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Application of HRGS for forensic characterization of uranium oxides, pure uranium metals and uranium alloys

A nondestructive iterative method for uranium-bearing material characterization with HRGS developed earlier in (J. appl. Rad. Isotopes, 166 (2020) 109433) is applied to determine matrix densities and uranium isotope masses of a variety of uranium materials, namely uranium ore, UO 2 and U 3 O 8 powders, fuel elements in the form of UO2 microspheres, uranium metal and uranium alloys. It is shown that U 3 O 8 powders with uranium mass fraction of about 84% can be distinguished from the powders of UO 2 with uranium mass fraction of about 87%; uranium products in the form of liquid or loose powder with matrix density of 0.5-2.0 g/cm 3 can be distinguished from uranium products in the form of compacted fuel elements with matrix density of 6.0-10.0 g/cm 3 and from pure metal uranium and uranium alloys with matrix density of 14.0- 19.0 g/cm 3 . In fuel microspheres based on UO2 the uranium mass fraction 88.02% measured by HRGS is consistent, within the measurement uncertainties, with the results of isotope dilution mass spectrometry 87.76±0.64 % and also is confirmed by X-ray diffraction technique. The uranium mass fraction of the uranium ore estimated as 0.08% by HRGS is consistent, within the measurement uncertainties, with the value 0.09±0.01% determined with WDXRF. Densities of two different uranium metal samples, estimated as 18.42 g/cm 3 and 19.33 g/cm 3 by HRGS are consistent with values 18.24±0.55 g/cm 3 and 18.86±0.59 g/cm 3 , respectively, obtained by the gas pycnometry technique.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Formation of uranium oxy-carbide and uranium carbide via conversion of polymer covered uranium dioxide by laser-based thermal processing

Conventional formation of carbonaceous uranium compounds requires bulk processing using furnace-based approaches. Here, a methodology employing polymer covered uranium dioxide and laser-based heating is explored to enable rapid, localized formation of carbonaceous uranium compounds. Specifically, heating of poly(methyl methacrylate) covered uranium dioxide powder to high temperatures using laser irradiation in argon and methane gaseous environments was investigated. Decomposition of material and reactions induced by laser irradiation were probed in situ by residual gas analysis using a benchtop mass spectrometer. In this study, to determine the effect on the resultant material phase, three different process parameters were varied: gaseous atmosphere, laser power, and laser irradiation time. Material processed under varying conditions was analyzed using powder X-ray diffraction and scanning electron microscopy. This work realized the conversion of uranium dioxide into uranium oxy-carbide(s) and uranium carbide(s) phases, at over 60 wt.%, via the polymer surface application and laser-based thermal decomposition methodology.

36 MATERIALS SCIENCE↗

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium diboride and uranium carbide

Here, the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breach of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium diboride and uranium monocarbide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UB 2 , and UC. Presented here is the literature for UB 2 and UC degradation modes, thermodynamics, and oxidation performance of the pure compounds and reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation behavior for high uranium density fuels for use in LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Uranium recovery from seawater using amidoxime-based braided polymers synthesized from acrylic fibers

Global demand for nuclear energy is expected to rise in the coming decades. To meet these growing needs, new uranium resources must be explored. One of the potential alternatives to traditional uranium mining is oceanic uranium. The capture and recovery of uranium from the ocean has been under investigation for some time, with many recent studies focused on amidoxime based adsorbents. These adsorbents while able to achieve high uranium recovery capacities are, nevertheless, expensive to produce and adsorb a significant amount of hard to remove vanadium. The purpose of this study is to evaluate the adsorption performance of amidoxime based polymer braids synthesized from acrylic fibers which are designed to significantly cut polymer synthesis and conditioning costs. Adsorption experiments were performed in a recirculating raceway flume system at environmental conditions, approximately 10.8 oC, with 40-micron prefiltered seawater over 28 days for small and large sized LCW polymer braids. For both braid sizes, the adsorption of vanadium was far lower on the LCW materials than on Oak Ridge National Laboratory’s amidoxime based AF1 and AI8 small braids, with the LCW small braids outperforming both materials with respect to uranium adsorption under similar conditions. Adsorption modeling was used to simulate the performance at higher temperatures based on 20oC experiments previously performed with similar materials. Simulation results indicated that the small LCW braids would have a less significant advantage with respect to uranium adsorption over both AF1 and AI8 at 20oC and adsorb slightly less uranium than AF1 at 31oC. Vanadium adsorption by LCW small braids was less than one third of vanadium adsorption by either AF1 or AI8 for all temperatures. This behavior indicates that the LCW materials are able to achieve comparable uranium adsorption to other amidoxime adsorbents previously developed, while being cheaper to produce and adsorbing significantly less vanadium.

Wiechert, Alexander I.↗

Using Fission-Track Radiography Coupled with Scanning Electron Microscopy for Efficient Identification of Solid-Phase Uranium Mineralogy at a Former Uranium Pilot Mill (Grand Junction, Colorado)

At a former uranium pilot mill in Grand Junction, Colorado, mine tailings and some subpile sediments were excavated to various depths to meet surface radiological standards, but residual solid-phase uranium below these excavation depths still occurs at concentrations above background. The combination of fission-track radiography and scanning electron microscope energy-dispersive X-ray spectroscopy (SEM-EDS) provides a uniquely efficient and quantitative way of determining mineralogic associations of uranium that can influence uranium mobility. After the creation of sample thin sections, a mica sheet is placed on those thin sections and irradiated in a nuclear research reactor. Decay of the irradiated uranium creates fission tracks that can be viewed with a microscope. The fission-track radiography images indicate thin section sample areas with elevated uranium that are focus areas for SEM-EDS work. EDS spectra provide quantitative elemental data that indicate the mineralogy of individual grains or grain coatings associated with the fission-track identification of elevated uranium. For the site in this study, the results indicated that uranium occurred (1) with coatings of aluminum–silicon (Al/Si) gel and gypsum, (2) dispersed in the unsaturated zone associated with evaporite-type salts, and (3) sorbed onto organic carbon. The Al/Si gel likely formed when low-pH waters were precipitated during calcite buffering, which in turn retained or precipitated trace amounts of Fe, As, U, V, Ca, and S. Understanding these mechanisms can help guide future laboratory and field-scale efforts in determining long-term uranium release rates to groundwater.

58 GEOSCIENCES↗

Uncovering Uranium Isotopic Heterogeneity of Fuel Pellets from the Fifth Collaborative Materials Exercise of the Nuclear Forensics International Technical Working Group

In 2017, the Nuclear Forensics International Technical Working Group (ITWG) organized their fifth 37 Collaborative Materials Exercise (CMX-5). The exercise samples were two uranium dioxide fuel pellets 38 manufactured from the same starting materials by different processes to have similar bulk isotopic 39 composition, but different spatial uranium isotopic distributions. Sets of identical materials were sent to 40 all participating laboratories, who then utilized their existing nuclear forensic capabilities to 41 independently analyse fuel pellets and identify similarities and differences of the materials’ 42 characteristics. The analytical methods used to probe the fuel pellets included ex situ, such as sectioning 43 or breaking up the pellets and analyzing dissolved pieces using inductively coupled plasma mass 44 spectrometry (ICP-MS), analyzing particles collected from intact or fragmented pellets by secondary ion 45 mass spectrometry (SIMS), as well as in situ methods, such as laser ablation coupled with ICP-MS, 46 autoradiography and nanoSIMS. In this paper we present the results of these independent analyses and 47 compare the capabilities of those nuclear forensic analytical methods to uncover details of the isotopic 48 heterogeneity of uranium fuel pellets.

Nuclear Forensic Analysis of Uranium Fuel Pellets,↗

Uranium-Mediated Peroxide Activation and a Precursor toward an Elusive Uranium cis -Dioxo Fleeting Intermediate

The activation of chalcogen–chalcogen bonds using organometallic uranium complexes has been well documented for S–S, Se–Se, and Te–Te bonds. In stark contrast, reports concerning the ability of a uranium complex to activate the O–O bond of an organic peroxide are exceedingly rare. Herein, we describe the peroxide O–O bond cleavage of 9,10-diphenylanthracene-9,10-endoperoxide in nonaqueous media, mediated by a uranium(III) precursor [(( Me,Ad ArO) 3 N)U III (dme)] to generate a stable uranium(V) bis-alkoxide complex, namely, [(( Me,Ad ArO) 3 N)U V (DPAP)]. This reaction proceeds via an isolable, alkoxide-bridged diuranium(IV/IV) species, implying that the oxidative addition occurs in two sequential, single-electron oxidations of the metal center, including rebound of a terminal oxygen radical. Furthermore, this uranium(V) bis-alkoxide can then be reduced with KC 8 to form a uranium(IV) complex, which upon exposure to UV light, in solution, releases 9,10-diphenylanthracene to generate a cyclic uranyl trimer through formal two-electron photooxidation. Analysis of the mechanism of this photochemical oxidation via density functional theory (DFT) calculations indicates that the formation of this uranyl trimer occurs through a fleeting uranium cis-dioxo intermediate. At room temperature, this cis-configured dioxo species rapidly isomerizes to a more stable trans configuration through the release of one of the alkoxide ligands from the complex, which then goes on to form the isolated uranyl trimer complex.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

In-field alpha spectrometry system for uranium enrichment verification of uranium hexafluoride

A portable In-Field Alpha Spectrometry (IFAS) system was developed to support on-site uranium enrichment determination for samples taken at facilities working with gaseous UF 6 . It consists of a modified commercial alpha spectrometer, a vacuum pump, a laptop computer, and associated connectors and cables. The IFAS uses specially designed Single-Use Destructive Assay (SUDA) samplers to capture gaseous UF 6 and convert and store it as UO 2 F 2 in a thin-film zeolite. To examine the quantitative performance of the IFAS system a large number of replicate SUDA samples was acquired for natural uranium (NU, 0.72 wt%, 23 samples), low enriched uranium 1 (LEU1, 2.33 wt% 235 U, 10 samples), and low enriched uranium 2 (LEU2, 4.62 wt% 235 U, 10 samples). The standard deviation (SD) values for uranium enrichment determination with the IFAS for NU, LEU1, and LEU2, in 20-hour assays, were 0.037 wt% 235 U, 0.063 wt% 235 U, and 0.072 wt% 235 U, respectively. In addition, a separate analysis suggests that comparable SD values can be achieved in less time, approximately 8-10 hours. For context, these SD values are well below the International Atomic Energy Agency’s International Target Values for the determination of uranium enrichment in non-destructive assay measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A High-Assay Low-Enriched Uranium Fuel Transportation Concept

The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production cycle leads to new challenges caused by various technical and regulatory hurdles. Current U.S. Nuclear Regulatory Commission–approved transportation package designs for UF6 with enrichments above 5 wt% provide relatively small payloads [=116 kg (250 lb)]. Furthermore, in accordance with 10 CFR 71.55, package design activities for fissile material enriched above 5 wt% need to consider water infiltration in the containment as part of the criticality safety evaluations. This study presents a transportation package concept for HALEU advanced nuclear reactor fuel with a significantly higher payload of up to 376 kg (830 lb) of fissile material per package and up to 1881 kg (4149 lb) of HALEU per legal weight truck. The anticipated chemical form of the transported material is UO2 downblended from available highly enriched uranium. The concept utilizes a combination of existing transportation packaging, 18 inner canisters, and a novel basket design that includes a borated aluminum flux trap. Criticality and shielding evaluations; fundamental structural, confinement, and thermal assessments; and studies on package operations are presented. The results of this study build significant confidence in the technical feasibility of a high-capacity HALEU transportation package concept while demonstrating the concept’s potential to meet U.S. regulatory requirements.

Eidelpes, Elmar↗

Xenon mobility in γ-uranium and uranium–molybdenum alloys

Diffusion in bcc uranium and U–Mo alloys is of great interest because fission gas and other fission products impact the performance of nuclear fuels. We investigate the mobility of xenon and molybdenum in bcc uranium (⁠$γ$-U) and metallic U–Mo alloys by calculating the migration energies of xenon and molybdenum for various local compositions using density functional theory. We also calculate the solute–vacancy binding energies of different solutes to vacancies in bcc uranium. We find that the solute–vacancy binding energy in bcc uranium is significantly higher than it is in other bcc metals (e.g., Fe and W). We also find that the migration energy of molybdenum is substantially higher than the migration energy of xenon, indicating that xenon is much more mobile than molybdenum in bcc uranium. The presence of molybdenum in the nearest-neighbor shell around a xenon atom typically increases the migration energy of xenon, which indicates a reduction of xenon mobility in U–Mo alloys compared to pure bcc uranium.

36 MATERIALS SCIENCE↗

Capability to Process and Characterize Uranium-Zirconium (U-Zr) and Uranium-Zirconium-Plutonium (U-Zr-Pu) Alloys

This work investigated co-reduction of anhydrous compounds of uranium, zirconium, and plutonium to produce uranium rich ternary nuclear fuel alloys. Metallothermic co-reduction is a novel method to produce all-metal nuclear fuels. Metal fuels offer thermal, fissility, compatibility, and security benefits over oxide fuels. Alloys of uranium–10% zirconium with plutonium contents of 0%, 2.5%, 5%, and 10% were produced, with yields of 60–85% of theoretical values in a traditional calciothermic bomb reduction apparatus. Microstructural analysis indicated transformation of uranium phase from gamma to beta and then alpha, in alternating lamellar plates typical of alpha phase uranium and delta phase uranium-zirconium, with zirconium and carbides at prior grain boundaries. Some of the analyses were inconclusive in their results and therefore require additional testing. Successful ternary co-reduction would simplify production of homogeneous feedstock and thereby streamline manufacture of homogeneous ternary metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Quantification of uranium metal and total uranium in U-UO 2 mixtures using H 3 PO 4 -HCl

Here, the selective dissolution of uranium metal from its mixtures with UO 2 was studied to establish a simple and rapid quantitative method. Uranium metal was found to selectively dissolve as U(IV) into a solution of phosphoric and hydrochloric acids, with quantitative dissolution occurring within 1 h. Total dissolution of uranium as U(VI) could then be obtained by nitric acid addition and heating for 0.5–1 h. Analysis of the resulting U(IV) and U(VI) solutions by ICP-OES or UV–Vis spectroscopy yielded uranium metal-to-UO 2 ratios with an accuracy of ± 10% or better. Accuracy of the ICP-OES method was slightly better than UV–Vis, but the simplicity of the latter makes it an attractive option for some applications.

36 MATERIALS SCIENCE↗

Formation of uranium disulfide from a uranium thioamidate single-source precursor

A single-source-precursor approach was developed to synthesize uranium-based materials outside of the typically-studied oxides. This approach allows for shorter reaction times, milder reaction conditions, and control over the chemicals present in synthesis. To this end, the first homoleptic uranium thioamidate complex was synthesized as a precursor for US 2 materials. Pyrolysis of the thioamidate results in decomposition via an alkene elimination pathway and formation of γ-US 2 , which has historically been hard to access without the need for a secondary sulfur source. Despite the oxophilicity of uranium, the method successfully forms US 2 without the inclusion of oxygen in the bulk final product. These findings are supported by simultaneous thermal analysis, elemental analysis, powder X-ray diffraction, and uranium L 3 -edge X-ray absorption fine-structure spectroscopy. This work represents the first example of a single-source precursor approach to target and synthesize actinide materials other than the oxides.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Certification of Uranium Isotope Amount Ratios in a Suite of Uranium Ore Concentrate Certified Reference Materials

Here, in this study, data from thirteen laboratories from around the world are presented for a successful certification of uranium isotope ratios in uranium ore concentrate (UOC) certified reference materials from the National Research Council Canada. Different mass spectrometric measurement principles (including SF-ICP-MS, quadruple ICP-MS, TIMS, MC-ICP-MS, SIMS and AMS) and measurement procedures were employed. In general, data from all sources exhibit good consistency with TIMS and MC-ICP-MS showing superior performance. The three candidate UOC CRMs (UCLO-1, UCHI-1 and UPER-1) have natural uranium isotope ratios with certified values of 0.5304(7) x 10 -4 , 0.5475(2) x 10 -4 and 0.5361(4) x 10 -4 for n( 234 U)/n( 238 U) and 0.007 2563(13), 0.007 2563(10) and 0.007 2542(11) for n( 235 U)/n( 238 U), respectively, with expanded uncertainty (k = 2) applicable to the last digit of the value given in the parentheses. Information values for n( 236 U)/n( 238 U) in these three CRMs, measured by AMS, are also provided: 10 x 10 -12 , 200 x 10 -12 and 22 x 10 -12 . The uncertainties of the proposed certified values of uranium isotope ratios in UOC CRMs are superior to available reference materials, and the values of n( 234 U)/n( 238 U) and n( 236 U)/n( 238 U) show significant variation among the three CRMs.

58 GEOSCIENCES↗

Uranium Carbide and Uranium-Metal Thermal Scattering Law Evaluations and Cross Sections [Slides]

New Uranium Carbide and Uranium Metal evaluations have been developed using the FLASSH platform. Ab initio LD/MD models converged electronic, magnetic, and dynamical properties and demonstrate good agreement with experimental data. Thermal scattering cross sections generate inelastic and elastic thermal neutron scattering libraries in FLASSH . New TSL evaluations occurred in the production of ENDF File 7 data for Uranium Carbide and Uranium Metal submitted to NNDC for ENDF/B-VIII.1 release.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

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. Here, 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In-Field Alpha Spectrometry to Assess Uranium Enrichment In Uranium Hexafluoride

A new In-Field Alpha Spectrometry (IFAS) system is being developed to allow quick turnaround measurements of uranium enrichment in uranium hexafluoride (UF6) samples. Samples are collected using specially-designed Single-Use Destructive Assay (SUDA) wafers which have an absorptive film of zeolite deposited onto a quartz or metal substrate. The SUDA samples, along with a special holder, are designed to be attached directly to a sampling tap at a gas centrifuge enrichment plant or a uranium conversion facility, allowing gaseous UF6 to come into direct contact with the zeolite. The zeolite absorbs this material and converts it into stable, safe UO2F2 within a few minutes. Once the sample has been collected it is transferred to the IFAS system where a solid-state, silicon alpha spectrometer is used to quantitatively measure the alpha emissions from 234U, 235U, and 238U in the sample. The IFAS system is a small, light-weight device that can be quickly turned on and put into use. The SUDA samples and IFAS hardware are designed so that thin-film spectra are produced, allowing a complete measurement with a final 235U:Utotal determination having a precision <±2% to be made in less than eight hours. This short time frame will allow an inspector to get results on-site within one working shift; further innovations may reduce this to even shorter times.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗