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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Energetics of oxidation and formation of uranium mononitride

Uranium mononitride (UN) is an advanced nuclear fuel currently being considered for use in several generation IV fast and thermal neutron spectrum core designs, with additional applications to thermal and electric nuclear propulsion reactors. Here, to better understand the thermal behavior and thermodynamic stability of UN, we investigated the bulk thermal oxidation process and thermochemical reactions, including the enthalpy of oxidation and standard enthalpy of formation, by conducting thermalgravimetric analysis – differential scanning calorimetry coupled with mass spectrometry (TGA-DSC-MS), and high temperature transposed temperature drop and oxide melt drop solution calorimetry. The bulk oxidation of UN (containing a small amount of α-UN 1.5+x ) in air was found to follow a step-wise process characterized by consecutive oxidative reactions UN-UN 1.5+x -UO 2 → UO 2 -UO 3 —N k → UO 3 —N k → UO 3 → U 3 O 8 . TGA results support that the UO 2 – U 2 N 3+x passivating layer delays the onset of rapid bulk oxidation of UN in air up to 662 K. Synchrotron X-ray diffraction (XRD) and extended X-ray absorption fine structure (EXAFS) analyses were performed to characterize UN and its final oxidized product. The standard enthalpy of formation (ΔH° f ) of UN was determined to be –144.4 ± 5.9 kJ/mol·atom, in good agreement with previously determined values from Pt encapsulation and bomb calorimetric experiments. Lastly, a negative linear correlation between ΔH° f and the N/U molar ratio was established based on the thermochemical data obtained in this work and previously reported enthalpies of formations of β-UN 1.5-x and α-UN 1.5+x .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Energetics of oxidation and formation of uranium monocarbide

To enable better implementation of uranium monocarbide (UC) as an advanced nuclear fuel for future high-temperature reactors, it is essential to have a thorough knowledge of its thermal and thermodynamic properties under reactor operational conditions. In this work, we studied thermal bulk oxidation of UC by simultaneous thermal analysis consisting of thermogravimetric analysis – differential scanning calorimetry coupled with evolved gas mass spectrometry (TGA-DSC-MS), and we examined the thermodynamic stability of UC using high temperature oxide melt drop solution calorimetry. Further, in air, our studied UC sample (which contains ~5 mol% UO 2 ) was found to undergo a step-wise thermal oxidation process consisting of consecutive oxidations and thermal decomposition reactions: 0.95UC·0.05UO 2 → UO 3 ·0.29(C x O y ) + 0.66CO 2 → UO 3 ·0.20(C x O y ) + 0.09CO 2 → UO 3 ·0.03(C x O y ) + 0.17CO 2 → U 3 O 8 + 0.03CO 2 + 0.166O 2 . DSC was further used to determine the enthalpies of reactions associated with this series of oxidation reactions. Synchrotron X-ray diffraction (XRD) and extended X-ray absorption spectroscopy (EXAFS) were performed to characterize both the long- and short-range structures of UC. The standard enthalpy of formation (ΔH° f ) of UC was determined to be –50.7 ± 10.8 kJ/mol·atom, in good agreement with previous values measured by bomb calorimetry. Lastly, the enthalpic landscape of U-C compounds, including UC, U 2 C 3 , and α-UC 1.94 , were established based on the enthalpy normalized per mole atom, which suggests that U-C phases are thermodynamically stable at lower C/U ratios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Diffusion study of uranium mononitride/zirconium carbide composite for space nuclear propulsion

Here, the next generation of space exploration will require extensive developments in rocket technology. Space nuclear propulsion is of interest due to its high fuel density and power, however, it also has high temperature and stability demands. This study examines a uranium nitride/zirconium carbide ceramic-ceramic particle composite as a potential fuel for these missions. Based on the constituent properties, this fuel composite is expected to be highly thermally efficient and resistant to the hot hydrogen propellant. One of the main concerns with this composite is unknown diffusion behavior between UN and ZrC over time. Diffusion couples and composites were used to determine how these constituents will behave when left in contact at high temperatures. UN and ZrC were found to have limited but observable diffusion at the phase boundary. The resulting phase is a UZr(CN) quaternary phase. UZr(CN) has been examined for other high temperature and gas nuclear reactors with preliminary success; however, there lacks sufficient data to fully qualify this phase. Based on the limited information available, the resultant quaternary phase could be deemed acceptable and even provide better adhesion for the fuel particles (UN) to the matrix (ZrC).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Determination of single crystal thermal expansion in Uranium-6wt%Niobium shape memory alloy using in-situ diffraction and modeling of textured polycrystalline samples

In-situ neutron diffraction and ex-situ dilatometry experiments were performed on a chemically banded, quenched uranium-6wt% niobium shape memory alloy to study the impact of deformation-(detwinning-) induced texture on the evolution of thermally induced strains (and associated stresses). Thermal heating and cooling cycles between 5K and 473K were performed in-situ with neutron diffraction, and the lattice strain evolution is reported for the observed monoclinic α" phase. Comparisons between the measured diffraction strains and macroscopic dilatometry experiments reveal relationships between micro- and macro-level thermal expansion. Softening of the texture during heating suggests that twin boundary motion can accommodate the large internal thermal strains which are approximately 2% greater than that typically observed during the 175K heating interval used to age more randomly oriented polycrystalline material. Assuming weak constraint of neighboring grains in randomly textured polycrystals, the single crystal thermal expansion tensor is extracted from measurements of lattice strains over the range 5K to 473K. Predictions of polycrystalline thermal expansion, based upon this single crystal thermal expansion tensor, are shown to compare favorably with bulk thermal expansion observations of the as quenched microstructure. However, such a lower-bound estimate is insufficient to explain all aspects of the behavior of the textured material, where the matrix is not isotropic. In conclusion, it is hypothesized that relaxation processes which occur within the quenched microstructure during heating are responsible for the distinct thermal expansion behavior observed during the initial heating cycle compared to cooling and subsequent cycling.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Glass formulation and lab-scale testing of glasses designed for in-can Melter and in-container Vitrification of high-assay low-enriched uranium aqueous polishing Raffinate waste

Glasses were designed for processing a nuclear waste from aqueous polishing of high-assay low-enriched uranium using either In-Can Melter (ICM) or GeoMelt® In-Container Vitrification™ (ICV) technologies, which operate at temperatures of T p ≤ 1100 °C and T p ≤ 1450 °C, respectively. Due to the different operating conditions, the melt and glass properties were optimized differently for each technology. Each glass was designed to optimize for maximum waste loading while simultaneously satisfying processing (e.g., crystallization, viscosity, and conductivity) and product quality (e.g., durability, hazard characteristic, and crystal content) constraints. Here, the raffinate waste contains high nitric acid (4 M) and low total solids (9 g·L -1 ) concentrations. Feed preparation processes were tested to facilitate concentration and nitrate destruction/removal while controlling redox of the melter feed. A successful feed process including sugar addition and spray-drying was performed to generate an adequate melter feed for ICV processing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Measuring thermal diffusivity and gap conductance in uranium nitride and Zircaloy relevant for microreactor applications

Heat transfer across nuclear fuels and structural interfaces is an important factor for evaluating the performance of nuclear power systems. Specifically, heat generated as nuclear fuel fissions must be transported through the cladding material and through the reactor to reach the steam turbine for power generation. As new microreactor designs emerge, maximizing the efficiency of this heat transfer process becomes crucial to make them commercially viable. This article examines thermal diffusivity and gap conductance in uranium nitride (UN) fuel and Zircaloy-4 (Zry4) cladding using light flash analysis (LFA). Thermal diffusivity measurements were made on monolithic UN pellets and Zry4 exposed to carbon at peak operating temperatures of microreactors and show that carbon ingress has a minimal effect on thermal diffusivity when compared with identical materials not exposed to carbon. Evaluation of gap conductance at the UN-Zry4 interface was done using one-dimensional two-layer thermal transport models as a function of applied pressure. Here the results show that increasing pressure on the UN-Zry4 interface leads to gains in gap conductance per unit area in fuel-cladding assemblies at microreactor operating temperatures. While many other variables are expected to influence UN-Zry4 interfacial gap conductance (e.g. contact surface roughness, porosity, localized heating, environmental gas pressure), the work offers a demonstration of using a conventional LFA apparatus to determine this parameter at elevated temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Chemical interaction and compatibility of uranium mononitride and alumina forming austenitic stainless steel

Uranium mononitride (UN) and alumina forming austenitic (AFA) stainless steel are a potential fuel-cladding combination for the lead-cooled fast reactor (LFR). Chemical compatibility between UN and AFA steel needs to be verified before implementation in a nuclear reactor. Diffusion couple experiments at 823 K and 1023 K were conducted for nonirradiated UN samples in contact with as-cast (no thermally grown Al 2 O 3 ) and preoxidized (with thermally grown Al 2 O 3 ) AFA for 500 and 1000 h in an inert environment. Preoxidized AFA exhibited little to no interaction with all UN samples tested at both 823 K and 1023 K, displaying the stability and capability of the Al 2 O 3 layer to prevent chemical interaction and inter-diffusion with UN. Chemical interaction occurs between UN and as-cast AFA. At 1023 K, an aluminum and nitrogen rich phase (likely AlN) formed along the interface of as-cast AFA and UN samples. At 823 K the AlN phase was not prominently observed due to the reduced diffusivity of aluminum through AFA. The aluminum and nitrogen-enriched phase was also observed in a high temperature pressure-assisted test sample of UN and as-cast AFA thermally treated at 1373 K. Finally, in UN samples doped with a low weight percent of UO 2 (< 3 wt%), AlN was not detected along the interface at either temperatures, and an Al 2 O 3 layer likely formed along the interface and prevented further chemical interaction between UN and as-cast AFA.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A systematic exploration of the hydrolysis products of the uranium trioxide polymorphs and their optical vibrational spectra

Study of the uranium trioxide (UO 3 )-water system is complex with inconclusive results and limited details in the literature. The UO 3 system is home to at least seven structural polymorphs and an amorphous phase. The proposed hydrolysis products of UO 3 are just as numerous, yet investigations of these alteration products are sporadic and generally antiquated, thus requiring systematic investigations. Recent developments in the understanding of UO 3 phase space, aided by improvements in analytical and computational techniques, necessitate more modern investigations into the uranyl hydroxide family and their naturally occurring mineral counterparts. We present findings from a systematic investigation of the products formed via hydrothermal reactions of common UO 3 polymorphs and discuss how the equatorial coordination of the uranyl/uranyl-like ions within the UO 3 precursors leads to differences in the optical vibrational spectra of the resulting hydrolysis products. The hypo-stochiometric nature of α-UO 2 (OH) 2 allows for the formation of multiple unique uranyl sites and a distortion of the unit cell to a lower symmetry. This study provides, for the first time, an analysis of β-UO 2 (OH) 2 using modern techniques and instrumentation (Raman/infrared spectroscopy and powder x-ray diffraction) and lays a foundation for future time-dependent investigations into the structural dependence of the hydrolysis kinetics of the UO 3 phases. In conclusion, given the prevalence of UO 3 at both ends of the nuclear fuel cycle, an understanding of its behavior with water has applications ranging from nuclear forensics to waste management and environmental transport.

Uranium trioxide↗
DOE OSTI2025

Molecular-dynamics study of diffusional creep in uranium mononitride

Uranium mononitride (UN) is a promising advanced nuclear fuel due to its high thermal conductivity and high fissile density. Yet, many aspects of its mechanical behavior and microstructural features are currently unknown. In this paper, molecular dynamics (MD) simulations are used to study UN's diffusional creep. Nanometer-sized polycrystals are used to simulate diffusional creep and to calculate an effective GB width. It is found that Nabarro-Herring creep is not dominant in the temperature range of 1700–2000 K and that the dominant diffusional creep mechanism is Coble creep with an activation energy of 2.28 ± 0.09 eV. A method is proposed to calculate the diffusional GB width and its temperature dependence in polycrystals. The effective GB width of UN is calculated as 2.69 ± 0.08 nm. This value fits very well with the prefactor of the phenomenological Coble creep formula. It is demonstrated that the most comprehensive thermal creep model for UN can be represented as the combination of our Coble creep model and the dislocation creep model proposed by Hayes et al.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Gamma-ray spectrum variations for surface measurements of uranium hexafluoride cylinders

Here, this work aims to characterize features that contribute to the spectral variations that appear in gamma-23 ray spectra of filled uranium hexafluoride (UF 6 ) cylinders. While the enrichment of the UF 6 in a 24 cylinder can greatly affect gamma-ray spectra at peaks such as the 185.7 keV peak and be used to 25 verify enrichment, many other characteristics of a filled UF 6 cylinder can affect the spectra. These 26 characteristics vary on a per-cylinder basis and can include: the time between when a cylinder was 27 filled and when it was measured; the previous fill history of a cylinder; how the cylinder was most 28 recently filled; the UF 6 geometric distribution within the cylinder; and the measurement location 29 along the cylinder. This work explores spectral variations seen in measurement data and uses 30 validated simulation models to find trends in these uncertainties. These trends are analyzed to observe 31 how each parameter can contribute to the overall uncertainty of UF 6 spectral measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Machine learning framework for predicting uranium enrichments from M400 CZT gamma spectra

A machine learning framework was developed for predicting uranium enrichments from M400 CZT gamma spectra. This framework leverages the availability of a large amount of measured M400 gamma spectra and uses a recently updated version of Gamma Detector Response and Analysis Software (GADRAS) for gamma spectrum analysis and generation. It also leverages the existing machine learning modules in Python for gamma spectrum data processing, curation, model training, benchmarking, and optimization of the deep machine learning models. The framework is used to develop a deep learning model to analyze gamma spectra from a set of U 3 O 8 samples with enrichments ranging from 0.31 to 93.17% and UF 6 cylinders with enrichments ranging from 0.2 to 4.95%, and the model performance is tested using a set of measured spectra and the respective declared enrichment values. Results show that the model can correctly classify 99.35% of the U 3 O 8 sample enrichments, and can predict the samples’ enrichments within an average absolute error of 0.099% (in percentage points of enrichment). For the UF 6 cylinders, the average absolute error was approximately 0.03%, with an accuracy of 98% in classifying discrete enrichment values of UF 6 samples. Finally, the results also show that the model has performed significantly better in terms of predicting enrichments in UF 6 cylinders based on measured gamma spectra than the GEM code, with a standard deviation (of the relative errors) of 2.23% (compared with the 11.51% value for the GEM code) based on results from a set of test data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Bridging the Transuranics with Uranium(IV) Sulfate Aqueous Species and Solid Phases

Isolating isomorphic compounds of tetravalent actinides (i.e., Th IV , U IV , Np IV , and Pu IV ) improve our understanding of the bonding behavior across the series, in addition to their relationship with tetravalent transition metals (Zr and Hf) and lanthanides (Ce). Similarities between these tetravalent metals are particularly illuminated in their hydrolysis and condensation behavior in aqueous systems, leading to polynuclear clusters typified by the hexamer [M IV 6 O 4 (OH) 4 ] 12+ building block. Prior studies have shown the predominance and coexistence of smaller species for Th IV (monomers, dimers, and hexamers) and larger species for U IV , Np IV , and Pu IV (including 38-mers and 70-mers). We show here that aqueous uranium(IV) sulfate also displays behavior similar to that of Th IV (and Zr IV ) in its isolated solidphase and solution speciation. Two single-crystal X-ray structures are described: a dihydroxide-bridged dimer (U 2 ) formulated as U 2 (OH) 2 (SO 4 ) 3 (H 2 O) 4 and a monomer-linked hexamer framework (U-U 6 ) as (U(H 2 O) 3.5 ) 2 U 6 O 4 (OH) 4 (SO 4 ) 10 (H 2 O) 9 . These structures are similar to those previously described for ThIV. Moreover, cocrystallization of monomer and dimer and of dimer and monomer–hexamer phases for both Th IV (prior) and U IV (current) indicates the coexistence of these species in solution. Because it was not possible to effectively study the sulfate-rich solutions via X-ray scattering from which U 2 and U-U 6 crystallized, we provide a parallel solution speciation study in low sulfate conditions, as a function of the pH. Raman spectroscopy, UV–vis spectroscopy, and small-angle X-ray scattering of these show decreasing sulfate binding, increased hydrolysis, increased species size, and increased complexity, with increasing pH. Furthermore, this study describes a bridge across the first half the actinide series, highlighting U IV similarities to Th IV , in addition to the previously known similarities to the transuranic elements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Synthesis of Trimethyltriazacyclohexane (Me 3 tach) Sandwich Complexes of Uranium, Neptunium, and Plutonium Triiodides: (Me 3 tach) 2 AnI 3

1,3,5-Trimethyl-1,3,5-triazacyclohexane (Me 3 tach) readily complexes uranium triiodide to form (Me 3 tach) 2 UI 3 . Further, the complex is soluble in THF and arenes and can function as a source of UI 3 to form organometallic U(III) complexes. When dissolved in pyridine (py), (Me3tach)2UI3 forms (Me 3 tach)UI 3 (py) 2 . A related complex with the larger 1,4,7-trimethyl-1,4,7-triazacyclononane (Me 3 tacn) ligand, namely (Me 3 tacn)UI 3 (THF), was synthesized for comparison. Since X-ray quality crystals of (Me 3 tach) 2 UI 3 can be synthesized in high yield even with small-scale reactions, the system is ideal for extension to transuranium elements. Accordingly, the neptunium and plutonium complexes (Me 3 tach) 2 NpI 3 and (Me 3 tach) 2 PuI 3 were synthesized in an analogous manner from NpI 3 (THF) 4 and PuI 3 (THF) 4 , respectively.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Linear versus Bent Uranium(II) Metallocenes—A Local Vibrational Mode Study

Uranium metallocenes have recently attracted attention driven by their use as catalysts in organometallic synthesis. In addition to bent U(IV) and U(III), an U(II) metallocene [(η 5 -C 5 i Pr 5 ) 2 U] was synthesized with an unusual linear Cp–U–Cp angle. Here, in this work, we investigated 22 U(II) metallocenes, (i) assessing the intrinsic strength of the U-ring interactions in these complexes with a novel bond strength measure based on our local vibrational mode analysis and (ii) systematically exploring what makes these U(II) metallocenes bent. We included relativistic effects through the NESCau Hamiltonian and complemented the local mode analysis with natural bonding orbital (NBO) and quantum theory of atoms in molecules (QTAIM) data. Our study led to the following results: (i) reduction of bulky U-ring ligand substituents does not lead to bent complexes for alkyl substituents ( i Pr and i Bu) in contrast to SiMe 3 ring substituents, which are all bent. (ii) The most bent complexes are [(η 5 -C 5 H 4 SiMe 3 ) 2 U] (130°) and [η 5 -P 5 H 5 ) 2 U] (143°). (iii) Linear complexes showed one hybridized NBO with s/d character, while bent structures were characterized by s/d/f mixing. (iv) We did not observe a correlation between the strength of the U-ring interaction and the amount of the ring-U-ring bend; the strongest interaction was found for [η 5 -Cp) 2 U] and the weakest for [η 5 -P 5 H 5 ) 2 U]. In conclusion, our results provide a foundation for the design of U(II) metallocenes with specific physicochemical properties and increased reactivity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Mapping the Electronic Structure of the Uranium(VI) Dinitride Molecule, UN 2

A combined anion photoelectron spectroscopic and relativistic electronic-structure computational study of the electronic structure of the UN 2 molecule is reported. Since the photoelectron spectrum of uranium dinitride’s negative ion, UN 2 - , (UN 2 plus an additional electron) directly reflects the electronic structure of neutral UN 2 , we have measured and relied upon the photoelectron spectrum of the UN 2 – anion as a means of mapping the electronic structure of neutral UN 2 . In addition to the electron affinity of the UN 2 ground state (the energy released when attaching an electron to UN 2 molecule), energy levels of the UN 2 excited states (electronic states with higher energies than the ground state) were well characterized by the close interplay between experiment and high-level theory. We found that both electron attachment and electronic excitation significantly bend the UN 2 molecule and elongate its U≡N bond. Implications for the activation of UN 2 are discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Electronic and Geometric Contributors to Hydrogen Binding in Uranium Oxide Grain Boundaries

Hydrogen induced corrosion of uranium, which leads to the formation of toxic and pyrophoric UH 3 , raises significant safety concerns for long-term storage of nuclear materials. Previous work suggests hydrogen diffuses through the grain boundaries (GBs) of the passivating oxide layer to initiate hydriding reactions. However, the atomistic mechanisms underlying this phenomenon and the structural factors that control its initiation are not well understood. To address this knowledge gap, here we use a high-throughput density functional theory (DFT) workflow to investigate the adsorption of H and H 2 in the defective bulk UO 2 . Specifically, we have exhaustively investigated the adsorption of H (107 sites) and H 2 (26 sites) in three different coincident site lattice (CSL) GBs: Σ3, Σ5, and Σ9. Compared to the binding energies in pristine UO 2 , we observe significantly stronger hydrogen adsorption at these GB sites. Interestingly, we find that the trends in H and H 2 adsorption vary considerably across the three GB models. In particular, while a small number of sites in Σ5 and Σ9 show exothermic adsorption of H and H 2 , respectively, no such sites are found in Σ3. These results provide fundamental atomistic insights that could guide the development of future corrosion mitigation strategies for the storage of nuclear materials.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

U–C Bond Insertion, Ring-Opening, and C–H Activation in a Uranium Bis(diisopropylamino)cyclopropenylidene (BAC) Adduct

Reaction of [U(NR 2 ) 3 ] (R = SiMe 3 ) with 1 equiv of bis(diisopropylamino)cyclopropenylidene (BAC) in Et 2 O results in the formation of [(NR 2 ) 3 U(BAC)] (1), which can be isolated in modest yields. Thermolysis of 1 in C 6 D 6 at 85 °C results in the formation of the ring-opened U(IV) product, [(NR 2 ) 2 U{N(R)(SiMe 2 CH=C(NiPr 2 )C(NiPr 2 )=CH)}] (2), which can be isolated in low yields. Mechanistic studies suggest that the formation of 2 proceeds via dissociation of BAC from 1 to regenerate [U(NR 2 ) 3 ], which converts into [U{N(R)(SiMe 2 CH 2 )}(NR 2 ) 2 ] at the elevated temperatures. BAC then inserts into the U–C bond of [U{N(R)(SiMe 2 CH 2 )}(NR 2 ) 2 ] to generate a cyclopropenyl intermediate, which undergoes ring opening and C–H activation to afford the final product, 2. We hypothesize that the ring-opening generates an unobserved carbene intermediate. Notably, thermolysis of a 1:1 mixture of independently prepared [U{N(R)(SiMe 2 CH 2 )}(NR 2 ) 2 ] and BAC results in clean formation of 2, providing strong support for the proposed mechanism. The formulations of both 1 and 2 were confirmed by X-ray crystallography. Finally, theoretical calculations indicate that the hypothesized uranium carbene intermediate features strong U–C bonding, potentially with some carbyne character in the electronic structure.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗