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Redox-switchable carboranes electrochemically capture and release UO 2 2+ selectively from mixed metal aqueous solutions, mimicking in part spent nuclear fuel.
Reductive silylation of the uranyl dication with 1,4-bis(trimethylsilyl)dihydropyrazine, or “Mashima's reagent”, is detailed. The reagent simultaneously delivers silylium ions and electrons to multiple uranyl complexes, i.e. a pyridine dipyrrolide uranyl complex and a common uranyl-containing starting material, UO 2 Cl 2 (OPPh 3 ) 2 . This reaction results in quantitative activation of thermodynamically robust U=O bonds. Furthermore, the reductive functionalization of environmentally persistent actinyl species with Mashima's reagent is seen as a promising solution for nuclear waste remediation.
For this work, twelve novel Ag + /UO 2 2+ heterometallic complexes have been prepared and characterized via structural, spectroscopic, and computational methods to probe the effects of Ag–oxo interactions on bonding and photophysical properties of the uranyl cation. Structural characterization reveals Ag–oxo interaction distances ranging from 2.475(3) Å to 4.287(4) Å. These interactions were probed using luminescence and Raman spectroscopy which displayed little effect on the luminescence intensity and the energy of the Raman active U═O symmetric stretch peak as compared to previously reported Pb–oxo interactions. Computational efforts via density functional theory-based natural bond orbital analysis revealed that the highest stabilization energy associated with the Ag–oxo interaction had a value of only 11.03 kcal mol –1 and that all other energy values fell at 7.05 kcal mol –1 or below indicating weaker interactions relative to those previously reported for Pb 2+ /UO 2 2+ heterometallic compounds. In contrast, quantum theory of atoms in molecules analysis of bond critical point electron density values indicated higher electron density in Ag–oxo interactions as compared to Pb–oxo interactions which suggests more covalent character with the Ag + . Overall, this data indicates that Ag + has a less significant effect on UO 2 2+ bonding and photophysical properties as compared to other Pb 2+ , likely due to the high polarizability of the cation.
Pertechnetate ( 99 Tc VII O 4 − ) and autoreduced 99 Tc(v), species coordinate to UO 2 2+ to form hybrid clusters. These clusters have important implications to understanding technetium chemistry in spent nuclear fuel and legacy wastes.
Macrocyclic uranyl complexes have been prepared that incorporate a range of secondary metal cations. Structural and spectroscopic studies have been used to study how the cations and macrocyclic structure influence the [UO 2 ] core moiety.
Uranium tetrafluoride microrods were prepared by chemical transformation from the reaction of UO 2 microrods with HF(g).
Radiolysis of water in high radiation fields generates a variety of reactive oxygen species that influence the chemical behavior and complexation of hexavalent uranium. This study investigates the behavior of interaction of a uranyl cation (UO 2 2+ (VI)) with a series of free radicals that are formed in situ via activation of the free radical initiator persulphate (S 2 O 8 2− ), which releases both SO 4 ˙ − and ˙OH species in the solution. Electron Paramagnetic Resonance (EPR) and Raman spectroscopy were used to evaluate the presence of the hydroperoxyl radical (HO 2 ˙) and superoxide radicals (O 2 ˙ − ) that are formed within the solution through radical cascade reactions. In addition, a uranyl peroxide cluster solid (NaU 24 ) was crystallized and characterized using single crystal X-ray diffraction (SCXRD), vibrational spectroscopy, and EPR spectroscopy. The presence of the hydroperoxyl radical (HO 2 ˙) and superoxide radicals (O 2 ˙ − ) was also observed in the solid-state compound, but spectroscopic evidence suggests that it was associated with the Na + network and not the cluster itself. Density functional theory (DFT) calculations were also utilized to further confirm the radical species produced and determine the potential stabilization of radicals detected within the cluster and lattice.
Elements of the 5f row maintain a wide variety of oxidation states that have been exploited in synthesis, catalysis, and separations. Herein, we describe the complexation of an O,N,O chelator, ExPh, with uranium in the (IV) and (VI) (uranyl) oxidation states. These two uranium complexes, U(IV)ExPh and UO 2 ExPh, respectively, were characterized in the solid state via single-crystal X-ray diffraction (SC-XRD) analysis. Electrochemical studies using cyclic voltammetry were employed to investigate the multiple redox events associated with the U(IV) complex. Spectro-electrochemical analysis of U(IV)ExPh provided spectroscopic evidence of a stable U(V) species. Chemical oxidation of U(IV)ExPh allowed isolation of the U(V) complex, U(V)ExPh. All three stable uranium complexes produced in this study were characterized via IR, UV-vis and NMR spectroscopies, and micro-spectrophotometry. On the other hand, efforts to reduce U(IV)ExPh to the corresponding U(III) species or produce this putative complex directly from ExPh failed to yield an isolable product. The stabilization of three formal oxidation states of uranium, coupled with previous lanthanide-row results, paves the way for studies of ExPh and its analogues in minor actinide chemistry.
Butyrobetaine bis(trifluoromethylsulfonyl)imide, ([Hbbet][Tf 2 N]), is a systematic variation of the task-specific ionic liquid betaine bis(trifluoromethylsulfonyl)imide, ([Hbet][Tf 2 N]). [Hbbet][Tf 2 N] exhibits improved properties with lower water solubility and a wider electrochemical window while maintaining thermomorphic phase switching behavior. We report the isolation and characterization of [UO 2 (bbet) 3 ][Tf 2 N] 2 with spectroscopy suggesting the presence of this species in solution.
Temperature gradients in ceramic light water reactor (LWR) uranium dioxide (UO 2 ) nuclear fuel pellets generate thermal stresses that cause fractures in the fuel beginning early in the life of fresh fuel. The combination of heating due to fission and forced convective cooling on the exterior of LWR fuel rods generates a temperature profile that is difficult to replicate outside the reactor environment. In the present study, a state-of-the-art experimental set-up using electrical heating to replicate fission heating was built and surrogate fuel materials such as ceria (CeO 2 ) were used to validate the system. Cracking experiments were conducted on these surrogates by inducing reactivity-initiated-accident (RIA) like temperature gradients in the pellets via induction and direct resistance heating. Induction heating was done using copper coils and molybdenum susceptors which heated the surrogates to a threshold temperature that is sufficiently high for the fuel material to conduct current. Thereafter, direct resistance heating was used by a D.C. power supply to introduce volumetric heating to replicate LWR operating conditions analogous to fission heating. The pellets were held against nickel electrodes and mounted on a boron nitride test-stand. All the tests were carried out in a stainless-steel vacuum chamber. Simultaneous real-time dual imaging of the surrogate pellet surface has been implemented using an optical and infrared camera system which will be mounted along axial and perpendicular directions to the pellet surface respectively. A beam-splitter was used to split the incoming radiation from the sample into two halves. While one of the beams is transmitted from the splitter through a bandpass filter to obtain optical images, the other beam is reflected from the splitter to the thermal camera to capture full field temperature gradients of the as fabricated pellet surface during crack initiation and propagation. In the current series of tests, a 2-color pyrometer was used for recording and comparing the surface and centerline temperatures of the surrogate pellets in lieu of the thermal camera. A LabVIEW data acquisition system has been set up for collecting useful data during experiments.
A simple model calculating the performance of an IMT solar reference cell is proposed and tested under clear sky periods. One-minute data from EKO spectroradiometers collocated with an IMT reference cell are used for the analysis. Data are obtained at the NREL Solar Radiation Research Laboratory in Golden, Colorado and the UO Solar Radiation Monitoring Laboratory in Eugene, Oregon. The model is also applied and evaluated under cloudy conditions. A match of better than 1% is found between calculated clear sky values and IMT measurements for July, 2020. One- minute cloudy sky comparisons show a much larger variation, but when averaged over one hour, a standard deviation of 4% between the calculated and measured values is found.
Here, U 3 Si 2 is a potential candidate for accident tolerant fuels because of its high uranium density and excellent thermal conductivity in comparison to UO 2 . However, U 3 Si 2 suffers from oxidation, steam corrosion, and subsequent disintegration/pulverization. The detailed investigation of kinetics that incorporates fundamental treatment of oxidation of U 3 Si 2 is scarcely reported, and the oxidation mechanisms have not been fully elucidated. In this paper, the oxidation behavior of microcrystalline (mc - ) and nanocrystalline (nc - ) U 3 Si 2 have been systematically investigated using a thermogravimetric analysis (TGA) apparatus through a series of isothermal and non-isothermal kinetic studies. The isothermal kinetic study with a model-fitting approach indicates oxidation activation energy of 85 kJ/mol for dense mc - U 3 Si 2 and 96.4 kJ/mol for nc - U 3 Si 2 pellets, while the isoconversional approach leads to an activation energy in the range of 70–85 kJ/mol for mc - U 3 Si 2 and 75–86 kJ/mol for nc - U 3 Si 2 with three most common model-free methods, including Kissinger–Akahira–Sunose, Flynn–Wall–Ozawa, and Friedman methods. The derivation of oxidation activation energies using both isothermal and isoconversional methods highlights the approach to evaluate the oxidation resistance of nuclear materials using TGA quantitatively and makes it possible to compare among various nuclear fuels.
Zirconium carbide (ZrC) is commonly used for energy sector research, as well as a surrogate for the proposed advanced nuclear fuel candidate uranium carbide. Here this study investigates structural modifications to nanocrystalline and microcrystalline ZrC resulting from dense electronic excitations induced by swift heavy ion exposure. Samples were irradiated with 946 MeV Au ions to various fluences up to 6 × 10 13 ions cm –2 and characterized using synchrotron-based x-ray diffraction. The evolution of the unit-cell parameter and heterogeneous microstrain were evaluated as a function of fluence and compared with those of nanocrystalline and microcrystalline CeO 2 (a surrogate for UO 2 fuel) irradiated under identical conditions. Distinct differences were observed in the radiation responses of the carbide and oxide across both grain sizes. Most notably, microcrystalline ZrC exhibits swelling characterized by two distinct regimes, which does not result in saturation at the ion fluences achieved. This contrasts with CeO 2 , which exhibits the well-documented direct-impact defect accumulation mechanism, reaching a steady-state saturation of swelling at higher fluences. Nanocrystalline CeO 2 undergoes more pronounced swelling compared with microcrystalline CeO 2 , in contrast to nanocrystalline ZrC, which exhibits only minimal unit-cell changes. These results demonstrate that swift heavy ion-induced structural changes can be quite different in carbides and oxides, which must be considered when extrapolating fission-fragment type damage in current fuels to advanced fuels.
In this work, a guided ion beam tandem mass spectrometer was used to study the reactions of U + with N 2 and NO. Reaction cross sections were measured over a wide range of energy for both systems. In each reaction, UN + is formed by an endothermic process, thereby enabling the direct measurement of the threshold energy and determination of the UN + bond dissociation energy. For the reaction of U + + N 2 , a threshold energy (E 0 ) of 4.02 ± 0.11 eV was measured, leading to D 0 (UN + ) = 5.73 ± 0.11 eV. The reaction of U + + NO yields UO + through an exothermic, barrierless process that proceeds with 94 ± 23% efficiency at the lowest energy. Analysis of the endothermic UN + cross section in this reaction provides E 0 = 0.72 ± 0.11 eV and, therefore, D 0 (UN + ) = 5.78 ± 0.11 eV. Averaging the values obtained from both reactions, we report D 0 (UN + ) = 5.76 ± 0.13 eV as our best value (uncertainty of two standard deviations). Combined with precise literature values for the ionization energies of U and UN, we also derive D 0 (UN) = 5.86 ± 0.13 eV. Both bond dissociation energies agree well with high-level theoretical treatments in the literature. The formation of UN + in reaction of U + with NO also exhibits a considerable increase in reaction probability above ~3 eV. Theory suggests that this may be consistent with the formation of UN + in excited quintet spin states, which we hypothesize are dynamically favored because the number of 5f electrons in reactants and products is conserved.
Ionic liquids (ILs) are a unique class of solvents with potential applications in advanced separation technologies relevant to the nuclear industry. ILs are salts with low melting points and a wide range of tunable physical properties, such as viscosity, hydrophobiciy, conductivity, and liquidus range. ILs have negligible vapor pressure, are often non-flammable, and can have high thermal stability and a wide electrochemical window, making them attractive for use in separations processes relevant to the nuclear industry. Metal salts generally have a low solubility in ILs; however, by incorporating new functional groups into the IL cation or anion that promote complexation with the metal, the solubility can be greatly increased. One such task-specific ionic liquid (TSIL) is 1-carboxy-N, N, N-trimethylglycine bis(trifluoromethylsulfonyl)imide ([Hbet][Tf 2 N]). Water, which is detrimental for electrochemical separations, is a common impurity in ILs and can coordinate with actinyl cations, particularly in ILs containing only weakly coordinating components. Understanding the behavior of actinides in TSIL/water mixtures on a molecular level is vital for designing improved separations processes. Classical molecular dynamics simulations of uranyl(VI) and plutonyl(VI) in 1-ethyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([EMIM][Tf 2 N]) with deprotonated Hbet (betaine) and water have been performed to understand the coordination and dynamics of the actinyl cations. We find that betaine is a much stronger ligand than water and prefers to coordinate the metal in a bidentate manner. Potential of mean force simulations yield a relative free energy for betaine coordination of approximately -120 to -90 kJ/mol in mixtures with water. As the amount of betaine coordinated to the actinide increases, the diffusion coefficient of the actinyl cation decreases. Moreover, the betaine ligand is able to bridge between two metal centers, resulting in dimeric complexes with actinide–actinide distances of ~5 Å. Potential of mean force simulations show that these structures are stable, with relative free energies of up to -40 kJ/mol. The crystal structure for [(UO 2 ) 2 (bet) 6 (H 2 O) 2 ][Tf 2 N] 4 shows that the betaine bridges between two uranium atoms to form dimeric complexes similar to those found in our simulations.
Accident tolerant fuel (ATF) is expected to delay or prevent core damage by providing additional coping time under accidents involving loss of core cooling. The effect of extended coping time may vary depending on the plant response to accidents. Age-related component degradation that deteriorates plant performance over time could have an impact on the actual advantages of ATF. The potential safety benefits of two near-term ATF candidates, including Cr-coated Zr cladding and FeCrAl cladding, are assessed for a 2-in. loss-of-coolant accident with failed high-pressure safety injection using the dynamic event tree (DET) approach considering possible stress corrosion cracking of steam generator (SG) tubing under aging. The DET approach allows likelihood quantification of accident sequences leading to core damage, including stochastic variation of system response and human actions during accident mitigation. The safety benefits of the selected ATF claddings in terms of additional coping time and the core damage frequency reduction rate under specified accident situations were quantitatively estimated. The results show that the deployment of the two selected ATF claddings is expected to lead to longer coping times and lower core damage frequency due to the wider safety margin to peak cladding temperature they provide. The safety advantages would be greater as SG tube degradation proceeds. Thus, the two ATF candidates would lead to less severe consequences in terms of likelihood of core damage and susceptibility to the SG tube degradation than UO 2 -Zr fuel.
Spent nuclear fuel repository simulations are currently not able to incorporate detailed fuel matrix degradation (FMD) process models due to their computational cost, especially when large numbers of waste packages breach. The current paper uses machine learning to develop artificial neural network and k-nearest neighbor regression surrogate models that approximate the detailed FMD process model while being computationally much faster to evaluate. Further, using fuel cask temperature, dose rate, and the environmental concentrations of CO 3 2- , O 2 , Fe 2+ , and H 2 as inputs, these surrogates show good agreement with the FMD process model predictions of the UO 2 degradation rate for conditions within the range of the training data. A demonstration in a full-scale shale repository reference case simulation shows that the incorporation of the surrogate models captures local and temporal environmental effects on fuel degradation rates while retaining good computational efficiency.
The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.