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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Framework Polymorphism and Modular Crystal Structures of Uranyl Vanadates of Divalent Cations: Synthesis and Characterization of M (UO 2 )(V 2 O 7 ) ( M = Ca, Sr) and Sr 3 (UO 2 )(V 2 O 7 ) 2

In this work, Uranyl vanadate compounds with divalent cations, M (UO 2 )(V 2 O 7 ) ( M = Ca, Sr) and Sr 3 (UO 2 )(V 2 O 7 ) 2 , were synthesized by flux crystal growth, and their crystal structures were solved using single-crystal X-ray diffraction data. Ca(UO 2 )V 2 O 7 and Sr(UO 2 )V 2 O 7 were synthesized from reactants with molar ratios M:U:V of 1:1:2 and identical heating conditions, and increasing the M:U:V ratio to 3:1:4 resulted in Sr 3 (UO 2 )(V 2 O 7 ) 2 . Crystallographic data for M(UO2)V2O7 compounds are: a = 7.1774(18) Å, b = 6.7753(17) Å, c = 8.308(2) Å; V = 404.01(18) Å 3 ; space group Pmn2 1 , Z = 2 for Ca; a = 13.4816(11) Å, b = 7.3218(6) Å, c = 8.4886(7) Å; V = 837.91(12) Å3; space group Pnma , Z = 4 for Sr. Compound Sr 3 (UO 2 )(V 2 O 7 ) 2 has a = 6.891(3) Å, b = 7.171(3) Å, c = 14.696(6) Å, α = 85.201(4)°, β = 78.003(4)°, γ = 89.188(4)°; V = 707.9(5) Å 3 ; space group P1, Z = 2. The framework structure of Sr(UO 2 )(V 2 O 7 ) is related to that of Pb(UO 2 )(V 2 O 7 ) reported previously, while that of Ca(UO 2 )(V 2 O 7 ) has a different topology. The topological polymorphism of the [(UO 2 )(V 2 O 7 )]-type framework may be due to the differing ionic radii of the guest M 2+ cations. Compound Sr 3 (UO 2 )(V 2 O 7 ) 2 has a modular structure based on two different types of electroneutral layers: [Sr(UO 2 )(V 2 O 7 )] and [Sr 2 (V 2 O 7 )]. This report calculated Structural complexities, and Raman spectra were collected and their peaks were assigned.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Changing the Rules of the Game: Used Fuel Studies Outside of a Remote Handling Facility

Recent advances in micro-fabrication tools like the FEI Helios NanoLab 660 dual Focused Ion Beam (FIB) and Scanning Electron Microscope (SEM) now make it possible to easily and reliably prepare micron (or smaller) specimens for analyses or further experimentation. Pacific Northwest National Laboratory (PNNL) has leveraged this capability at the Category II Nuclear Facility, the Radiochemical Processing Laboratory, to facilitate nuclear materials analysis and experimentation of used nuclear fuel. Micron-scale specimens of un-irradiated UO2 fuel pellets of various enrichments from depleted to 4.95% enriched, relative to 235U, were prepared at PNNL and shipped to the University of Missouri’s Research Reactor (MURR) for irradiation. Prior to shipping, samples were attached to commercially available Transmission Electron Microscopy half-grids via ion deposition, which were then housed in specially designed aluminum sample holders approved for use at the reactor facility. This minimized handling and preparation requirements prior to materials analysis after irradiation. Samples were irradiated to a burnup equivalent of 8 – 3,700 MWd/MTU. Additionally, by reducing the size of the specimens, and therefore their emitted radiation dose, the required cooling time after irradiation and prior to analysis was greatly diminished, and handling of these samples could proceed outside of a hot cell facility. Results from these experiments provided insight into the initial production of noble metal phase particles in used nuclear fuel at extremely low burnup levels. Some incompatibility with the container was observed at higher burnup. Micro-scale irradiation experiments were performed at roughly 1/7th of the time, effort and expense a similar bulk-scale study requiring a shielded hot cell facility would require. Conducting nuclear materials analysis at the micron-scale holds the promise of enhancing collaboration and material accessibility to the broader research community, improving accessibility to cutting edge research tools by the nuclear materials science community, lowering the risk to the scientist by reducing radiation dose and exposure, reducing facility and research costs associated with the requirement to conduct nuclear materials science within glovebox or remote handling facilities, and minimizing waste.

Schwantes, Jon M.↗
DOE OSTI2022

Direct isotopic analysis of solid uranium particulates on cotton swipes by microextraction-ICP-MS

Direct isotope ratio analysis of solid uranium particulates on cotton swipes was achieved using a solution-based microextraction technique, coupled to a quadrupole inductively coupled plasma - mass spectrometer (ICP-MS). This microextraction-ICP-MS methodology provides rapid isotopic analysis which could be applicable to nuclear safeguards measurements. Particulates of uranyl nitrate hexahydrate (UO 2 (NO 3 ) 2 ·6H 2 O) and uranyl fluoride (UO 2 F 2 ) ranging from 6 μm to 40 μm in length were transferred to cotton swipes with a particle manipulator. The microextraction probe then delivers a 5% nitric acid (HNO3) solvent onto the swipe surface to extract the uranium species. The extracted sample is then delivered to the ICP-MS for isotopic determination. The majority of uranium signal (~99% and ~94% for UO2(NO3)2·6H2O and UO 2 F 2 , respectively) was detected in the first 15 s extraction, while subsequent extractions on the same location had low or no U signal, suggesting near complete removal of the solid uranium compounds from the swipe surface. Ten samples (for each of the uranium compounds), were analyzed for their isotopic composition. For UO 2 (NO 3 ) 2 ·6H 2 O, the determined isotope ratios resulted in a % relative difference (% RD) from the referenced isotope ratios of 0.97, 1.0, and 7.3% for 234 U/ 238 U, 235 U/ 238 U, and 236 U/ 238 U, respectively. The % RD of the UO 2 F 2 isotope ratios were 1.9 and 0.60% for 234 U/ 238 U and 235 U/ 238 U, respectively. The preliminary limits of detection were determined to be 0.002, 0.4, and 60 pg for 234 U, 235 U and 238 U, respectively This work demonstrates that microextraction ICP-MS is a rapid and sensitive method that could directly determine uranium isotope ratios of UO 2 (NO 3 ) 2 ·6H 2 O and UO 2 F 2 particulates on cotton swipes.

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

A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide

During reactor operation, nuclear fuels are subject to extreme temperature and irradiation conditions which can significantly degrade the fuel's thermal transport properties. The reduction in thermal conductivity of the fuel as a result of irradiation-induced lattice defects is arguably the most important fuel performance metric in regard to reactor efficiency and safety. Because thorium dioxide (ThO2) is suitable as a model system for more complex materials such as UO2 and its mixed oxides, we present a theoretical investigation of thermal conductivity reduction seen in defect-bearing thorium dioxide and compare directly to experimental measurements. Phonon-mediated thermal transport of the fuel is modeled by a solution to the Boltzmann transport equation (BTE) for phonons. A cluster dynamics (CD) model for lattice defect evolution during irradiation predicts defect densities which are used as input to the BTE for modeling phonon-defect scatterings. Phonon scatterings by lattice defects include those from point defects and vacancy clusters and interstitial clusters of various sizes. The CD model is benchmarked against structural defect characterization of irradiated thorium dioxide using electron microscopy. Thermal conductivity predicted by the BTE model is compared to measured values for irradiated thorium dioxide specimens below room temperature to isolate effects of phonon-defect scattering from intrinsic 3-phonon processes, which dominate at higher temperatures. The computed conductivity values are in partial agreement at temperatures close to room temperature while slight deviations are observed at the lowest measured temperatures, suggesting that implemented phonon-defect scattering cross-section expressions may not be adequate for low temperatures. The presented work provides a necessary investigation of the influence of irradiation induced defects on fuel performance and represents a first step toward a full characterization of phonon mediated thermal transport in irradiated materials with complex defect microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Oxide structure of air-passivated U-6Nb alloy thin films

U-6Nb is a uranium alloy containing 6 wt% niobium that possesses high corrosion resistance. The structure and composition of the passivating oxide layer formed on air-aged U-6Nb, which gives the material its corrosion resistant properties, was characterized using surface scattering techniques. Stable oxide layers formed on the surface of a set of U-6Nb alloy thin films under ambient conditions were investigated using neutron reflectometry (NR), x-ray reflectometry (XRR) and grazing incidence x-ray diffraction (GIXD). The passivating oxide was composed of approximately 27% U, 5% Nb, and 68% O, primarily consisting of a thin niobium oxide layer (5.5 ± 0.4 nm) separating a thicker UO2 layer (27.1 ± 2.3 nm) from the underlying U-6Nb alloy. A critical density of enriched niobium oxide at the metal-oxide interface is hypothesized to limit oxygen diffusion and confer high corrosion resistance to the alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Controlled current-rate AC flash sintering of uranium dioxide

Uranium dioxide (UO2) pellets with controlled microstructures were densified up to 93.4% of their theoretical density in less than 25 minutes at a furnace temperature of 873 K, utilizing controlled current-rate alternating current (AC) flash sintering (FS). Using this AC-FS method it was possible to control the sintering rate and thermal gradients, resulting in dense pellets with no appreciable hourglassing and good mechanical integrity. Moreover, the apparent sintering activation energy for FS and for conventionally sintered samples was estimated to be 108 kJ mol -1 and 380 kJ mol -1 , respectively using the master sintering curve method. The apparent activation energy for FS was remarkably close to those reported in the literature for spark plasma sintering of UO 2 . Furthermore, both these field assisted sintering methods utilize fast heating rates and electrical effects that are likely enhancing the grain boundary diffusion mechanism. Finally controlled current-rate AC-FS has been demonstrated as a technological advancement, capable of producing ceramic nuclear fuels in a fraction of the conventional processing time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Characterization of high thermal conductivity fuel surrogates before and after ion irradiation

High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the system. One proposed approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives with UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling in the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO2 behavior. Additives of 10 vol. % Al 2 O 3 or SiC in the form of short fibers or platelets were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50 °C to 900 °C and confirmed the desirable increase compared to pure CeO 2 /ZrO 2 pellets. Samples were irradiated with 20 MeV Ni 6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300 °C to 700 °C. Post irradiation characterization revealed a good stability of the samples at low to medium doses with matrix lattice parameter swelling of < 0.14 % but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Analytical workflow dependence of experimental observables for uranium chemistries

In this work, we investigated how the sequencing of laboratory analytical methods used for chemical and morphological characterization influences analytical findings for particulate materials relevant to the nuclear fuel cycle, including UO2, U3O8, studtite (UO2O2·4H2O), and β-UO3, in the context of nuclear forensic analysis. Particles of each chemistry obtained from consistent production batches were exposed to Raman spectroscopy and scanning electron microscopy in varying orders to elucidate how the order in which the techniques are applied influences morphological and chemical observations as a function of particle size. The results indicate that particles from all four chemistries exposed to high-resolution electron imaging before Raman spectral analysis demonstrate optical vibrational spectral changes that reduce accurate interpretation of the underlying chemistry via Raman spectral analysis. We hypothesize that these changes are due to the thermal load of the electron beam imparted to the sample being unable to be dissipated by materials with poor thermal conduction properties. Results from this study will aid in determining best practices for forensic analysis procedures to reduce uncertainty in chemical determination of unknown particulate samples.

Manns, Rebecca [University of Nevada, Las Vegas]↗
DOE OSTI2022

Clinical 6 MV X-Ray Facility Photo-Neutron/Fission Interrogations with TMFD Sensors

Detection of several targets (U, Be, D2O) of interest to national security has been achieved at Purdue University’s 6 MV (~104 R/h) X-Ray clinical linear accelerator (CLINAC), from photoneutron/fission induced neutron signatures using a Centrifugally Tensioned Metastable Fluid Detector (CTMFD). Traditional neutron detection schemes were saturated and ineffective under the intense (to ~104 R/h) photon environment; however, the CTMFD sensor technology was capable of measuring the relatively insignificant (~10-12x lower intensity) neutron field produced from 6 MeV-end point X-Ray beam interactions with the CLINAC components and surroundings. This is in stark contrast to published results based on conventional belief for immeasurable neutron dose in 6 MV CLINAC facilities. Photoneutron and photofission neutron production rates were detected and analyzed at various standoffs from the isocenter, with and without special nuclear material (SNM) related targets ranging in mass from 5 g of Be, to ~0.6 kg of UO2. It was confirmed that 6 MV X-Ray photons do result in measurable neutron dose above normal cosmic background. Additionally, in threshold mode, the CTMFD operating under a tensioned metastable pressure of -5bar, enabled conclusive rejection of background photoneutrons, while enabling the rapid detection of: Be, D2O, C and U targets within seconds to minutes while operating in and around 104 R/h photon fields.

Ozerov, Stepan↗
DOE OSTI2020

Transformational Challenge Reactor preconceptual core design studies

In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. Furthermore, the designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability. Additionally, the TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Transformational challenge reactor analysis to inform preconceptual core design decisions: Sensitivity study of transient analysis in a hydride-moderated microreactor

The Transformational Challenge Reactor (TCR) program aims to demonstrate a revolutionary design approach enabled by advanced manufacturing and data analytics in the design of nuclear reactors. This article discusses scoping analyses of preconceptual designs to inform TCR design decisions and the evaluation of sensitivities and uncertainties on postulated transient scenarios. The applicability of the systems codes TRACE and RELAP5-3D to TCR transient analysis are examined, and RELAP5-3D models are used to examine the transient response of two candidate core designs at multiple power levels. Then, the uncertainty quantification code RAVEN is used to quantify the effect of several design parameters on reactivity-initiated accident (RIA) progression at hot zero power (HZP) and hot full power (HFP) as well as to assess the impact of uncertainties in transient parameters for the pressurized loss of forced cooling (PLOFC).When results were compared, TRACE and RELAP5-3D showed good agreement in their ability to predict system behavior, but RELAP5-3D calculations were closer to analytical predictions for the RIA. Furthermore, models for a PLOFC accident in two designs (a UO2 and TRISO core) at multiple power levels showed greater temperature margins for the TRISO core at all power levels. Using this information, along with other scoping analyses and constraints, the TCR design team selected a power level of 3 MWth and a TRISO-based core design. For this design RAVEN was applied to vary RIA parameters in RELAP5-3D models at HZP and HFP to understand the effect on figures of merit such as peak power, fuel and coolant temperature, and energy deposition. This sensitivity study found that the inserted reactivity worth was the most important parameter controlling all figures of merit, but for insertion up to 1.5$ no failure of TCR fuel is anticipated. For constant reactivity insertions, the magnitude of the fuel temperature coefficient was found to have the greatest effect on all figures of merit under most circumstances. These results not only demonstrate the anticipated robust safety of the proposed TCR fuel form but also provide a reference for future metal-hydride moderated systems to understand RIA behavior. In the PLOFC, the impact of heat transfer enhancement due to wavy flow channel effects dominated the variance in peak temperatures, and variations in heat exchanger elevation provided the greatest control on natural circulation flow rate. Additionally, no fuel particle failure is anticipated in the PLOFC.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Destruction of the Uranyl Moiety in a U(V) “Cation–Cation” Interaction

A gas-phase uranyl peroxide dimer supported by three 12-crown-4 ether (12C4) ligands, [(UO 2 ) 2 (O 2 )(12C4) 3 )] 2+ (A), was prepared by electrospray ionization. Density functional theory (DFT) indicates a structure with two terminal 12C4 and the third 12C4 bridging the uranium centers. Collision induced dissociation (CID) of A resulted in elimination of the bridging 12C4 to yield a uranyl peroxide dimer with two terminal donor ligands, [(12C4)(UO 2 )(O 2 )(UO 2 )(12C4)] 2+ (B). Remarkably, CID of B resulted in elimination of the bridging peroxide concomitant with reduction of U(VI) to U(V) in C, [(12C4)(UO 2 )(UO 2 )(12C4)] 2+ . DFT studies indicate that in C there is direct interaction between the two UO2+ species, which can thus be considered as a so-called cation-cation interaction (CCI). This formal CCI, induced by tetradentate 12C4 ligands, corresponds to destruction of the linear uranyl moieties and creation of bridging U-O-U oxo-bonds. Finally, on the basis of the structural rearrangement to achieve the structurally extreme CCI interaction, it is predicted also to be accessible for PaO 2 + but is less feasible for transuranic actinyls.

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

Rational Construction of Porous Metal–Organic Frameworks for Uranium(VI) Extraction: The Strong Periodic Tendency with a Metal Node

Although metal-organic frameworks (MOFs) have been reported as important porous materials for the potential utility in metal ion separation, coordinating the functionality, structure, and component of MOFs remains a great challenge. Herein, a series of anionic rare earth MOFs (RE-MOFs) were synthesized via a solvothermal template reaction and for the first time explored for uranium(VI) capture from an acidic medium. The unusually high extraction capacity of UO2 2 + (e.g., 538 mg U per g of Y-MOF) was achieved through ion-exchange with the concomitant release of Me 2 NH 2 + , during which the uranium(VI) extraction in the series of isostructural RE-MOFs was found to be highly sensitive to the ionic radii of the metal nodes. That is, the uranium(VI) adsorption capacities continuously increased as the ionic radii decreased. Finally, in-depth mechanism insight was obtained from molecular dynamics simulations, suggesting that both the accessible pore volume of the MOFs and hydrogen-bonding interactions contribute to the strong periodic tendency of uranium(VI) extraction.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Structural and Optical Properties of Phase-Pure UO 2 , α-U 3 O 8 , and α-UO 3 Epitaxial Thin Films Grown by Pulsed Laser Deposition

Fundamental understanding of the electronic, chemical, and structural properties of uranium oxides requires the synthesis of high-crystalline-quality epitaxial films of different polymorphs of one material or different phases with various oxygen valence states. Here we report the growth of single-phase epitaxial UO 2 , α-U 3 O 8 , and α-UO 3 thin films using pulsed laser deposition. Both oxygen partial pressure and substrate temperature play critical roles in determining the crystal structure of the uranium oxide films. X-ray diffraction and Raman spectroscopy demonstrate that the films are single phase with excellent crystallinity and epitaxially grown on a variety of substrates. Chemical valance states and optical properties of epitaxial uranium oxide films are studied by X-ray photoelectron spectroscopy and UV–vis spectroscopy, which further confirm the high-quality stoichiometric phase-pure uranium oxide thin films. Epitaxial UO2 films show a direct band gap of 2.61 eV, while epitaxial α-UO 2 , α-U 3 O 8 and α-UO 3 films exhibit indirect band gaps of 1.89 and 2.26 eV, respectively. The ability to grow high-quality epitaxy actinide oxide thin films and to access their different phases and polymorphous will have significant benefits to the future applications in nuclear science and technology.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Multifunctional two-dimensional metal-organic frameworks for radionuclide sequestration and detection

Two lanthanide containing porous coordination polymers, [Ln2(bpdc)6(phen)2] · nH2O (1) and [Ln2(bpdc)6(terpy)2] · 3H2O (2) (Ln = either Pr, Nd or Sm-Dy); bpdc: 2,2'-bipyridine-5,5'-dicarboxylic acid, phen: 1,10-phenanthroline, and terpy: 2,2':6',2''-terpyridine, have been hydrothermally synthesized and structurally characterized by powder and single crystal X-ray diffraction. The crystallographic analyses reveal compounds 1 and 2 feature Ln3+ containing dimeric nodes that form a porous two-dimensional (2D) and non-porous three-dimensional (3D) framework, respectively. Each material is stable in aqueous media between pH 3-10 and exhibits modest thermal stability to ~400ºC. Notably, a portion of the phen and bpdc ligands in 1 can be removed thermally, without compromising the crystal structure, causing the surface area and pore volume to increase. The optical properties of 1 and 2 with Gd3+, Sm3+, Tb3+, and Eu3+ were explored in the solid-state using absorbance, fluorescence, and lifetime spectroscopies. The analyses reveal a complex blend of metal and ligand emission in the materials containing Sm3+ and Tb3+, while those featuring Eu3+ are dominated by intense metal-based emission. Compound 1 with Eu3+ displayed promise for the capture and detection of the uranyl cation, (UO2)2+, from aqueous media. In short, uranyl capture was observed at pH 4 and the adsorption thereof was detectable via vibrational and fluorescence spectroscopies and colorimetrically as the off-white color of 1 turned yellow with uptake. Finally, both 1 and 2 with Eu3+ produce bright red emission upon irradiation with Cu ka X-ray radiation (8.04 keV) and are candidate materials for applications in solid-state scintillation.

Surbella, Robert G.↗
DOE OSTI2019

Actinyl cation–cation interactions in the gas phase: an accurate thermochemical study

Gas phase actinyl cation–cation interactions (CCIs) were studied by an accurate composite coupled cluster thermochemical approach for the first time. A number of CCI dimers were constructed from the monomers UO 2 2+ , UO 2 + , NpO 2 2+ , NpO 2 + , PuO 2 + , and AmO 2 + . All CCI dimers studied were calculated to be thermodynamically unstable, with dissociation energies ranging from -60 to -90 kcal mol -1 , but in many cases kinetic stability was indicated by calculated local minima with well depths as large as ~15 kcal mol-1. Most of the dimers studied involved a T-shaped geometry, although one side-on dimer, (UO2+) 2 , was included since it was amenable to coupled cluster methods. In the T-shaped isomers the most stable dimers were calculated to arise when the oxo-group of an An(V) actinyl cation was oriented towards the metal center of an An(VI) actinyl cation. For both mixed-valent An(VI)/An(V) and mono-valent An(V) dimers, the stability as estimated from the depth of the calculated local minimum decreased in the donor series U(V) > Np(V) > Pu(V) > Am(V). These trends correlate well with experimental trends in condensed phase CCIs. A rationale for the bonding in CCIs was investigated by carrying out charge transfer analyses using the natural bond orbital (NBO) method. Augmenting the usual Lewis acid-base explanation, CCIs are the direct result of a competition between charge transfer stabilization, which can be as much as 0.11e or 30.7 kcal mol -1 at equilibrium, and Coulombic repulsive destabilization.

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

Actinide oxide dissolution in tributyl phosphate

An alternative to dissolving used nuclear fuel (UNF) in an acidic solution during reprocessing is direct dissolution in an organic solution, which would eliminate an aqueous dissolution step, decrease the amount of nitrate needed, and reduce the facility size. The flowsheet for this potentially less expensive alternative is first to voloxidize the UNF to remove fission product gases and form an oxide. After voloxidation, the UNF is then dissolved in an organic solution containing an extractant mixed with an aliphatic diluent and pre-equilibrated with nitric acid. The organic solution then goes through a solvent extraction process to recover the uranium and/or other desired radionuclides. This work qualitatively studied the dissolution of actinide oxides (UO2, NpO2, and PuO2) in tributyl phosphate using UV-Vis-NIR absorbance spectroscopy to ascertain dissolution behavior. Initial studies included material that is otherwise difficult to dissolve in only nitric acid, specifically CeO2, that is sometimes used as a dissolution surrogate for PuO2. This work confirmed that CeO2, NpO2, and PuO2 are difficult to dissolve in 30 vol% TBP-dodecane pre-equilibrated with 10 M HNO3 and will readily dissolve when co-precipitated with U (i.e., the mixed oxides U-Ce, U-Np, and U-Pu), surrogates for voloxidized nuclear fuel.

Gogolski, Jarrod [Savannah River National Laborato↗
DOE OSTI2026

The 0E2 benchmarks for PWR UO 2 decay heat: an analysis from the NEA WPNCS

This paper presents the work performed in the subgroup 16 of the Working Party for Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. The main goal was to define two decay heat benchmarks for Spent Nuclear Fuel (one pincell and one assembly), perform calculations and compare and analyze the results in light of existing calorimetric measurements. The selected case is the PWR UO2 assembly 0E2, irradiated at the Ringhals-3 reactor and measured at the Clab facility in Sweden. In total, 21 institutes worldwide participated to the exercise, leading to 55 calculated results (named C). It was found that the measured decay heat values (E) can be satisfactorily reproduced with two-dimensional assembly calculations, leading to an average C/E value of 0.99, with an uncertainty (or one standard deviation) of ±0.01.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗