Search NASA⌕ Search

SEARCH · Search NASA

Results for “URANIUM”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 217 records · Page 12

Isocyanide adducts of tri- and tetravalent uranium metallocenes supported by tetra(isopropyl)cyclopentadienyl ligands

In this work, reaction of the uranium( III ) metallocenium salt [(Cp iPr4 ) 2 U][B(C 6 F 5 ) 4 ] with tert-butyl isocyanide ( t BuNC) yielded the dicationic uranium( IV ) complex [(Cp iPr4 ) 2 U(CN t Bu) 4 ][B(C 6 F 5 ) 4 ] 2 (1), which displays a linear metallocene geometry. Use of crude mixtures of [(Cp iPr4 ) 2 U][B(C 6 F 5 ) 4 ], which contain a soluble source of iodide, led instead to isolation of the monocationic uranium( IV ) iodide complex [(Cp iPr4 ) 2 U(I)(CN t Bu) 2 ][B(C 6 F 5 ) 4 ] (2). Adduct formation with no change in oxidation state was observed upon addition of t BuNC to the neutral uranium( III ) species (Cp iPr4 ) 2 UI, resulting in isolation of (Cp iPr4 ) 2 U(I)(CN t Bu) (3). X-ray crystallographic and IR spectroscopic studies both showed effects ascribed to the presence of multiple strongly donating isocyanide ligands in 1.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Multiple oxidation states of uranium stabilized by an O , N , O -ligand

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.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗

High Flux Isotope Reactor Low-Enriched Uranium Low Density Silicide Fuel Design Parameters

High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) to low-enriched uranium (LEU) conversion activities are ongoing as part of the Department of Energy (DOE) National Nuclear Security Administration (NNSA)’s nuclear nonproliferation mission. Design activities studying the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a low density uranium-silicide dispersion (U 3 Si 2 -Al) fuel, which has a uranium density of 4.8 gU/cm 3 . The goal of these studies is to generate several HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. The documented designs will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U 3 Si 2 -Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U 3 Si 2 -Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Immobilization of Uranium in a Wetland 8-km Downstream from an Industrial Source (Report)

A three-day walk-over survey with portable gamma detecting equipment was conducted over a uranium-contaminated wetland located 8-km downstream from a former fuel fabrication facility. 97,000 spectra were collected, individually processed to extract uranium activity data, and then integrated with topographical information (LIDAR) to create maps. These maps reveal that 50 years after being released into the environment, 11% of the uranium is retained in the 9-ha wetland area. Combined with measurement collected closer to the source, it was determined that 94% of the released uranium is presently being held in downstream wetlands. This demonstrates that wetlands can be very effective at immobilizing contaminants for long durations. The survey also showed that previous flyover maps overestimated the size of the contaminated areas by 140% but had the advantage of being completed in orders-of-magnitude fewer labor hours. Depending on project needs, the walk-over technique may be useful to include in tandem with flyover measurements to acquire maps with greater sensitivity and spatial resolution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

High Flux Isotope Reactor Low Enriched Uranium U-10Mo Fuel Design Parameters

Activities to convert the HFIR from HEU to LEU are ongoing as part of the US Department of Energy (DOE) National Nuclear Security Administration (NNSA) nuclear nonproliferation mission. Design activities to study the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a uranium-molybdenum (U-10Mo) monolithic alloy fuel. This high-density alloy contains 90 wt % uranium and 10 wt % molybdenum and has a uranium density of 15.318gU/cm 3 . The goal of these studies is to generate several candidate HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. Recent advancements in modeling and simulation tools and design methods enabled a thorough analysis of the available design space with U-10Mo fuel. A surrogate model used this analysis as training data to quickly determine the performance of a design given specific design parameters. An optimization module used this surrogate model to quickly search this multidimensional search space given specific desired performance characteristics. This approach was made possible by the large available design space with U-10Mo fuel. Shift, a Monte Carlo tool optimized for high-performance computing (HPC) architectures, was used for faster calculation and better data management for reactor physics simulations. Once most of these design studies were complete, a new suite called the Python HFIR Analysis and Measurement Engine (PHAME) was developed to connect all fuel design analysis steps, making design studies more efficient and reproducible. The post-processing capabilities of these new tools are leveraged for the information provided herein. Leveraging these tools, several candidate fuel designs were selected with varying levels of feature complexity and reactor performance. This report provides design feature details for four selected HFIR LEU U-10Mo fuel designs and their corresponding performance and safety metrics. Nominal best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities, are provided. Simulations show that the high uranium density of U-10Mo fuel provides a large potential design space that enables various LEU designs to meet HEU core performance metrics and safety requirements with a power increase from 85 MW (HEU) to 95 MW or 100 MW (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluation of Thermal Neutron Scattering Cross Section of Uranium Silicide with Ab Initio Lattice Dynamics

Uranium silicide (U 3 Si 2 ) is a candidate material for the high-density nuclear fuel in commercial light water reactors [1], [2]. Its higher uranium density, 11.3 g-U/cm3, compared to that of uranium dioxide (UO 2 ), 9.7 g-U/cm3, can improve the performance of a nuclear reactor while using low enriched uranium (LEU) and diversify the choice of cladding materials [1]–[3]. It also has a higher thermal conductivity than UO 2 , which can reduce the thermal stress on the material caused by a temperature gradient across the fuel pellet and provide a larger margin for some postulated accidents [1], [2], [4]. Furthermore, compared to U3Si, another high-density fuel candidate, it has better resistance to in-pile swelling due to less irradiation-induced rapid amorphization [1], [3]. Corresponding to its importance in nuclear engineering, many previous studies have reported the properties of U3Si2. Experiments showed that U 3 Si 2 is a paramagnetic (PM) metal, where a slight linear increase in magnetic susceptibility was measured with increasing temperature [5], [6]. In addition, thermodynamic quantities such as thermal expansion coefficient, heat capacity, and thermal conductivity were experimentally determined over a wide temperature range [1], [7], [8]. In several computational studies, ab initio atomistic simulations based on density functional theory (DFT) were performed to calculate various properties including elastic constants, electronic density of states (DOS), and phonon dispersion curves [9]–[12]. Nevertheless, thermal neutron scattering cross sections, which are critical to the prediction of the parameters in reactor physics that are ultimately related to reactor criticality, have not yet been evaluated for U3Si2. The scattering cross section can be calculated from the phonon DOS, or the energy spectrum of lattice vibrations, of the crystalline system [13], [14]. However, there is also no experimental data available for the phonon DOS of U 3 Si 2 . While some computational studies reported the phonon DOS and/or dispersion curves from ab initio simulations [9]–[12], the accuracy cannot be guaranteed because it is unclear whether the spin-polarization behavior of PM U 3 Si 2 was properly described. In the present study, the thermal neutron scattering cross section for U 3 Si 2 is evaluated for the first time by calculating the phonon DOS for U3Si2 from ab initio lattice dynamics (AILD) simulations based on DFT. First, U 3 Si 2 is modeled based on the experimental structure, and AILD simulations are performed on the modeled U3Si2 to optimize the structure. Next, AILD simulations are performed for supercells with atomic displacement to calculate Hellmann-Feynman forces. Based on the calculated forces, partial phonon DOSs for U and Si are obtained, and the thermal neutron scattering law (TSL) for U 3 Si 2 is finally evaluated. To verify the accuracy of the calculations in the present study, the calculation results are compared with experimental data on the structure and heat capacity of U3Si2 [1], [7], [8], [15].

Geometry Optimization↗

Computational study of tungsten and depleted uranium photoneutron targets for a 20 MeV electron linear accelerator

Neutron production can be realized with a high energy electron linear accelerator by using Bremsstrahlung and photoneutron converters. In this study, Monte Carlo N-Particle Code (MCNP) was used to evaluate potential photonuclear target designs for a high energy electron linear accelerator for applications such as neutron radiography and neutron resonance spectroscopy. A computational model was developed to inform a target design that would yield a high number of neutrons. It consists of a 20 MeV electron beam incident on a Bremsstrahlung target and a photonuclear target to generate neutrons. This computational model showed that a thickness of 0.75 inches for both tungsten and depleted uranium yields the most neutrons from photoneutron reactions. Saturation in the total number of generated neutrons was observed at over 0.75-inch thickness for both evaluated materials. Depleted uranium yielded approximately twice the number of neutrons overall compared to tungsten. The highest neutron surface flux for Depleted Uranium was 1.06 × 10-4 neutrons/cm2/source electron, and for Tungsten it was 5.12 × 10-5 neutrons/cm2/source electron. The optimal target design for this study’s application would consist of a 0.75 inch-thick block of depleted uranium with the length, width, and/or diameter varying dependent on application.

43 PARTICLE ACCELERATORS↗

Radiation Dose Modeling for Niowave’s Accelerator Driven Uranium Target Assembly 3

Molybdenum-99 is a high-value radionuclide commonly used for medical purposes within the United States. The National Nuclear Security Administration (NNSA) seeks to reliably produce the radioisotope 99 Mo without the use of highly enriched uranium. NNSA’s Office of Material Management and Minimization (M3) provides funding and government laboratory expertise to private companies to expedite the production process domestically and currently funds designs that use low-enriched uranium or other 99 Mo production pathways. Several production designs are being explored across the industry, including uranium fission and photonuclear conversion of 100 Mo targets. Niowave Inc. seeks to produce 99 Mo via a high-energy electron accelerator that strikes a lead-bismuth eutectic target that ultimately produces a consistent neutron flux. The neutron flux then interacts in a subcritical reactor core configuration to produce fission in low-enriched or natural uranium targets. These fissionable targets are then processed to extract 99 Mo. The purpose of this work is to estimate the neutron and photon dose response across Niowave’s proposed facility for worker safety during operation. Owing to the size of the proposed Niowave facility and necessary shielding, unbiased Monte Carlo radiation transport is impractical, and variance reduction methods are required. This work focuses on the weight window variance reduction method to produce high confidence dose response results within a Monte Carlo radiation transport code. Specifically, an adjoint-informed weight window methodology was created to improve the dose response estimates for accelerator-driven subcritical reactor designs. This adjoint-informed methodology was implemented for Niowave’s proposed design and improved dose results at far-field locations across the facility. Acceptable dose rate contours for the proposed facility were generated across the facility and are presented in this work.

07 ISOTOPE AND RADIATION SOURCES↗

Speciation Correlations between Uranium and Iodine

The National Nuclear Security Administration’s Office of Conversion has been supporting development of molybdenum-99 production methods utilizing non-highly enriched uranium targets. One of the novel methods being considered is the use of an aqueous homogeneous subcritical system. Concerns about controlling potentially volatile radioiodine have led to efforts to characterize the iodine species in this system. This report examined the potential to characterize the chemical species of the iodine by correlating it with analyses of uranium, which is more ubiquitous in the target material and suitable for on-line monitoring. Direct measurement of the iodine species would be infeasible due to the difficulty in performing analyses. The use of chemometric modeling utilizing spectral analyses is viable for the uranium measurements, however, a better understanding of the radiolytic effects of neutrons and the corresponding relationship between uranium and iodine is required to correlate any measurements to iodine speciation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Development of a Chlorine Cathode and Anode Basket Assembly for Production of Uranium Chloride

A chlorine cathode has been developed for in situ chlorination of metals, oxides, and oxychlorides in molten chloride electrolytes that could be used to support synthesis of chloride fuel salts for molten salt reactors. The chlorine cathode is designed to electrochemically reduce chlorine gas to generate chloride ions that, when paired with a metal anode, chlorinate that metal as it is oxidized into the salt. The designed porous carbon electrode effectively distributes Cl 2 to the electrode surface and efficiently generates Clions in the molten chloride electrolyte. This report highlights recent improvements made to the electrode assembly, with a focus on operational control of the anode basket stability. Synthesis tests demonstrated the successful chlorination of uranium metal, resulting in 3.6 wt% uranium generated in the LiCl-KCl base salt in 45 minutes, performed in a bench-scale chlorination apparatus. Higher concentrations could be achieved by applying longer chlorination times or increasing the amount of uranium loaded in the anode basket. Overall, the chlorine cathode can be used with an appropriately designed anode to chlorinate uranium in situ for synthesis of molten salt reactor fuel salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radiological Characterization of Uranium Decay Products in LEU U-10Mo Before and After Heat Treatment

Fuel for the U.S. high-performance research reactor fleet is undergoing significant development as the United States moves away from using highly enriched uranium dispersion fuels. The proposed fuel is a high-density, low-enriched uranium (LEU, 19.75 wt% U 235) alloyed with ten weight percent molybdenum fuel (known as LEU U 10Mo). The LEU U 10Mo fuel samples studied in this report were made by down-blending highly enriched uranium with a master alloy, which is cast from molybdenum rods and depleted uranium, via vacuum induction melting and then casting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Mechanical and Thermal Properties for Uranium and U–6Nb Alloy from First-Principles Theory

Elasticity, lattice dynamics, and thermal expansion for uranium and U–6Nb alloy (elastic moduli) are calculated from density functional theory that is extended to include orbital polarization (DFT+OP). Introducing 12.5 at.% of niobium, substitutionally, in uranium softens all the cii elastic moduli, resulting in a significantly softer shear modulus (G). Combined with a nearly invariant bulk modulus (B), the quotient B/G increases dramatically for U–6Nb, suggesting a more ductile material. Lattice dynamics from a harmonic model coupled with a DFT+OP electronic structure is applied for α uranium, and the obtained phonon density of states compares well with inelastic neutron-scattering measurements. The Debye temperature associated with the lattice dynamics falls within the range of experimentally observed Debye temperatures and it also validates our quasi-harmonic (QH) phonon model. The QH Debye–Grüneisen phonon method is combined with a DFT+OP electronic structure and used to explore the anisotropic thermal expansion in α uranium. The anomalous negative thermal expansion (contraction) of the b lattice parameter of the α-phase orthorhombic cell is relatively well reproduced from a free-energy model consisting of QH-phonon and DFT+OP electronic structure contributions.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Update on the High Precision Titration Method for Uranium Assay Supported by NBL Program Office

The NBL Program Office (NBL PO) has coordinated with Oak Ridge National Laboratory (ORNL) to implement the High Precision Titration (HPT) method for uranium assay measurements. The measurement method has been successfully developed and qualified, therefore ORNL has progressed to performing analyzes critical to the mission of the NBL PO. The HPT method is vital to producing the next generation of certified reference materials for uranium assay and isotopic abundance. The work described here focused on establishing traceability of the method to the NIST SRM 136 potassium dichromate series, and focused on investigating small differences between SRM 136e and 136f. The NBL PO evaluated the history of all of its primary uranium reference materials, including CRM 112A natural uranium metal, and worked with NIST in investigating the small differences in the dichromate SRM’s. ORNL performed comparative experiments on SRM 136e and 136f using CRM 112A. The results of the experiment will be presented here, along with NBL PO’s plans to re-evaluate the CRM 112A certified uncertainty to comply with JCGM 100, “Guide to the expression of uncertainty in measurement.”

Rogers, Kayron↗

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Nuclear forensic analysis on uranium ore concentrates: A multi-laboratory intercomparison study

A collaboration between the United States and Japan performed nuclear forensic analyses on a set of 7 uranium ore concentrate (UOC) samples with known origins to improve methodologies and best practices in the field. Overall, good agreement between the 4 participating laboratories was achieved for uranium assay, uranium isotope composition, strontium isotope composition, and trace element analysis. In conclusion, the combination of these four signatures is sufficient to distinguish the different origins among these studied UOCs.

and nuclear chemistry↗

Computational studies of impurity migration during induction stirring of molten uranium

Understanding and controlling impurity behavior in actinide metal casting processes are foundational for efficient part production yet remain major challenges for researchers and industry. To help provide insight regarding impurity distribution during actinide metal casting, we have developed computational fluid dynamics (CFD) models for a laboratory-scale system using commercial and open-source code. Multiple simulation frameworks allow for improved confidence in the resulting outputs, while taking advantage of the maturity and convenience offered by commercial platforms and simultaneously maintaining the transparency and flexibility often provided by open-source software. Here, we describe multiple experiment-informed models designed to simulate a specific laboratory system in which uranium melt, containing a known starting concentration of carbon impurity, is electromagnetically stirred in an induction furnace. Here, the goal of the simulations is to predict the motion of uranium carbide microparticles in the velocity field of the melt. Prior to simulating the uranium-carbon system, numerical models were validated using a previously published nonradioactive experimental system. Effects of the size and shape of impurity particles in the models were investigated and agree with experimental findings. Simulation of smaller particles (< 50 µm) shows more homogenous distribution throughout the stirred melt. With increased particle size (100 µm), the body forces, which include the buoyancy force, dominate over the drag force, causing larger particles to move toward the crucible walls and upward in the system.

36 MATERIALS SCIENCE↗