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At least 469 records · Page 26

Rapid Oxidation of Uranium Steel Alloys: Combustion Synthesis

This project explored the use of combustion synthesis as a rapid, high-temperature method for oxidizing uranium-bearing steel alloys. Traditional laboratory-scale synthesis methods often fail to replicate the thermal and kinetic conditions experienced by real-world particulates, particularly those formed under rapid quenching or high-temperature scenarios. Combustion synthesis offers a promising alternative by enabling fast, localized heating and flexible precursor selection. A series of targeted experiments were conducted using a U 2 NiCrFe 4 alloy as the precursor. The alloy was oxidized using combustion synthesis reactions fueled by uranyl nitrate and glycine, achieving peak temperatures exceeding 1,200 °C. Postreaction analysis using scanning electron microscopy (SEM), elemental mapping, and Raman spectroscopy revealed the formation of iron-based oxides, with limited but detectable evidence of uranium oxide phases such as UO 2 . The results indicate that under the rapid reaction and cooling conditions of combustion synthesis, iron oxides form preferentially, but uranium oxide formation is kinetically limited. These findings validate combustion synthesis as a viable method for simulating the oxidation behavior of uranium steels in extreme environments and lay the groundwork for future studies aimed at enhancing uranium oxide formation through higher temperatures or modified precursor compositions.

36 MATERIALS SCIENCE↗

Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

Grisaffe, Salvatore J.↗

Measured and calculated fast neutron spectra in a depleted uranium and lithium hydride shielded reactor

Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.

Lahti, G. P.↗

Production of fissioning uranium plasma to approximate gas-core reactor conditions

The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

Lee, J. H.↗

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Briggs, Maxwell H.↗

Recovering and recycling uranium used for production of molybdenum-99

A processes for recycling uranium that has been used for the production of molybdenum-99 involves irradiating a solution of uranium suitable for forming fission products including molybdenum-99, conditioning the irradiated solution to one suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina. Another process involves irradiation of a solid target comprising uranium, forming an acidic solution from the irradiated target suitable for inducing the formation of crystals of uranyl nitrate hydrates, then forming the crystals and a supernatant and then separating the crystals from the supernatant, thus using the crystals as a source of uranium for recycle. Molybdenum-99 is recovered from the supernatant using an adsorbent such as alumina.

36 MATERIALS SCIENCE↗

Gamma ray Imaging Based Methods to Improve the Accuracy of Uranium Holdup Quantification

Deposits in commercial uranium processing facilities of interest to nuclear safeguards are predominantly low enriched uranium (LEU). The shapes and sizes of holdup deposits can vary significantly, thus making it difficult to model the deposits accurately for calibration purposes. This presents a challenge to currently employed methods such as the Generalized Geometry Holdup (GGH) that rely on quantifying 235U mass by simplifying deposit shapes as a point, a line, or an area. In this work, gamma ray imaging using high energy resolution Germanium Gamma Imagers (GeGI™) is employed to determine the distribution of uranium inside the source containment. Two types of imaging methods are employed; coded-aperture imaging and Compton imaging. Gamma-ray emissions seen in different regions of the image are quantified using an inverse gamma-ray transport solver being developed at ORNL. Using measured data, the inverse method will be used to solve for unknown source parameters such as source matrix thickness, density, and attenuation due to container wall and shielding. The intrinsic efficiency of the detector is determined based on a library of built-in response functions and key dimensions of the detector (e.g., thickness and radius of the detector crystal). Measurements are performed in both imaging modalities using uranium sources of well-known masses and enrichments configured inside mocked-up holdup fixtures. Results from the two imaging modalities are intercompared. The paper presents preliminary results of the imaging measurements and the progress made in the computational methods to quantify uranium mass.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

R “SHINY” GUI DEVELOPMENT FOR URANIUM ISOTOPIC ANALYSIS WITH MATRIX-ASSISTED IONIZATION MASS SPECTROMETRY

The international nuclear safeguards community continues to seek rapid, accurate, and precise characterization capabilities for the in-field measurement of uranium isotopic compositions in nuclear facilities. Mass spectrometry (MS) is considered the “gold standard” for analysis of relatively long-lived actinides such as uranium (U) and plutonium; however, conventional MS analysis often requires time consuming sample preparation and complex analytical methodologies that are difficult to perform in-field or in-facility. Matrix assisted ionization (MAI) is a novel ambient ionization MS technique (i.e., MAI-MS) that potentially addresses these challenges due to the relative simplicity of the ionization phenomenon and ruggedness of ambient MS instrumentation. Savannah River National Laboratory (SRNL, USA) has demonstrated this technique for nanogram-level 235U/238U isotope ratio measurements within seconds, with percent-level analytical uncertainties capable of discriminating depleted, natural, and low-enriched uranium. Current experimental work on developing MAI methods for uranium isotopic analysis has been enabled by parallel development of a comprehensive MAI-MS data analysis suite at SRNL. Development of this bespoke data analysis software was necessary because commercially available ambient MS software is poorly suited for uranium isotope ratio measurement. The effort leverages the power of R, a popular open-source programming language, and Shiny, an R package providing tools for graphical user interface (GUI) and web interface coding. This software allows researchers without any programming experience to harness and utilize R’s considerable data analysis/visualization power.

LaBone, Elizabeth D.↗

Laser-Induced Spectrochemical Assay for Uranium Enrichment (LISA-UE)

Uranium hexafluoride (UF6) is the uranium compound typically involved in uranium enrichment process. As the first line of defense against nuclear proliferation, accurate determinations of the uranium enrichment ratio in UF6 are critical for materials verification, accounting and safeguards. Shipping gaseous UF6 samples off-site for analysis with mass spectrometry is cumbersome and costly, and results are not available for some time (months). In-field UF6 enrichment assay has the potential to substantially reduce the time, logistics and expense of sample handling. At present, COMPUCEA is the only accepted method for UF6 enrichment assay in the field. Laser-Induced Spectrochemical Assay for Uranium Enrichment (LISA-UE) is an all-optical (based on laser induced plasma emission) analytical technique intended for fieldable, accurate, precise and rapid UF6 enrichment assay. In its operation, laser induced plasma is created directly in the gaseous UF6 sample. Because different U isotopes emit at slightly different wavelengths, the isotopic information of the UF6 sample is inherently encoded in the atomic emission from the plasma. Isotopic emissions from 235U and 238U are measured simultaneously, which eliminate correlated noise from the laser induced plasma. Isotopic information of the UF6 sample can be extracted from the acquired spectrum with theoretical multi-variable non-linear spectral fitting. To date, advances made by the LISA-UE research team include optimization of the spectral window for direct gaseous UF6 enrichment assay with laser induced plasma, development of data reduction algorithms, and demonstrations of the LISA-UE technique with gaseous UF6 samples. In this presentation, the technical aspect of LISA-UE will be overviewed, the data reduction algorithm will be described, and performance of the technique will be discussed.

Chan, George↗

Determination of Uranium Metal Concentration in Irradiated Fuel Storage Basin Sludge Using Selective Dissolution

Uranium metal corroding in water-saturated sludges now held in the US Department of Energy Hanford Site K West irradiated fuel storage basin can create hazardous hydrogen atmospheres during handling, immobilization, or subsequent transport and storage. Knowledge of uranium metal concentration in sludge thus is essential to safe sludge management and process design, requiring an expeditious routine analytical method to detect uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of 30 wt% or higher total uranium concentrations.

Nuclear fuel storage basin sludge↗

Chemical and Sr isotopic characterization of North America uranium ores: Nuclear forensic applications

This study reports major, minor, and trace element data and Sr isotope ratios for 11 uranium ore (uraninite, UO 2+x ) samples and one processed uranium ore concentrate (UOC) from various U.S. deposits. The uraninite investigated represent ores formed via different modes of mineralization (e.g., high- and low-temperature) and within various geological contexts, which include magmatic pegmatites, metamorphic rocks, sandstone-hosted, and roll front deposits. In situ trace element data obtained by laser ablation-ICP-MS and bulk sample Sr isotopic ratios for uraninite samples investigated here indicate distinct signatures that are highly dependent on the mode of mineralization and host rock geology. Relative to their high-temperature counterparts, low-temperature uranium ores record high U/Th ratios (>1000), low total rare earth element (REE) abundances (<1 wt%), high contents (>300 ppm) of first row transition metals (Sc, Ti, V, Cr, Mn, Co, Ni), and radiogenic 87 Sr/ 86 Sr ratios (>0.7200). Comparison of chondrite normalized REE patterns between uraninite and corresponding processed UOC from the same locality indicates identical patterns at different absolute concentrations. Lastly, this result ultimately confirms the importance of establishing geochemical signatures of raw, uranium ore materials for attribution purposes in the forensic analysis of intercepted nuclear materials.

07 ISOTOPE AND RADIATION SOURCES↗

Rare-earth element fractionation in uranium ore and its U(VI) alteration minerals

We developed a cation exchange chromatography method employing sulfonated polysterene cation resin (DOWEX AG50-X8) in order to separate rare-earth elements (REEs) from uranium-rich materials. The chemical separation scheme is designed to reduce matrix effects and consequently yield enhanced ionization efficiencies for concentration determinations of REEs without significant fractionation using solution mode-inductively coupled plasma mass spectrometry (ICP-MS) analysis. This method was then applied to determine REE abundances in four uraninite (ideally UO 2 ) samples and their associated U(VI) alteration minerals. In three of the samples analyzed, the concentration of REEs for primary uraninite are higher than those for their corresponding secondary uranium alteration phases. The results for U(VI) alteration minerals of two samples indicate enrichment of the light REEs (LREEs) over the heavy REEs (HREEs). This differential mobilization is attributed to differences in the mineralogical composition of the U(VI) alteration. There is a lack of fractionation of the LREEs in the uraninite alteration rind that is composed of U(VI) minerals containing Ca 2+ as the interlayer cation (uranophane and bequerelite); contrarily, U(VI) alteration minerals containing K + and Pb 2+ as interlayer cations (fourmarierite, dumontite) indicate fractionation (enrichment) of the LREEs. Our results have implications for nuclear forensic analyses since a comparison is reported between the REE abundances for the CUP-2 (processed uranium ore) certified reference material and previously determined values for uranium ore concentrate (UOC) produced from the same U deposit (Blind River/Elliott Lake, Canada). UOCs represent the most common form of interdicted nuclear material and consequently is material frequently targeted for forensic analysis. The comparison reveals similar chondrite normalized REE signatures but variable absolute abundances. Based on the results reported here, the latter may be attributed to the differing REE abundances between primary ore and associated alteration phases, and/or is related to varying fabrication processes adopted during production of UOC.

58 GEOSCIENCES↗

Understanding neutron capture processes in uranium deposits using combined U-Sm-Nd isotopic compositions

Valuable insights into the history and evolution of a geologic deposit can be found by investigating neutron capture reactions. Thermal neutron capture reactions occur within both the samarium (Sm) and the uranium (U) systems, where 149 Sm and 235 U can capture neutrons to become 150 Sm and 236 U, respectively. Although largely unexplored, paired measurements of 150 Sm and 236 U could be important for understanding neutron capture effects within uranium ore bodies, and such measurements are potentially useful in nuclear forensics for assessing a material's provenance or mineral exploration. In this work, we refined measurement procedures of Sm isotope compositions utilizing MC-ICPMS. While geologic reference materials were found to have indistinguishable Sm isotope compositions, we found significant isotope variations consistent with nuclear field shift among synthetic Sm standards. Here, this observation highlights that future high-precision Sm isotope investigations need to carefully evaluate synthetic standard(s) against geologic reference materials until an unfractionated and agreed-upon standard is identified. Here, we applied this method to a set of nine uranium ores from the South Australian Beverley North uranium deposits. Although 236 U excesses had been previously reported for these U ores, we found no measurable isotopic shifts in 149 Sm- 150 Sm at the current level of precision (±5 parts per million). One possible explanation for this disparity in the observed neutron capture signatures between U and Sm is that the source(s) of the U and Sm in these ores may be decoupled. This is consistent with the finding that these ores have variable 143 Nd/ 144 Nd, thus demonstrating that diverse sources were involved in the formation of the Beverley North deposits. Alternatively, this deposit may be too young (<50Ma) to have accumulated measurable neutron capture effects in Sm to be detected with the methods employed here.

Mineral exploration↗

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗

The effects of low- Z shielding on uranium isotope discrimination using the time-emission profiles of long-lived delayed neutrons

Characteristic radiation signals can often be obscured or distorted by relatively small amounts of shielding, as is the case for highly-enriched uranium (HEU). Methods that can detect and identify special nuclear material (SNM) even in the presence of shielding are therefore of high value in many nuclear security and nonproliferation applications. It has been previously shown that the buildup and decay time profiles of long-lived delayed neutron emission from induced fission are unique and can be used to successfully discriminate between uranium isotopes and infer enrichment without the need for a calibration standard. Because the long-lived delayed neutron precursors exhibit decay times on the order of tens of seconds, their time profiles are not sensitive to the delays associated with scattering or diffusion in shielding materials, which occur on much shorter time scales. The use of capture-based neutron detectors can also mitigate the effects of shielding, because unlike recoil-based detectors, they do not have a low-energy detection threshold, so that neutrons that lose their energy in the shielding are still detectable. In this work, the effects of low-Z shielding on the discrimination of uranium isotopes using delayed neutron time profile measurements are investigated for the first time. Furthermore, the delayed neutron buildup and decay time profiles of polyethylene-shielded HEU and depleted uranium objects are measured using two different capture-gated detector designs. The recorded profiles show good agreement with the shapes predicted from six-group delayed neutron models even with the shielding in place, and the enrichment of the HEU and DU objects was estimated accurately to within 5%. While this method shows potential for discriminating and characterizing extreme cases of enrichment when a moderate amount of shielding is present, simulation results suggest that significant limitations exist for intermediate enrichment levels due to distortions to the delayed neutron time profile caused by neutron reflection in the shielding.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Strategies toward the Synthesis of Advanced Functional Sorbent Performance for Uranium Uptake from Seawater

A polymer fiber-based adsorbent (AF1) composed of acrylonitrile and itaconic acid functional groups was synthesized by a radiation-induced graft polymerization technique onto hollow-gear shaped polyethylene (PE) fibers. Investigation of the optimum reaction parameters for the conversion of grafted cyano moieties into amidoxime groups was conducted by the reaction with hydroxylamine at different temperatures and periods in a variety of aqueous and organic solvents. The 13 C CP/MAS spectra of amidoximated AF1 adsorbent fibers from water–methanol and dimethyl sulfoxide (DMSO) revealed that both the cyclic imide dioxime and open-chain amidoxime were formed through the reaction. The conversion from amidoxime to imide dioxime was found to occur slowly and gradually with the increasing reaction time. The quicker diffusion of DMSO as compared to that of water–methanol, in the grafted trunk PE fiber, resulted in faster kinetics of the amidoximation reaction. The uranium adsorption capacity of the amidoximated AF1 samples was determined after (i) 24 h contact with sodium-based brine spiked with uranium and (ii) 56 days of exposure to filtered seawater (Sequim Bay, WA, USA) in flow-through columns. Finally, the uranium extraction performance of the adsorbents after exposure to filtered seawater was consistent with the laboratory screening results, and the amidoximated AF1 samples (in DMSO at 70 °C for 3 h) exhibited the highest 56 day uranium adsorption capacity (5.04 ± 0.15 g U/kg-ads) with faster adsorption kinetics compared to the pristine AF1 adsorbents.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Amidinate Supporting Ligands Influence Molecularity in Formation of Uranium Nitrides

Uranium nitride complexes are attractive targets for chemists as molecular models for the bonding, reactivity, and magnetic properties of next-generation nuclear fuels, but these molecules are uncommon and can be difficult to isolate due to their high reactivity. Here, we describe the synthesis of three new multinuclear uranium nitride complexes, [U(BCMA) 2 ] 2 (μ-N)(μ-κ 1 :κ 1 -BCMA) (7), [(U(BIMA) 2 ) 2 (μ-N)(μ-N i Pr)(K 2 (μ-η 3 :η 3 -CH 2 CHN i Pr)] 2 (8), and [U(BIMA) 2 ] 2 (μ-N)(μ-κ 1 :κ 1 -BIMA) (9) (BCMA = N,N-bis(cyclohexyl)methylamidinate, BIMA = N,N-bis(iso-propyl)methylamidinate), from U(III) and U(IV) amidinate precursors. By varying the amidinate ligand substituents and azide source, we were able to influence the composition and size of these nitride complexes. 15 N isotopic labeling experiments confirmed the bridging nitride moieties in 7–9 were formed via two-electron reduction of azide. The tetra-uranium cluster 8 was isolated in 99% yield via reductive cleavage of the amidinate ligands; this unusual molecule contains nitrogen-based ligands with formal 1–, 2–, and 3– charges. Additionally, chemical oxidation of the U(IV) precursor U(N 3 )(BCMA) 3 yielded the cationic U(V) species [U(N 3 )(BCMA) 3 ][OTf]. Magnetic susceptibility measurements confirmed a U(IV) oxidation state for the uranium centers in the three nitride-bridged complexes and provided a comparison of magnetic behavior in the structurally related U(III)-U(IV)-U(V) series U(BCMA) 3 , U(N 3 )(BCMA) 3 , and [U(N 3 )(BCMA) 3 ][OTf]. At 240 K, the magnetic moments in this series decreased with increasing oxidation state, i.e., U(III) > U(IV) > U(V); this trend follows the decreasing number of 5f valence electrons along this series.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Theoretical analysis of uranium-doped thorium dioxide: Introduction of a thoria force field with explicit polarization

Thorium dioxide is used industrially in high temperature applications, but more insight is needed into the behavior of the material as part of a mixed-oxide (MOX) nuclear fuel, incorporating uranium. We have developed a new interatomic potential model including polarizability via a shell model, and commensurate with a prominent existing UO{sub 2} potential, to conduct configurational analyses and to investigate the thermophysical properties of uranium-doped ThO{sub 2}. Using the GULP and Site Occupancy Disorder (SOD) computational codes, we have analyzed the distribution of low concentrations of uranium in the bulk material, where we have not observed the formation of uranium clusters or the dominance of a single preferred configuration. We have calculated thermophysical properties of pure thorium dioxide and Th{sub (1−x)}U{sub x}O{sub 2} which generated values in very good agreement with experimental data.

36 MATERIALS SCIENCE↗