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At least 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Characterizing the Chemical Behavior of Uranium Compounds for Nuclear Forensics

Better characterization of uranium compounds related to the nuclear fuel cycle is a key goal of the nuclear forensics community. Understanding the chemical behavior of these compounds in various environmental conditions and the effect on observable properties is crucial to the identification of unknown materials. Uranyl fluoride, the hydrolysis product of uranium hexafluoride, is a material of particular interest. Previous studies have suggested that uranyl fluoride may degrade upon exposure to high humidity, but the chemical pathway of degradation was previously not well understood. Our work demonstrates that uranyl fluoride undergoes a chemical reaction with water vapor to form uranyl hydroxide and peroxide species. This understanding will help guide forensic analysis in the future.

Kirkegaard, Marie↗
DOE OSTI2018

Concurrent Uranium Overmass and Hydraulic Fluid Leak

In the Spring of 2017, the Y-12 National Security Complex had a NCS operational event involving hydraulic fluid leaking into a glovebox containing fissile material. The hydraulic fluid intermixed with fissile material in the glovebox but remained in a safe slab-like configuration. Unknown at the time, a large enough mass of uranium to support a moderated nuclear criticality accident resided below the glovebox floor in an unfavorable volume sand separator system. Thanks to the operator following procedures and shutting down the hydraulic pump supplying the jolter, a potential accident was adverted. As a result of the event, significant physical modifications were made to the process to enhance safety. Additionally, changes were made in how the materials in the glovebox are processed in order to prevent liquid intrusion and significant uranium accumulation in the sand separator system.

42 ENGINEERING↗
DOE OSTI2018

Uranium Accumulation Discoveries at Y-12

Since discovering an unexpected accumulation of uranium in the Reduction Area of Building 9212 last June, CNS has been executing an extent-of-condition review that has led to additional discoveries of uranium accumulation. In all instances of accumulation discoveries, CNS has paused operations while executing corrective actions to address identified issues. CNS is aggressively addressing the underlying causes of the accumulations across its facilities and programs, to include expanding and accelerating the scope of extent-of-condition walk-downs and improvement actions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

An Evaluation of Systematic Biases of the Point Kinetics Model for Active Interrogation of Uranium

The point kinetics model relies upon a series of assumptions that simplify the process of neutron fission chain propagation. Tagged neutron interrogation of uranium using 14.1 MeV neutrons can potentially benefit from the use of the point kinetics model, but the appropriateness of the model's assumptions have yet to be validated for this application. The two assumptions considered by this work are the assumption that a single distribution can be used to describe the neutron multiplicity of a chain-starting event and the assumption that parasitic neutron capture within uranium is negligible. Monte Carlo simulations are used to evaluate the appropriateness of each assumption, as well as the consequences on the emitted doubles rate. This work demonstrates that the use of the point kinetics model for this application is possible, but corrections to the source multiplicity assumption are necessary that would likely require prior knowledge of sample isotopics.

Tweardy, Matthew↗
DOE OSTI2018

Transient Reactor Test Facility (TREAT) Low Enriched Uranium (LEU) Conversion Progress

The National Nuclear Security Administration Office of Material Management and Minimization (NNSA M3) is working towards converting the Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory (INL) from its original high-enriched uranium (HEU) core to a low-enriched uranium (LEU) core. TREAT reactor operations were restarted in November of 2017 and data obtained from the initial testing will be used to validate the existing ANL reactor modeling. Fuel block development has progressed with the completion of 2 and 4 inch crack free surrogate fuel blocks. Alternative fabrication methods for the fuel blocks continue to be explored and are showing promising results. Additionally, cladding oxidation studies have been performed and analyzed to initiate the down selection of cladding material to be used for the TREAT LEU fuel elements. This paper describes the current status of the above mentioned activities, and discusses the anticipated path forward for the conversion program as it progresses toward a final fuel design and fuel fabrication process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

A Prediction of Uranium(VI) Diffusion in Montmorillonite at Various Chemical Solution Conditions

Most nuclear waste disposal options include compacted bentonite, predominately consisting of montmorillonite clay, as part of a barrier system to minimize contaminant mobility. Uranium (U) is the primary element in spent nuclear fuel, and a potential contaminant of water resources. Uranium sorption onto clay and its slow diffusive transport away from waste canisters are expected to limit U(VI) mobility, and possibly also control nuclear fuel degradation rates based on mass action considerations. A prediction of U(VI) diffusion in these systems is complicated by two main factors: (1) the complex mineralogical structure of montmorillonite clay, leading to two types of clay porosities and surface sites, and (2) the dependence of U(VI) sorption on U(VI) solution speciation and the specific chemical system conditions. For instance, depending on pH and ionic strength, U(VI) may sorb onto montmorillonite due to weak ion exchange reactions or the formation of stable surface complexes. Furthermore, in contrast to conventional porous media, U(VI) diffusion in montmorillonite is not only limited by contaminant sorption, but also by a (potential) lack of access to the full clay porosity for some U(VI) solution species. For instance, at the high degrees of clay compaction expected in future barrier systems, the partial or full exclusion of U(VI) from negatively-charged clay interlayer spaces may substantially decrease U(VI) fluxes. In this modeling study, we specifically investigated how the charge of U(VI) solution species affects U(VI) sorption and diffusion behavior. For this purpose, we first simplified complex chemical solution speciation diagrams to charge speciation diagrams. Then, we coupled an existing U(VI) surface complexation model (Tournassat et al., Geochim. Cosmochim. Acta, 2018) with an analytical solution of the diffusion equation to simulate U(VI) mobility in lab-scale through-diffusion experiments for various chemical conditions. Our results suggest that there are two important parameters that govern U(VI) diffusion in these systems. First, U(VI) surface complexation on clay edge surfaces leads to contaminant retardation. On the field scale, this will determine how early a plume may cross site boundaries and pose risks to drinking water resources. Second, the degree of access of U(VI) solution species to diffuse layers in clay interlayer spaces largely controls U(VI) diffusive fluxes. In the field, this parameter will govern the level of U(VI) concentrations that can be expected in contaminated waters.

Pistorino, Jonathan C.↗
DOE OSTI2019

Simulation of Uranium(VI) Speciation and Sorption in Montmorillonite Clay in the Presence of Calcite Impurities

Montmorillonite is an abundant clay mineral that has been identified as the primary component in engineered bentonite barrier systems in future nuclear waste repositories. Selective sorption onto planar and edge surfaces of clay particles make it a great candidate for the attentuation of uranium(VI). Although previous works have investigated uranium(VI) interactions with this clay, few have done so in the presence of calcite, a mineral impurity in bentonite that is understood to alter U(VI) speciation, sorption, and consequently its mobility. PHREEQC software is used to simulate the chemical solution speciation of U(VI) under various known repository conditions, in the absence and presence of calcite, and across a range of U(VI)concentrations (1.0E-7 to 1.0E 5 M), and partial pressures of CO 2 (zero to 2% CO2) (Fig. 1). In addition, the kinetic dissolution of calcite is simulated to gain a better understanding of how quickly chemical equilibrium can be reached in these systems. Lastly, surface complexation models will be developed to characterize the influence of calcite impurities on the formation of ternary U(VI)-Ca-carbonate solution species on U(VI) sorption behavior. Our results indiciate that the presence of Ca 2+ and CO 3 2- ions released from calcite leads to U(VI)-Ca-carbonate species with a negative or neutral charge even without the presence of CO 2 . Quantification of these species will allow for the development of more accurate U(VI)-montmorillonite sorption models. Calcite precipitation at alkaline pH further complicates speciation by creating a competation for Ca 2+ . This research is being performed using funding received from the DOE Office of Nuclear Energy’s NEUP.

Pistorino, Jonathan C.↗
DOE OSTI2020

Emulating Variable Uranium Enrichments With Radiation Signature Training Devices (RSTD)

Radiological signature training devices (RSTDs) are sealed sources (or assemblies of sources) that emulate the radiation signature of larger masses of special nuclear material for passive detection while maintaining a minimal logistical footprint in security, radiological safety, criticality safety, and transportation requirements. Previous work focused on emulation of large quantities of highly enriched uranium for search and field identification with radioisotope identification detectors and mobile systems. New work has expanded into investigating use of the RSTDs with laboratory safeguards tools, such as MGAU and FRAM, and emulating uranium spectra of multiple enrichments over more energy regions (particularly the x-ray region).

Blessinger, Christopher S.↗
DOE OSTI2022

A Comparison of Homogeneous and Heterogeneous Uranium Metal-Water Systems using Calculated and Experimental Critical Data

Heterogeneous effects of fissile units latticed in water has been documented in several forms in commonly referenced handbooks and guides. Many of these documents provide guidance for when heterogeneous systems are more reactive than their homogenous counterparts if controlling fissile mass or volume. This information provides insight into when heterogeneity should be considered for conservatism, however this information is not always displayed in a manner that is comparable to commonly referenced data, such as critical curves based on spherical systems. This paper aims to provide a juxtaposition of homogenous and heterogeneous uranium metal-water systems using both experimental and calculated critical data displayed in the format of commonly referenced critical curves. This format allows for easy comparison between the two systems against parameters that are often considered for single unit analysis such as: uranium density, fissile mass, system volume, and total system mass (fissile mass plus moderator mass). Calculated critical spherical systems provide an equal comparison between latticed geometry types as the latticed array can be cut off or restricted in the same manner in each series of models, which is not true for experimental values. Additionally, enrichment of the latticed units can be made uniform in the calculated systems, which is also difficult to achieve when referencing experimental data of varying latticed geometries. A comparison of calculated critical values and available experimental values is provided to show how these parameters can impact the critical spherical system result.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A Comparison of Homogeneous and Heterogeneous Uranium Metal-Water Systems using Calculated and Experimental Critical Data

Heterogeneous effects of fissile units latticed in water has been documented in several forms in commonly referenced handbooks and guides. Many of these documents provide guidance for when heterogeneous systems are more reactive than their homogenous counterparts if controlling fissile mass or volume. This information provides insight into when heterogeneity should be considered for conservatism, however this information is not always displayed in a manner that is comparable to commonly referenced data, such as critical curves based on spherical systems. This paper aims to provide a juxtaposition of homogenous and heterogeneous uranium metal-water systems using both experimental and calculated critical data displayed in the format of commonly referenced critical curves. This format allows for easy comparison between the two systems against parameters that are often considered for single unit analysis such as: uranium density, fissile mass, system volume, and total system mass (fissile mass plus moderator mass). Calculated critical spherical systems provide an equal comparison between latticed geometry types as the latticed array can be physically cut off or restricted in the same manner in each series of models, which is not true for experimental values. Additionally, enrichment of the latticed units can be made uniform in the calculated systems, which is difficult to achieve when referencing experimental data of varying latticed geometries. A comparison of calculated critical values and available experimental values is provided to show how these parameters can impact the critical spherical system result.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Ab Initio Evaluation of Uranium Carbide S(α,β) and Thermal Neutron Cross Sections

Uranium Carbide (UC) is a nuclear fuel material which offers better neutron economy and lower fuel-cycle costs compared to conventional mixed-oxide fuels. UC’s lattice binding and dynamical properties impact thermal neutron scattering and low temperature epithermal resonance absorption. The Thermal Scattering Law (TSL) describes the scattering system available energy and momentum transfer states. There is no TSL evaluation for UC in the ENDF/B-VIII.0 database; herein, ab-initio lattice dynamics (AILD) techniques are invoked to calculate the phonon spectrum for UC using spin-orbit-coupling density functional theory (DFT). The TSLs, inelastic and elastic thermal scattering cross sections for Uranium and Carbon in UC, respectively, are calculated in FLASSH for use in higher fidelity reactor design calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Neutron Coincidence Measurements of Uranium-233 Oxide

Renewed international interest in thorium-fueled advanced reactors has challenged the safeguards community to address future proliferation concerns. Thorium-based technology presents many benefits but does not eliminate the proliferation risks associated with producing and processing fissile material. A byproduct of thorium-fueled reactors is uranium-233, which is classified as a direct-use material. As a result, the development of new or improved methods to characterize and measure materials containing 233U must mirror the pace of development of reactors and facilities that produce such material. Research is underway to assess, develop, and test approaches for safeguarding nuclear materials within the thorium fuel cycle. Neutron signatures from the nondestructive assay (NDA) of materials containing 233U are being quantified to inform the potential characterization of these materials. Using a traditional neutron coincidence counter and a series of well-documented 233U oxide samples, initial measurements have been made to assess the feasibility of 233U characterization and discrimination from other uranium isotopes, primarily 235U, using a combination of measurement techniques and analysis methods. Data acquisition is performed in list mode, allowing for a variety of analyses to be performed on the raw data that is not available using traditional shift register technology. Measurements were performed in passive and active configurations to quantify the strength of signal and to validate simulations in support of this work. This paper presents and discusses the results of the initial measurements of 233U oxide performed at Oak Ridge National Laboratory.

Lockhart, Madeline↗
DOE OSTI2020

Leveraging Machine Learning Capabilities for the Characterization of Irradiated Uranium: A Case Study of Analysis Methods for Nuclear Safeguards and Nuclear Forensics

Nondestructively determining the initial enrichment of irradiated uranium is a complex and laborious multivariable problem due to the presence of fission products. This work demonstrates the capabilities of machine learning to analyze gamma-ray spectral data to determine initial enrichment without knowledge of the decay time of the sample. The approach developed is agnostic to the particular scenario and is applicable to a wide variety of applications in nuclear forensics and nuclear safeguards. We irradiated 5 mg uranium standard reference materials at discrete enrichment values ranging from 0.02% to 97% 235 U (weight percent) in UT Austin’s Nuclear Engineering Teaching Laboratory TRIGA Mark II 1.1 MW research reactor, allowed each to decay for 8 hours, and then measured each sample via gamma-ray spectrometry for 50 hours post-irradiation yielding 1,400 individual gamma-ray spectra discretized into 8,192 energy bins. We then trained decision trees models to analyze individual gamma-ray spectra and estimate the associated initial enrichment without knowledge of the time since end of irradiation. We evaluated the performance of the models with a reserved test set not used for training or calibrating the model. A decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 2.3% (weight percent 235 U). Next, we implemented a principal component analysis pre-processing routine of the gamma-ray spectrometry data to reduce the dimensionality of the dataset from 8,192 channels in the spectrum to 10 principal components while retaining over 99% of the inherent variance in the data. Decision tree models constructed with these data demonstrated decreased mean absolute error in enrichment determination, reduced computational time, and decreased complexity. A single decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 0.05% (weight percent 235 U). Furthermore, we analyzed these models with learning curves to ensure that overfitting did not occur. The capabilities provided by these models can be naturally extended to other application-focused measurements in the fields of nuclear safeguards, nuclear forensics, and nuclear non-proliferation.

Drescher, Adam↗
DOE OSTI2025

Hydrochlorination of Uranium Dioxide in a Molten Salt Mixture - Phase 1: Tube Furnace Cross Flow Experiments

In 2023, Metatomic® Inc., a South Carolina based company, was awarded a Gateway for Advanced Innovation in Nuclear (GAIN) research voucher for a proposed series of experiments aimed at demonstrating the viability of a spent nuclear fuel (SNF) recycling process patented by Met atomic® Inc. For the GAIN voucher, Metatomic® Inc. selected Savannah River National Laboratory (SRNL) as a partner in executing the proposed proof-of-concept experiments. This report outlines the proof-of-concept experiments performed by SRNL for Metatomic® Inc. during Phase 1 (of 2) experimentation. The Phase 1 hydrochlorination experiments consisted of weighing UO2 into alumina crucibles with a eutectic mixture of NaCl and CsCl, heating the uranium/salt mixture to a varied temperature (550, 650, or 750 °C), and flowing anhydrous hydrogen chloride (AHCl) gas across the surface of the uranium-bearing molten salt mixture. The hydrochlorination process conditions were maintained for 4-5 hours and the percent conversion for each batch of UO2 was determined using a suite of analytical characterization techniques. The degree of UO2 conversion was found to be greatest at the highest tested temperature, ultimately achieving 32.9% conversion of UO2 to water soluble uranium-chloro species (e.g., UO2Cl2, UCl4) after exposure to AHCl for 4.25 hours at 750 °C.

Nguyen, Vinh T. [Savannah River National Laborator↗
DOE OSTI2022

The Use of Acoustic Cavitation to Recovery High-Assay Low-Enriched Uranium (HALEU) form TRISO Fuel Particles

Recent interest in advanced nuclear reactor concepts such as small modular reactor (SMR), micro reactor or versatile test reactor (VTR) has increased the demand for high-assay, low-enriched uranium (HALEU). The sustainable management of this resource is integral to successful resurgence of nuclear energy in the United States. The High Temperature Gas-Cooled Reactor (HTGR) concept is a prime candidate for deployment of the new fleet of SMRs, and it will become very important to have a simple and cost-effective technology to recover HALEU from short-burned, defective, or off-specification TRISO fuel particles. We have attempted to harness the high energetics of sonochemistry to penetrate the TRISO fuel particles and recover HALEU via acidic leaching. While the TRISO fuel particle design is inherently “bulletproof,” (due to the silicon carbide (SiC) layer), ultrasonic irradiation may hold the key to convenient access to its valuable contents. Sonochemistry is a field of chemistry based on acoustic cavitation, which is the formation, growth, and collapse of bubbles in liquid media.1,2 The oscillating bubble formation is produced by irradiation of a liquid media with sound waves. Literature reports indicate that collapsing bubbles induced by cavitation produces intense local heating, high pressures, and short lifetimes.1 These localized hot spots reach temperatures of ˜5000 K, pressures approaching 500 atm, and heating and cooling rates exceeding 100 K/s.1 The large temperature and pressure differentials deliver high-energy heating and microscopic explosive shock waves to a liquid media or liquid/solid interface. Cavitation at the surface of a solid in solution induces a deformation in the bubble cavity upon collapse. This deformation reinforces the bubble structure and sends a fast-moving stream of liquid through the cavity at the surface with velocities greater than 100 m/s.3 These energetic impacts have demonstrated an ability to penetrate or simply destroy the SiC shell surrounding the uranium fuel kernel during prolonged exposures to the high-power acoustic waves. Some preliminary results demonstrating HALEU recovery using sonochemistry techniques will be reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Internal-Gelation Production of Uranium Oxide Sol-Gel Particles for Forensic Applications

The purpose of this project was to develop and demonstrate a novel method for the production of uranium oxide microsphere particles with tunable chemical compositions via a sol-gel process using a 3D-printer setup. These particles can serve several purposes in research and development as a forensic training tool or as standard reference materials. A key component of the project was to demonstrate the ability to control physical and chemical parameters of the particles created. First, we demonstrated the ability to employ an internal gelation sol-gel process to create individual uranium oxide particles. The particles were successfully dispensed using a unique 3D-printing setup onto a substrate to react and then were collected and thermally processed. A series of temperatures for the annealing process was tested on individual samples to investigate the effect on the sol-gel chemical composition and physical integrity. Next, we demonstrated the ability to control matrix composition of the particles by separately incorporating fission product isotopes as well as Np-237 into the sol-gel solution. It was shown by gamma-ray spectroscopy that these matrix elements were successfully retained during the gelation process. We studied the retention of the elements across a series of annealing temperatures. Additionally, we demonstrated the ability to quantitatively control the isotopic composition of the particles by altering the U-237/U-238 ratio to a controlled value. Finally, X-ray diffraction analysis (XRD) was used to investigate the oxidation state of the sol-gel after annealing at different temperatures.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2014

Communication: Relativistic Fock-space coupled cluster study of small building blocks of larger uranium complexes

We present a study of the electronic structure of the [UO{sub 2}]{sup +}, [UO{sub 2}]{sup 2} {sup +}, [UO{sub 2}]{sup 3} {sup +}, NUO, [NUO]{sup +}, [NUO]{sup 2} {sup +}, [NUN]{sup −}, NUN, and [NUN]{sup +} molecules with the intermediate Hamiltonian Fock-space coupled cluster method. The accuracy of mean-field approaches based on the eXact-2-Component Hamiltonian to incorporate spin–orbit coupling and Gaunt interactions are compared to results obtained with the Dirac–Coulomb Hamiltonian. Furthermore, we assess the reliability of calculations employing approximate density functionals in describing electronic spectra and quantities useful in rationalizing Uranium (VI) species reactivity (hardness, electronegativity, and electrophilicity)

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2017

Evaluation of 235 U Isotopic Dilution with the Addition of Depleted Uranium to Supernatant Liquid Waste

Laboratory-scale results on experiments performed to examine the feasibility of isotopic dilution of 235 U in supernatant liquid storage tanks at the Savannah River Site are described. Here, the isotopic dilution tests were accomplished by adding an alkaline depleted uranium solution to small portions of simulated and actual storage tank waste solutions with enriched 235 U compositions. Based on the laboratory observations, recommendations were made, which involved the addition of significant quantities of uranyl carbonate solution to more than 4 million l of 235 U enriched waste stored in tank 43H at the site to reduce the risk for criticality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗