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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.

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At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Metallographic Characterization of Wrought Depleted Uranium

Metallographic characterization was performed on wrought depleted uranium (DU) samples taken from the longitudinal and transverse orientations from specific locations on two specimens. Characterization of the samples included general microstructure, inclusion analysis, grain size analysis, and microhardness testing. Comparisons of the characterization results were made to determine any differences based on specimen, sample orientation, or sample location. In addition, the characterization results for the wrought DU samples were also compared with data obtained from the metallographic characterization of cast DU samples previously characterized. No differences were observed in microstructure, inclusion size, morphology, and distribution, or grain size in regard to specimen, location, or orientation for the wrought depleted uranium samples. However, a small difference was observed in average hardness with regard to orientation at the same locations within the same specimen. The longitudinal samples were slightly harder than the transverse samples from the same location of the same specimen. This was true for both wrought DU specimens. Comparing the wrought DU sample data with the previously characterized cast DU sample data, distinct differences in microstructure, inclusion size, morphology and distribution, grain size, and microhardness were observed. As expected, the microstructure of the wrought DU samples consisted of small recrystallized grains which were uniform, randomly oriented, and equiaxed with minimal twinning observed in only a few grains. In contrast, the cast DU microstructure consisted of large irregularly shaped grains with extensive twinning observed in most grains. Inclusions in the wrought DU samples were elongated, broken and cracked and light and dark phases were observed in some inclusions. The mean inclusion area percentage for the wrought DU samples ranged from 0.08% to 0.34% and the average density from all wrought DU samples was 1.62E+04/cm 2 . Inclusions in the cast DU samples were equiaxed and intact with light and dark phases observed in some inclusions. The mean inclusion area percentage for the cast DU samples ranged from 0.93% to 1.00% and the average density from all wrought DU samples was 2.83E+04/cm 2 . The average mean grain area from all wrought DU samples was 141 μm 2 while the average mean grain area from all cast DU samples was 1.7 mm2. The average Knoop microhardness from all wrought DU samples was 215 HK and the average Knoop microhardness from all cast DU samples was 264 HK.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Uranium Prototype Glovebox and Furnace Design and Build (SPC-2954)

This specification contains the requirements for design, analysis, fabrication, testing and delivery of the Uranium Prototype Glovebox (UPG) and Furnace, and the associated equipment as described herein. The UPG will be primarily used for the containment of uranium and the prototyping of equipment to be used in higher hazard category gloveboxes. This will allow complex pieces of equipment to be tested in a less hazardous environment prior to fabrication and installation in other gloveboxes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

TEX-HEU: Integral Experiments with Highly Enriched Uranium and Polyethylene [Slides]

This presentation discusses Thermal/Epithermal eXperiments (TEX) in relation to integral experiments spanning a wide range of fission energy spectra utilizing existing NCSP fissile material assets, TEX-Pu for plutonium, and TEX-HEU highly enriched uranium (HEU) for highly enriched uranium. It also discusses the TEX-HEU design, the use of HEU fuel, and polyethylene parts. The experiment, executed in the Nuclear Criticality Experiment Research Center of the Nevada National Security Site is discussed, specifically how it was assembled on the Comet Universal Critical Assembly Machine. Measured quantities and various critical configurations are listed. In conclusion, the presentation states that they successfully assembled and documented the baseline TEX-HEU experiments, that six critical configurations and one reproducibility measurement were produced, and that they are compiling results to submit to the International Criticality Safety Benchmark Evaluation Project. A future experiment campaign is planned in 2021 with Hafnium as a diluent.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transformative Manufacturing of Adaptive Wire Feedstocks with Atomic Layer Deposition of Nb 2 O 5 onto Depleted Uranium [Slides]

Project Motivation: Uranium corrosion susceptible and highly catalytic; Exposure thickens oxide impacting material performance. This worsens with time; Sigma is uniquely positioned to tackle R&D challenges with Uranium processing. Possible Solutions: 1) Make new material for each weld. Occupational exposure, difficult to align resources ; 2) Coatings? Coating adds foreign material: can this be minimized?

36 MATERIALS SCIENCE↗
DOE OSTI2025

TEX-Chlorine Assemblies: Highly Enriched Uranium Plates with Sodium Chloride Absorbers Using Polyethylene Moderator and Polyethylene Reflector

This evaluation documents highly enriched uranium (HEU) experimental critical configurations with polyethylene moderators and sodium chloride absorbers conducted as part of the United States Nuclear Criticality Safety Program’s Thermal/Epithermal eXperiments (TEX) program. HEU-MET-MIXED-021 provides the benchmark evaluation of five TEX experiments designed to establish baseline configurations with HEU Jemima plates moderated by high density polyethylene (HDPE). The TEX-HEU experiments were designed to cover five different fission energy regimes by varying the thickness of the interstitial HDPE moderator, with varying fractions of thermal, intermediate, and fast fissions, and to be easily modified to accommodate test materials of interest. HEU-MET-INTER-013 documents the first TEX-HEU variation, incorporating hafnium in seven different experimental configurations. This evaluation covers an additional variant that incorporates absorber plates of compacted high-purity sodium chloride salt. These experiments were motivated by a criticality safety need for validation data for uranium purification by means of electrorefining with chloride salts, especially thermal and intermediate energy configurations resulting from moderator upset conditions, and their design was optimized by matching sensitivity profiles from application cases. All three experimental configurations are judged to be acceptable as benchmark cases. The main parameter varied between the configurations is the thickness of the polyethylene moderators and the sodium chloride absorbers between the HEU plates. Varying the thickness of the polyethylene tunes the neutron energy spectrum between majority thermal (Case 1 and 2) and intermediate (Case 3). The fission fractions, presented in Table 1, are determined calculationally. Case 3 is cross listed as HEU-MET-INTER-014.

42 ENGINEERING↗
DOE OSTI2017

Modeling Substrate Utilization, Metabolite Production, and Uranium Immobilization in Shewanella oneidensis Biofilms

In this study, we developed a two-dimensional mathematical model to predict substrate utilization and metabolite production rates in Shewanella oneidensis MR-1 biofilm in the presence and absence of uranium (U). In our model, lactate and fumarate are used as the electron donor and the electron acceptor, respectively. The model includes the production of extracellular polymeric substances (EPS). The EPS bound to the cell surface and the EPS distributed in the biofilm were considered bound EPS (bEPS) and loosely associated EPS (laEPS), respectively. COMSOL ® Multiphysics finite element analysis software was used to solve the model numerically (model file provided in the Supplementary Material). The input variables of the model were the lactate and fumarate concentrations, cell density, EPS density, half saturation constant for fumarate, and diffusion coefficients of the substrates and metabolites. To estimate unknown parameters and calibrate the model, we used a custom designed biofilm reactor placed inside a nuclear magnetic resonance (NMR) microimaging and spectroscopy system and measured substrate utilization and metabolite production rates. From these data we estimated the yield coefficients, maximum substrate utilization rate, half saturation constant for lactate, stoichiometric ratio of fumarate and acetate to lactate and stoichiometric ratio of succinate to fumarate. These parameters are critical to predicting the activity of biofilms and are not available in the literature. Lastly, the model was used to predict uranium immobilization in S. oneidensis MR-1 biofilms by considering reduction and adsorption processes in the cells and in the EPS. We found that the majority of immobilization was due to cells, and that EPS were less efficient at immobilizing U. Furthermore, most of the immobilization occurred within the top 10 µm of the biofilm. To the best of our knowledge, this research is one of the first biofilm immobilization mathematical models based on experimental observation. It has the ability to predict the relative contributions to U immobilization of laEPS, bEPS and cells.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2023

Online monitoring the hydrolysis of uranium hexafluoride for intermediates by cryogenic layering and FTIR

Uranium hexafluoride (UF 6 ) is a commonly utilized material feedstock in uranium enrichment processes due to its high vapor pressure and ease of sublimation. When exposed to air, UF 6 undergoes spontaneous hydrolysis to form uranyl fluoride (UO 2 F 2 ) particulates which are utilized for the detection of undeclared nuclear activities by nuclear safeguards organizations. In this study, the kinetics of the hydrolysis reaction and how they relate to particle morphology of the product are still debated in the literature. Here, we report the direct, in situ observation of UF 6 reaction intermediates by cooling the reaction to cryogenic temperatures to significantly reduce the rate of hydrolysis. The reaction is then observable by Fourier transform infrared (FTIR) spectroscopy. The conversion of UF 6 to UOF 4 is observed as well as several other bands associated with possible long lived intermediate complexes. Chemometrics are used to further elucidate the reaction pathway from UF 6 to UO 2 F 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Expanding the Chemistry of Actinide Metallocene Bromides. Synthesis, Properties and Molecular Structures of the Tetravalent and Trivalent Uranium Bromide Complexes: (C 5 Me 4 R) 2 UBr 2 , (C 5 Me 4 R) 2 U(O-2,6- i Pr 2 C 6 H 3 )(Br), and [K(THF)][(C 5 Me 4 R) 2 UBr 2 ] (R = Me, Et)

The organometallic uranium species (C 5 Me 4 R) 2 UBr 2 (R = Me, Et) were obtained by treating their chloride analogues (C 5 Me 4 R) 2 UCl 2 (R = Me, Et) with Me 3 SiBr. Treatment of (C 5 Me 4 R) 2 UCl 2 and (C 5 Me 4 R) 2 UBr 2 (R = Me, Et) with K(O-2,6- i Pr 2 C 6 H 3 ) afforded the halide aryloxide mixed-ligand complexes (C 5 Me 4 R) 2 U(O-2,6- i Pr 2 C 6 H 3 )(X) (R = Me, Et; X = Cl, Br). Complexes (C 5 Me 4 R) 2 U(O-2,6- i Pr 2 C 6 H 3 )(Br) (R = Me, Et) can also be synthesized by treating (C 5 Me 4 R) 2 U(O-2,6- i Pr 2 C 6 H 3 )(Cl) (R = Me, Et) with Me 3 SiBr, respectively. Reduction of (C 5 Me 4 R) 2 UCl 2 and (C 5 Me 4 R) 2 UBr 2 (R = Me, Et) with KC 8 led to isolation of uranium(III) “ate” species [K(THF)][(C 5 Me 5 ) 2 UX 2 ] (X = Cl, Br) and [K(THF) 0.5 ][(C 5 Me 4 Et) 2 UX 2 ] (X = Cl, Br), which can be converted to the neutral complexes (C 5 Me 4 R) 2 U[N(SiMe 3 ) 2 ] (R = Me, Et). Analyses by nuclear magnetic resonance spectroscopy, X-ray crystallography, and elemental analysis are also presented.

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

Restoration Insights Gained from a Field Deployment of Dithionite and Acetate at a Uranium In Situ Recovery Mine

Mining uranium by in situ recovery (ISR) typically involves injecting an oxidant and a complexing agent to mobilize and extract uranium in a saturated ore zone. This strategy involves less infrastructure and invasive techniques than traditional mining, but ISR often results in persistently elevated concentrations of U and other contaminants of concern in groundwater after mining. These concentrations may remain elevated for an extended period without remediation. Here, we describe a field experiment at an ISR facility in which both a chemical reductant (sodium dithionite) and a biostimulant (sodium acetate) were sequentially introduced into a previously mined ore zone in an attempt to establish reducing geochemical conditions that, in principle, should decrease and stabilize aqueous U concentrations. While several lines of evidence indicated that reducing conditions were established, U concentrations did not decrease, and in fact increased after the amendment deployments. We discuss likely reasons for this behavior, and we also discuss how the results provide insights into improvements that could be made to the restoration process to benefit from the seemingly detrimental behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Production of small uranium dioxide microspheres for cermet nuclear fuel using the internal gelation process

The U.S. National Aeronautics and Space Administration (NASA) is developing a uranium dioxide (UO2)/tungsten cermet fuel for potential use as the nuclear cryogenic propulsion stage (NCPS). The first generation NCPS is expected to be made from dense UO2 microspheres with diameters between 75 and 150 m. Previously, the internal gelation process and a hood-scale apparatus with a vibrating nozzle were used to form gel spheres, which became UO2 kernels with diameters between 350 and 850 m. For the NASA spheres, the vibrating nozzle was replaced with a custom designed, two-fluid nozzle to produce gel spheres in the desired smaller size range. This paper describes the operational methodology used to make 3 kg of uranium oxide microspheres.

CERMET fuel↗
DOE OSTI2014

Molecular Dynamics Simulation of Thermodynamic Properties in Uranium Dioxide

In the present study, we investigated the thermodynamic properties of uranium dioxide (UO2) by molecular dynamics (MD) simulations. As for solid UO2, the lattice parameter, density, and enthalpy obtained by MD simulations were in good agreement with existing experimental data and previous theoretical predictions. The calculated thermal conductivities matched the experiment results at the midtemperature range but were underestimated at very low and very high temperatures. The calculation results of mean square displacement represented the stability of uranium at all temperatures and the high mobility of oxygen toward 3000 K. By fitting the diffusivity constant of oxygen with the Vogel-Fulcher-Tamman law, we noticed a secondary phase transition near 2006.4 K, which can be identified as a ‘‘strong’’ to ‘‘fragile’’ supercooled liquid or glass phase transition in UO2. By fitting the oxygen diffusion constant with the Arrhenius equation, activation energies of 2.0 and 2.7 eV that we obtained were fairly close to the recommended values of 2.3 to 2.6 eV. Xiangyu Wang, Bin Wu, Fei Gao, Xin Li, Xin Sun, Mohammed A. Khaleel, Ademola V. Akinlalu and Li Liu

Computer Simulation↗
DOE OSTI2015

Evaluation of Computed Tomography of Mock Uranium Fuel Rods at the Advanced Photon Source

This study discusses a multi-year effort to evaluate the utility of computed tomography at the Advanced Photon Source (APS) as a tool for non-destructive evaluation of uranium based fuel rods. The majority of the data presented is on mock material made with depleted uranium which mimics the x-ray attenuation characteristics of fuel rods while allowing for simpler handling. A range of data is presented including full thickness (5mm diameter) fuel rodlets, reduced thickness (1.8mm) sintering test samples, and pre/post irradiation samples (< 1mm thick). These data were taken on both a white beam (bending magnet) beamline and a high energy, monochromatic beamline. This data shows the utility of a synchrotron type source in the evealuation of manufacturing defects (pre-irradiation) and lays out the case for in situ CT of fuel pellet sintering. Finally, in addition data is shown from small post-irradiation samples and a case is made for post-irradiation CT of larger samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of molecular dynamics potential for uranium silicide fuels

Use of uranium–silicide (U-Si) in place of uranium dioxide (UO2) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. This is due to a higher thermal conductivity than UO2 that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. For example, silicide compounds (U3Si2 and U3Si) are known to become amorphous under irradiation. There was clear independent experimental evidence to support a crystalline to amorphous transformation in those compounds. However, it is still not well understood how the amorphous transformation will affect on fuel behavior. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In order to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters and increase the predictive capabilities of the developed model, inputs from atomistic simulations are essential. First-principles based density functional theory (DFT) calculations will provide the most reliable information. However, it is probably not possible to obtain kinetic information such as amorphization under irradiation directly from DFT simulations due to size and time limitations. Thus, a more feasible way may be to employ molecular dynamics (MD) simulation. Unfortunately, so far no MD potential is available for U-Si to discover the underlying mechanisms. Here, we will present our recent progress in developing a U-Si potential from ab initio data. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impact of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.

Accident tolerance fuel↗
DOE OSTI2018

Technical Basis for the Uranium Processing Facility Criticality Accident Alarm System

The technical basis for the Y-12 National Security Complex (Y-12) Uranium Processing Facility (UPF) Criticality Accident Alarm System (CAAS) 90% design has been completed. The CAAS design was evaluated for two primary processing facilities, the Main Processing Building (MPB) and the Salvage and Accountability Building (SAB), and for connected facilities known as the Personnel Support Building (PSB) and the Highly Enriched Uranium Materials facility (HEUMF) Connector (HCON), which connects UPF to existing Y-12 facilities. The evaluation focused on three primary areas to address requirements of ANSI/ANS-8.3: (1) detectability of the minimum accident of concern; (2) determination of areas where total dose could exceed 12 rad for the maximum anticipated accident yield; and (3) evaluation of the radiation tolerance of the detectors and associated control cabinets. The design was based on extensive analyses performed in MCNP6.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Uranium Accumulation Discoveries at Y-12 (Paper)

In June 2017, reduction operations at the Y-12 National Security Complex (Y-12) experienced a uranium accumulation event that compromised the nuclear criticality safety (NCS) of reduction operations. This event led to the execution of an extensive Extent-of-Condition (EOC) review aimed at identifying the location of other significant uranium masses in unintended locations. This paper describes the reduction process, the equipment in which unexpected accumulation was discovered, and the timeline of events leading to the discovery. It highlights the multi-disciplinary effort taken by production to recover from the event and resume operations. Lastly, it includes a description of the inconsistencies between operational conditions and Criticality Safety Evaluation (CSE) assumptions and how the EOC led to the discovery of four more accumulations within Y-12 processes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Acceptable Knowledge Summary Report for REMOTE-HANDLED DEPLETED URANIUM INGOTS FROM MATERIALS AND FUELS COMPLEX

This Acceptable Knowledge (AK) Summary Report has been prepared for the Central Characterization Program (CCP) for remote-handled (RH) transuranic (TRU) waste generated and managed by the Materials and Fuels Complex (MFC), formerly Argonne National Laboratory-West (ANL-W) and part of the Idaho National Laboratory (INL). The waste described in this report, depleted uranium ingots (waste stream ID-MFC-DU-INGOT), was historically generated in Building 765, the Fuel Conditioning Facility (FCF), formerly, Hot Fuel Examination Facility (HFEF)-South. This AK Summary Report, along with the referenced supporting documents, provides a defensible and auditable record of AK for the designated waste stream, depleted uranium ingots, produced from the FCF electrorefining process. The references and AK source documents used to prepare this report are listed in Sections 8.0 and 9.0 respectively. The source documents cited throughout this report are identified by alphanumeric designations corresponding to a unique Source Document Tracking Number (e.g., AKA01, C001, CCE01, DR001, M001, P001, and U001).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Cleanup under Airlock of an Old Uranium Foundry - 13273

Since 2004, AREVA's subsidiary SICN has been conducting the cleanup and dismantling of an old uranium foundry located in the town of Annecy (France). The first operations consisted in the removal of the foundry's production equipment, producing more than 300 metric tons (MT) of waste. The second step consisted in performing the radiological characterization of the 1,600 m{sup 2} (17,200 ft{sup 2}) building, including underground trenches and galleries. The building was precisely inventoried, based on operations records and direct measurements. All sub-surfaces, which needed to be cleaned up were characterized, and a determination of the contamination migration was established, in particular with trenches and galleries. The wall thicknesses to be treated were empirically justified, knowing that the maximal migration depth inside concrete is 5 mm for a liquid transfer vector. All singularities such as cracks, anchoring points, etc. were spotted for a complete and systematic treatment. Building structures not laying directly on the soil, such as floor slabs, were not cleaned up but directly deconstructed and disposed of as waste. The facility was located within the town of Annecy. Therefore, in order to avoid the risk of dusts dispersion and public exposure during the building deconstruction and the soil treatment, a third of the building's surface was confined in a sliding airlock built from a metal structure capable of resisting to wind and snow, which are frequent in this area. This particular structure provided a static confinement over the half of the building which was covered and a dynamic confinement using a ventilation and high efficiency air filtration system, sized to provide 2.5 air changes per hour. The enclosure and its metallic structure is 33 m long (108 feet), 25 m wide (82 feet), and 13 m high (42 feet), for a volume of 10,000 m{sup 3} (353,000 ft{sup 3}). It was made up of a double skin envelope, allowing the recycling of its structure and outside envelope. After cleaning up and dismantling the first portion of the building, the enclosure was repositioned on the second and the last third of the building, by sliding it on support pads. Almost 7,000 m{sup 2} of concrete surface has been treated with no dust dispersion outside the enclosure. After treatment, all the remaining surfaces were controlled by an independent entity to verify their acceptability with regards to residual contamination (less than 0.4 Bq/cm{sup 2} (24 DPM) for alpha contamination and less than 1 Bq/g of total uranium). Approximately 1,900 MT of equipment and waste were generated in batches of 1m{sup 3}, in order to be staged on site, and then characterized and packaged in 20 foot containers for shipment to the final ANDRA repository. The package certification included the verification of the physical and chemical characteristics and the radiological characteristics (mass activity, dose rate, and residual outside surface contamination). Finally, after cleanup and dismantling of the foundry, a concrete slab was poured on the free surface as a clean base for implementation of new activities. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗