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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 1,531 records · Page 85
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Analysis of Uranium Ion Exchange Resin from the 200 West Pump-and-Treat Facility

The 200 West Area Pump-and-Treat (P&T) Facility at the Hanford Site is a multifunctional groundwater treatment facility consisting of unit processes to remove radionuclides such as uranium (U) and technetium (Tc-99) as well as other inorganic (nitrate and hexavalent chromium), and organic (carbon tetrachloride and trichloroethene) contaminants. The RAD building of the P&T Facility removes U and Tc-99 using ion exchange chromatography. In 2016, the Purolite A530E resin used for the Tc-99 removal was replaced and the spent resin was analyzed. In addition to Tc-99, the Purolite A530E resin also retained I-129 and Co-60; however, the resin loading was dominated by common groundwater constituents such as calcium, iron, silicon, and sulfate.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

An Instrumented Uranium Hemispherical Casting

This work describes an instrumented uranium casting experiment and the analysis of the results. The casting is a hemisphere with inner radius of 75 mm and wall thickness of 10 mm. Casting was performed in a multi-coil vacuum induction melting (VIM) furnace with separate induction power supplies for each coil. One coil was used to heat the crucible and melt the charge material. The second coil was used to heat the top of the mold in order to preheat the mold and obtain the desired initial mold temperature. The current, voltage, and frequency of the mold coil was measured during mold heating. The mold filling rate was measured by a load cell was placed under the mold stack. The thermal history of the casting was measured by 7 thermocouples in the casting cavity and 22 thermocouples imbedded in the graphite mold. Contact pins were also located in the mold to attempt to measure gap formation in the metal-mold interface. This instrumentation provides an in-depth thermal history of the mold heat-up, mold filling, solidification, and cool down. The thermocouples were positioned to allow for the casting/mold interfacial heat transfer coefficient to be determined as a function of position and time. The casting was sectioned and metallography and chemical analysis was performed.

36 MATERIALS SCIENCE↗
DOE OSTI2013

In-Situ Leach Mining and Processing, Uranium

This unit process includes in-situ mining of uranium and associated processing activities, including milling and groundwater remediation. Data are based primarily on U.S. in-situ leach (ISL) mines. To access the model referenced in the report, please visit https://www.netl.doe.gov/energy-analysis/details?id=4625868C-66BB-4356-8EE6-93B9AC7DEEF3

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

X-Ray Diffraction Texture Analysis of Uranium Alloy Fuel PLN-5527 R0

This test plan describes XRD sample characterization for uranium alloys. This test plan includes preferred sample preparation information, characterization method, instrument parameters, calibration method, analysis software, data-acceptance requirements, documentation/reporting requirements. Texture measurements will be taken based on relative peak shapes (e.g., peak height, area, integrated intensity, etc.), relative to the individual XRD pattern.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Capture of Uranium Particles By Filtration: Effects of Filter Material and Air Humidity

Filtration of airborne particles is a standard technique for removal of radiological particles from the effluent of a nuclear reactor. The effects of humidity on filter performance and uranium particles are however less understood. We have conducted experimental runs in the Environmental Experimental Chamber (EEC) located at ORNL to study the capture of uranyl fluoride particles in supports of NNSA missions. Two filter types, one of Minimum Efficiency Reporting Value (MERV)-9 and the other of MERV-10 American Society of Heating, Refrigerating and Air-Conditioning Engineering) ASHRAE rating, were used in the investigations under two humidity conditions [15% and 65% Relative Humidity (Rh)]. Both MERV9 and MERV10 filters failed to capture uranyl fluoride particles that were smaller than 100 nm. The penetration efficiencies for both filter types were very high, greater than 80%, for particles from 10 to 100 nm. As particles became larger and wetter, the capturing efficiencies of both types of filter improved. Slightly greater than 50% of particles from 1 to 10 μm still escaped the capture by the MERV9 filters. However, MERV10 filters were able to capture higher than 50% of particles in the size range of 1 to 2 μm indicating it is likely to meet ASHRAE requirements. Overall, the capture efficiency of a filter is highly dependent on uranyl fluoride particle size, which humidity plays a major role. These two filter types simply are unable to capture particles smaller than 1 μm. A filter that has a higher ASHRAE rating like the HEPA filter has to be used to capture sub-micron particles. Questions on how uranyl fluoride behave in a condition higher than 65% Rh remain, because the particles could deliquescent at the higher humidity condition. That would complete change the size, shape, and surface properties of the uranyl fluoride particles which could lead to completely different filter performance and filtration dynamics. Furthermore, how high humidity affects filters of all MERV ratings remain unclear. Other filter characteristics such as surface charge could also be affected by high humidity condition and need to be investigated.

42 ENGINEERING↗
DOE OSTI2020

Dr. Ian Hutcheon Award: Spectroscopic Fingerprinting of Vapor-Phase Uranium Species

The Dr. Ian Hutcheon Post-Doctoral Fellowship provides resources to fund a TNFapplicable research project as well as promote mentorship, career development, and relevant training for the recipient. The two-year proposal incorporated a research plan to investigate uranium chemistry relevant to post-detonation nuclear forensics and a mentorship plan to address career development for the recipient in TNF-relevant areas.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Uranium Disposition Efforts at Los Alamos National Laboratory

Summary: Doing uranium work in a plutonium facility has challenges: NDA is geared to plutonium, Pu contamination, DU discards. The future volume of the feed stream to UED is dependent on pit disassembly. The dilute and dispose Pu disposition option will cause the HEU stream to grow significantly. Shipping HEU metal directly to Y-12 is a long term goal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

In-situ Electrochemical Testing of Uranium Oxide Alteration under Anoxic Conditions

Prediction of the corrosion behavior of spent uranium oxide (UO 2 ) fuel is needed for developing predictive performance assessment models for a geologic repository. Currently, the Fuel Matrix Dissolution Model (FMDM) is being used for modeling UO 2 chemistry in the Engineered Barrier System (EBS) and includes the effects of oxidants, oxygen and hydrogen peroxide generated from radiolysis of the fuel in contact with water, as well as the role of hydrogen on limiting corrosion. In this report, efforts to validate the FMDM model using in-situ liquid cells in the scanning electron microscopy (SEM) are described. The System for Analysis at the Liquid Vacuum Interface (SALVI), was invented at the Pacific Northwest National Laboratory (PNNL) and is a device that enables the investigation of solid-solution systems within the electron microscope. Initial efforts have concentrated on building the testing system for radioactive materials and exploring the use of the SALVI cell on UO 2 particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Characteristics of Uranium-Plutonium Mixed Oxide Nondestructive Assay Standards

A set of six well-characterized, homogeneous, mixed HEU-WGPu oxide standards has been manufactured for and are proposed for qualification as NDA working reference standards with SAFE-NMCA. These standards, if approved, will provide the basis for qualifying the FRAM isotopic systems for use in accountability measurements of 235U in the presence of Pu for unknown mixed U-Pu items by nondestructive assay. The U to Pu elemental mass ratios of the six analytical standards are nominally in the proportions 0.1, 0.25, 1.0, 2.5, 10, and 20; these values were chosen to span the U/Pu mass ratios of inventory items that ultimately will be measured by nondestructive assay and which currently reside in long term storage. The measured characteristics, including plutonium and uranium elemental masses and isotopic distribution, of each standard and relevant source documentation are reported here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

TEX-HEU: Integral Experiments with Highly Enriched Uranium and Polyethylene (IER-297 CED-3b Report)

This report documents the experimental critical configurations and measurements for IER-297, Thermal/Epithermal eXperiments (TEX) using highly enriched uranium (HEU) fuel, moderated and reflected by polyethylene. These experiments establish the baseline TEX-HEU configurations spanning thermal, intermediate, and fast fission energy regimes. The experiment campaign was completed in FY2020 at the National Critical Experiments Research Center at the Nevada National Security Site. The campaign produced six documented critical configurations (0”, ¹/₈” A, ¹/₈” B, ¹/₄”, ¹/₂”, and 1¹/₂”) and one reproducibility measurement (¹/₈” R). Table 1 summarizes these seven experiments along with their period and calculated reactivity, in order of execution. The Comet General Purpose Critical Assembly Machine was used to conduct the TEX-HEU experiments. The HEU fuel consists of U(93+) plates, collectively known as the “Jemima” plates. Each experiment consisted of an alternating stack of these HEU plates with or without polyethylene moderator plates. By varying the thicknesses of the polyethylene plates, the neutron spectrum of the experiment could be fine-tuned to a specific fission energy regime. The experiments were reflected by 1” of polyethylene consisting of a bottom and top plate and outer reflector rings to surround the HEU and polyethylene plates. The fine reactivity control of the system was provided by the thickness of the top reflector plate, which could be varied in increments of ¹/₃₂”.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Uranium Density, Thermal Conductivity, Specific Heat, and Thermal Diffusivity

Thermophysical properties for pure uranium are collected here as a common and maintained resource with a focus on casting. Source material details will be provided for decisions on fitting emphasis. The four material properties represented in this document were chosen based upon the definition of thermal diffusivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Uranium Immobility in Tims Branch Wetland

The objective of this study was to address an item identified by the Low-Level Waste Disposal Facility Federal Review Group (LFRG) panel during their review of the Savannah River Site’s Composite Analysis. Their concern was that the Tims Branch wetland, which has elevated levels of uranium (U) originating from site operations between 1952 to 1985, may be a secondary source term, posing a human health risk. This report briefly describes three studies dealing with: 1) quantifying the mass of U remaining in the wetland, 2) quantifying the tendency of U to release from Tims Branch sediment, and 3) describing the distribution of U in aqueous and suspended solids in Tims Branch water during episodic storm events. The Tims Branch wetland has been extremely effective for the last 50 years at immobilizing U released to the system. A detailed map including 270,000 gamma spectra of the U distribution in the wetland indicated that 84% of the initial 43,500 kg U introduced in the system remain there after 50 years (Kaplan et al., 2020). The effectiveness of immobilizing U is supported by stream water measurements collected over the last several decades indicating that U concentrations have been consistently below drinking water regulatory limits (i.e., the EPA Maximum Contaminant Level, 30 pg/L (SRNS, 2019)). Furthermore, using contaminated Tims Branch sediments, in situ desorption Kd values (ratio of U sediment/U Water) were extremely high, ranging from 104to 106 L/kg, suggesting the U is strongly bound to these wetland sediments. Finally, a detailed study of Tims Branch stream water properties during episodic rain events indicated that essentially all the U transported in this stream was associated with suspended solids (Batson et al., 1996). Together these findings indicate that the Tims Branch wetland is extremely effective in its present state at immobilizing U and that the small masses of U released downstream will be predominantly in association with suspended solids.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Protocols for Uranium Carbon Analysis

In 1998, MST-6 (now SIGMA) installed a Horiba EMIA-8200 Carbon/Sulfur analyzer to provide the ability to perform in-house chemical analysis on a variety of materials including uranium. While many of the results and procedures were detailed in program files, no compiled record of the overall analytic process was ever published. This retrospective utilizes existing logbooks, operator interviews, instrument manuals, and notes to catalog what is known about the past procedures and protocols for carbon analysis that were established and present the data informing those decisions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Protocols for Uranium Carbon Analysis - Testing Protocol Summary

In 1998, Sigma (then MST-6) purchased and installed a new carbon-sulfur analyzer in G105 as part of a substantial refresh of our analytic capabilities. Substantial effort was undertaken to establish analysis protocols for uranium and other materials. This summary describes the following protocols we put in place as a result of these studies: sample xize (for monolithic samples—chip and powder sample may need more current study), flux/accelerant recipe, and sample cleaning process.

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

Uranium-Titanium Alloys: The Effect of Ti Content on Phase Transformations, Microstructures, and Mechanical Properties

This report documents the findings of collaborative research done at Sandia National Laboratory and the Y-12 Development Division during the 1980s. Ongoing progress reports were presented at various U. S. Department of Energy (DOE) meetings throughout the 1980s, but the findings were never integrated into an inclusive final report. This report provides a summary of the key findings and implications of this research. This work was supported by the DOE National Nuclear Security Administration. The effect of titanium content on phase transformations, microstructures, and mechanical properties of uranium-titanium alloys has been investigated in alloys containing 0.3 wt.% to 2.0 wt.% Ti.

36 MATERIALS SCIENCE↗
DOE OSTI2016

A Compilation of Tensile Data for Quenched and Aged Uranium-Niobium Alloys

A compilation of tensile test data in uranium-niobium (U-Nb)binary alloys containing up to 10wt.% Nb is presented, capturing over900heat-treat conditions. It is based on a comprehensive survey of the literature and includes recent Los Alamos data, especially from long-term aging. The data are restricted to ambient-temperature, quasi-static strain-rate measurements. The emphasis is on controlled gamma-solutionize+quench and isothermal aging data from temperatures as high as 625oC and times as long as 10 years. Elastic, plastic (strength), and ductility properties reported include first-yield modulus, first-yield strength, second-yield strength, ultimate tensile strength,uniform elongation, total elongation, and reduction in area. Total elongation data were also converted to a common standard for ease in comparing data from different tensile specimen geometries. Material pedigree and machining damage condition are included among the metadata recorded here. With a view towards modeling these properties as a function of age(to be reported in future publications), the data were categorized according to the microstructure present (i.e., before or after the appearance of lamellar decomposition products) and the phenomenological time-trend of aging (initial softening, classic hardening, classic overaging-softening).

36 MATERIALS SCIENCE↗