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At least 775 records · Page 43

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NASA NTRS1976

Vacuum ultraviolet spectra of uranium hexafluoride/argon mixtures

The transmission properties of room temperature helium at pressures up to 20 atmospheres were determined in the wavelength range from 80 to 300 nm. Similarly, the transmission properties of uranium hexafluoride at 393 K (pressures less than 1.0 mm) were determined in the wavelength range from 80 to about 120 nm. The results show that high pressure helium is sufficiently transparent in the vacuum ultraviolet region (provided trace contaminants are removed) to be utilized as a transparent purge gas in future fissioning gaseous uranium plasma reactor experiments. Absorption cross sections for uranium hexafluoride were calculated from the data between 80 and 120 nm and were of the order of 10 to the -17 power sq cm.

Krascella, N. L.↗

NASA NTRS1976

Properties of radio-frequency heated argon confined uranium plasmas

Pure uranium hexafluoride (UF6) was injected into an argon confined, steady state, rf-heated plasma within a fused silica peripheral wall test chamber. Exploratory tests conducted using an 80 kW rf facility and different test chamber flow configurations permitted selection of the configuration demonstrating the best confinement characteristics and minimum uranium compound wall coating. The overall test results demonstrated applicable flow schemes and associated diagnostic techniques were developed for the fluid mechanical confinement and characterization of uranium within an rf plasma discharge when pure UF6 is injected for long test times into an argon-confined, high-temperature, high-pressure, rf-heated plasma.

Source record↗

NASA NTRS1976

Uranium plasma emission at gas-core reaction conditions

The results of uranium plasma emission produced by two methods are reported. For the first method a ruby laser was focused on the surface of a pure U-238 sample to create a plasma plume with a peak plasma density of about 10 to the 20th power/cu cm and a temperature of about 38,600 K. The absolute intensity of the emitted radiation, covering the range from 300 to 7000 A was measured. For the second method, the uranium plasma was produced in a 20 kilovolt, 25 kilojoule plasma-focus device. The 2.5 MeV neutrons from the D-D reaction in the plasma focus are moderated by polyethylene and induce fissions in the U-235. Spectra of both uranium plasmas were obtained over the range from 30 to 9000 A. Because of the low fission yield the energy input due to fissions is very small compared to the total energy in the plasma.

Williams, M. D.↗

NASA NTRS1978

Sputtering of uranium

Results are presented for an experimental study of the sputtering of U-235 atoms from foil targets by hydrogen, helium, and argon ions, which was performed by observing tracks produced in mica by fission fragments following thermal-neutron-induced fission. The technique used allowed measurements of uranium sputtering yields of less than 0.0001 atom/ion as well as yields involving the removal of less than 0.01 monolayer of the uranium target surface. The results reported include measurements of the sputtering yields for 40-120-keV protons, 40-120-keV He-4(+) ions, and 40- and 80-keV Ar-40(+) ions, the mass distribution of chunks emitted during sputtering by the protons and 80-keV Ar-40(+) ions, the total chunk yield during He-4(+) sputtering, and some limited data on molecular sputtering by H2(+) and H3(+). The angular distribution of the sputtered uranium is discussed, and the yields obtained are compared with the predictions of collision cascade theory.

Gregg, R.↗

NASA NTRS1985

Lead and uranium group abundances in cosmic rays

The importance of Lead and Uranium group abundances in cosmic rays is discussed in understanding their evolution and propagation. The electronic detectors can provide good charge resolution but poor data statistics. The plastic detectors can provide somewhat better statistics but charge resolution deteriorates. The extraterrestrial crystals can provide good statistics but with poor charge resolution. Recent studies of extraterrestrial crystals regarding their calibration to accelerated uranium ion beam and track etch kinetics are discussed. It is hoped that a charge resolution of two charge units can be achieved provided an additional parameter is taken into account. The prospects to study abundances of Lead group, Uranium group and superheavy element in extraterrestrial crystals are discussed, and usefulness of these studies in the light of studies with electronic and plastic detectors is assessed.

Yadav, J. S.↗

NASA NTRS2005

X-Ray Emission from "Uranium" Stars

The project aims to secure XMM observations of two targets with extremely low abundances of the majority of heavy elements (e.g., log[Fe/H] $\sim$-4), but that show absorption lines of uranium. The presence of an r-process element such as uranium requires a binary star system in which the companion underwent a supernova explosion. A binary star system raises the distinct possibility of the existence of a compact object, most likely a neutron star, in the binary, assuming it survived the supernova blast. The presence of a compact object then suggests X-ray emission if sufficient matter accretes to the compact object. The observations were completed less than one year ago following a series of reobservations to correct for significant flaring that occurred during the original observations. The ROSAT all-sky survey was used to report on the initial assessment of X-ray emission from these objects; only upper limits were reported. These upper limits were used to justify the XMM observing time, but with the expectation that upper limits would merely be pushed lower. The data analysis hinges critically on the quality and degree of precision with which the background is handled. During the past year, I have spent some time learning the ins and outs of XMM data analysis. In the coming year, I can apply that learning to the analysis of the 'uranium' stars.

Schlegel, Eric↗

DOE OSTI2014

Supercritical Fluid Extraction and Separation of Uranium from Other Actinides

This paper investigates the feasibility of separating uranium from other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of an extraction and counter current stripping technique, which would be a more efficient and environmentally benign technology for used nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U(VI), Np(VI), Pu(IV), and Am(III)) were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, the separation of uranium from plutonium in sc-CO2 modified with TBP was successful at nitric acid concentrations of less than 3 M in the presence of acetohydroxamic acid or oxalic acid, and the separation of uranium from neptunium was successful at nitric acid concentrations of less than 1 M in the presence of acetohydroxamic acid, oxalic acid, or sodium nitrite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

A Multi-Energy Approach for Neutron Interrogation of Uranium Items

Bulk uranium items are often measured using active neutron interrogation systems to take advantage of the high penetrability of neutrons, providing the ability to quickly and accurately measure effective uranium masses in large, dense configurations. These active techniques employ an external neutron source to induce fission in the uranium and subsequently measure emitted prompt fission or delayed neutrons. Unfortunately, the emitted neutrons from 235U and 238U are, for all practical purposes, indistinguishable; therefore, common systems such as the Active Well Coincidence Counter and other systems based on measurement of prompt or delayed fission neutrons require many representative calibration standards or known isotopic information to interpret the results, thus limiting the application of these techniques for safeguards purposes. This paper presents the motivation for a multi-energy neutron interrogation technique which aims to improve the accuracy of these NDA techniques through the use of multiple neutron interrogation sources. Potentially, the dual measurement approach will allow the mass values of the isotopes in the item to be independently determined while minimizing the number of representative standards needed and/or eliminating the need for prior knowledge of the isotopic information. This work examines the change in fission rates with differing interrogating neutron energy and, therefore, the potential of this new technique to independently measure the masses of 235U and 238U in the measurement item. Initial simulation and measurement results will be presented.

Hayward, Jason P.↗

DOE OSTI2019

A Metaheuristic Optimization Tool for High Flux Isotope Reactor Low-Enriched Uranium Core Design

A metaheuristic optimization tool was developed for design optimization studies of a High Flux Isotope Reactor low-enriched uranium core. Low-enriched uranium core design presents a multidimensional, multi-objective optimization problem that requires a systematic approach to avoid arbitrary design decisions. This tool searches this multidimensional design space to generate an optimum design, considering a weighted design performance metric that includes several parameters. The search methods are discussed herein, and the tool is applied to uranium silicide fuel designs.

Variansyah, I.↗

DOE OSTI2019

Method of separating and recovering uranium from aluminum-clad metallic nuclear fuel

According to one aspect of the invention, a method for separating and recovering uranium from a nuclear fuel element. The method includes immersing a nuclear fuel element containing nuclear fuel and cladding in a molten metal. The nuclear fuel includes uranium. The cladding is selectively dissolved from the nuclear fuel element when immersed in the molten metal. The nuclear fuel is separated from the cladding. The method then includes loading the nuclear fuel into a permeable basket that is electrically configured as an anode of an electrolytic cell. There are also a molten salt electrolyte and a cathode in the electrolytic cell. Then, the method includes applying an electric charge across the electrolytic cell. The molten salt electrolyte selectively transfers uranium from the anode to the cathode.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Nuclear Criticality Safety Lessons Learned in the Design of the Uranium Processing Facility at the Y-12 National Security Complex

The Uranium Processing Facility (UPF) is being built by the National Nuclear Security Administration (NNSA) to replace aging uranium processes at the Y-12 National Security Complex (Y-12 NSC) in Oak Ridge Tennessee. Specifically, the processes housed in building 9212 are being replaced as this facility dates to the World War II Manhattan Project days of Y-12. When the UPF Project is completed it will consist of three main facilities along with several support facilities. The Main Processing Building (MPB) will house oxide production and metal casting operations. The Salvage and Accountability Building (SAB) will house wet chemistry operations (examples include recovery evaporation and calcination of solutions). The Personnel Support Building will be connected to both the SAB and the MPB and will serve as a transition facility for materials and personnel between the two. The MPB will also be connected to the Highly Enriched Uranium Materials Storage Facility (HEUMF) via a connector (HCON) to permit transfers from storage into UPF directly. This paper will summarize some of the more important lessons learned for the NCS practitioner working in a design environment rather than in a production or experimental environment. As the Chief NCS Engineer for the UPF Project I have been responsible for ensuring the technical evaluations produced on the project are high quality, technically accurate, and conform to the Y-12 NCS Program expectations. Ensuring nuclear criticality safety for processes that do not yet exist in reality is a challenge very different than that of ensuring subcriticality for processes in an operating facility. The biggest difference being that in design space – everything is imaginary whereas in the operating facility you have the benefit of being able to go out and look at the equipment and interview the operators. Ensuring that Nuclear Criticality Safety is integrated into design is a process that requires constant communication between the NCS staff and the broader engineering design team. Until the facility is constructed – nothing is physically stable and so rigorous documentation of decisions – to include any assumptions – is necessary to minimize errors. Non-NCS engineers will not understand the hazard or the physics of nuclear criticality. It is up to the NCS staff to provide training and communication to the team. The NCS staff must be patient and prepared to explain themselves many times and be as flexible as possible.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2019

Simulating Uranium(VI) Diffusion in Sodium-Montmorillonite as a Function of 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 in these systems. However, their prediction is complicated by the complex mineralogical structure of montmorillonite clay, leading to two types of surface sites and clay porosities, and by the dependence of U(VI) solution speciation on chemical 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, at the high degrees of clay compaction expected in future barrier systems, the ability of U(VI) species to access small, highly negatively charged interlayer spaces may ultimately control diffusive fluxes. In this modeling study, we investigate how the charge of U(VI) solution species affects U(VI) sorption and diffusion behavior. For this purpose, we simplified complex chemical solution speciation diagrams to charge speciation diagrams. Furthermore, we coupled an existing U(VI) sorption model (in PHREEQC) with U(VI) diffusion calculations (in GLE) to simulate U(VI) mobility in lab-scale through-diffusion experiments over various chemical conditions. Our results suggest that there are two important parameters that govern U(VI) diffusion in these systems. Uranium(VI) surface complexation on clay edge surfaces leads to contaminant retardation, while the access of U(VI) solution species to diffuse layers in clay interlayer spaces controls diffusive fluxes. Hence on the field scale, these parameters will determine how early a plume may cross site boundaries and pose risks to drinking water resources, and the levels of U(VI) concentrations to be expected in contaminated waters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Polymers grafted with organic phosphorous compounds for extracting uranium from solutions

Complexing or chelating agents that offer strong, selective bonding with uranium as well as a broad pH range of effectiveness, specifically including the pH range around 8.2, together with the acrylic double bonds required for radiation-induced grafting on polymers to remove uranium from a solution such as seawater. The novel adsorbing species are phosphorus-containing molecules, in particular organic phosphates, phosphonates and phosphoric acids. Organic phosphorus compounds, for example, organic phosphates, phosphonates, and phosphoric acids, are attached to polymer fibers to form fibers, fiber fabrics or membranes that are effective, or show activity, in uranium adsorption.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

MULTIPHYSICS-MODELING OF FIRE-INDUCED URANIUM AEROSOL FORMATION – A-POSTERIORI BENCHMARKING OF EXPERIMENTS

The formulation and solution of a model that properly described the temperature profiles of uranium in fire conditions was prepared, verified and validated, to support a-posteriori benchmarking of historical experiments. Data from the multi-physics model combined with data regression from the experiments provides a useful tool for exploring sources of potential bias in historical experiments. The approaches used are augmented by visual observations and photographic evidence allowing for correlations between various phenomena, independent of whether or not that specific phenomena are part of the validated model. In this manner, the validated model is a tool for a-posteriori benchmarking, and not a model designed to replicate the experimental output exactly for any specific experiment. For experiments like the Clark (2015) experiments, this integrated model (including chemistry) demonstrated that the largest bias was associated in the metallurgy of the coupons. The model developed does serve as a basis for defending which of the historical alloys tested are the most appropriate for any other alloy being considered (e.g., one that has not yet been tested). The model identifies key reactions that impact uranium chemistry and can significantly bias experimental results. Through use of the model and energy balances, data from historical experiments by Elder and Tinkle (1980) were regressed and benchmarked. This study identifies that all uranium alloys are not equal and that there are distinct differences between the dominant alloys representing distinct metallurgical phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

High-Assay Low Enriched Uranium Pyroprocessing and Electrometallurgical Treatment at the Fuel Conditioning Facility

Fuel Conditioning Facility (FCF) fulfills part of INL’s mission by reprocessing irradiated sodium-bonded fuel from the EBR-ll reactor and participates in research to further explore the fuel cycle for recovery of fuel from a variety of current reactors in use today. The fuel cycle of a nuclear reactor under the current non-proliferation act, leaves unused fuel in the spent fuel rods. The rods are safely stored until their fuel can be recovered. ? Fuel recovery from the EBR-II fuel pins has several steps to separate the uranium from the salts. The pyroprocessing technology involves high - temperatures, chemical and electrochemical methods for separating the unused uranium from the salts, fission products, and used fuel. The recovered uranium is then remixed and formed into ingots or Regulus with an enrichment of < 20% U-235. This final product is a source to fuel current nuclear reactors as well as research and development for future nuclear reactors

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2023

Methods that purify uranium

A method for purifying uranium includes forming primary uranyl peroxide precipitates (UO 2 O 2 .4H 2 O). Forming the primary uranyl peroxide precipitates includes obtaining impure uranium dissolved in an acidic solution, evaporating the acidic solution to increase uranium concentration and to form a concentrated solution, mixing a hydrogen peroxide (H 2 O 2 ) solution with the concentrated solution in a first container, and forming uranyl peroxide precipitates in the first container. The method includes collecting the uranyl peroxide precipitates and washing and drying the uranyl peroxide precipitates. The method also includes converting the washed and dried uranyl peroxide precipitates into triuranium octoxide (U 3 O 8 ).

Du, Miting↗

DOE OSTI2017

Use of NAA to Detect Uptake of Uranium and other Elements in Adsorbents

A consortium of national lab and university partners, led by Oak Ridge (ORNL) and Pacific Northwest (PNNL) National Laboratories, has been working on the development of fibers with the ability to passively adsorb uranium out of seawater. In experiments by PNNL, the adsorption capacity of these fibers was observed to decrease by up to 30% as an effect of marine biofouling. In order to increase the cost-effectiveness of the process, the project ultimately aims to mitigate the effect of fouling by adding antimicrobial agents such as Ag nanoparticles to the adsorbents. To work towards this goal, a method of producing fiber surrogates for the specialized uranium adsorbents was established. This enabled researcher to complete fouling trials at a faster rate and lower cost. Additionally, the project has proven Neutron Activation Analysis (NAA) to be a viable process for measuring the uptake in fibers. The technique stands out because of its ability to measure low concentrations of uranium and other elements with high sensitivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Technology Development of a High-Capacity High-Assay Low Enriched Uranium Transportation Concept

This paper discusses the technology development (TD) efforts that led to the development of a High-Capacity High-Assay Low Enriched Uranium Transportation (HALEU) transportation concept. In 2018, the Department of Energy (DOE) Office of Nuclear Technology and Research Development tasked Idaho National Laboratory (INL) to investigate strategies to transport large quantities of HALEU. To complete this task, INL collaborated with Pacific Northwest National Laboratory and Oak Ridge National Laboratory. The project was completed in 2020, and one of the project outcomes was a transportation concept that consisting of five individual Type B packages transported on a single legal-weight truck (LWT). The total payload capacity of this concept is 1,881 kg (4,149 lb) of HALEU in the form of uranium dioxide (UO2) powder. The concept utilizes an existing Type B packaging design carrying a novel fuel basket design with an incorporated flux trap. The basket can be loaded with 18 individual fuel canisters. The research collaboration investigated the U.S. certification potential of this concept. This part of the project included evaluations of criticality safety, radiological safety, thermal safety, structural integrity, and confinement under hypothetical accident scenarios of transport. The results of these evaluations demonstrated a promising potential for U.S. certification of this concept. Eventually, the described efforts led to the pursuance and issuance of a U.S. patent, thus, protecting the associated intellectual property (IP). Current short-term goals include making this IP available to private industry partners for licensing, directly supporting DOE’s objectives of accelerating commercialization of national laboratory-generated IP. If additional funding becomes available, long-term research goals could include exploring the feasibility of transporting other uranium chemical forms (e.g., UF4) with this concept, or refining operational procedures to load or unload the packagings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗