Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion
No abstract available
Hickman, Robert↗
SEARCH · Engineering Papers
Results for “uranium Dioxide”
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
ENDF/B-VIII.1: Thermal Neutron Scattering Sublibrary
The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Benchmark Gap Assessment for the Manufacturing of High-Assay Low-Enriched Uranium Fuels
This document develops basic critical conditions for spheres—moderated and unmoderated, as well as reflected and unreflected—in consideration of nuclear criticality safety of a potential fuel production facility producing high-assay low-enriched uranium (HALEU) fuel of several different types like tristructural-isotropic (TRISO), uranium metal and alloys, oxide and non-metallic forms. In addition to spherical arrangements, TRISO particle manufacturing process–specific equipment is modeled as it would be for the criticality safety analysis. The objective is to develop representative systems that can then be used for comparison with existing benchmarks. SCALE/TSUNAMI is used to assess the similarity index between these systems to assess validation gaps for possible fuel production applications of proposed advanced reactors. Several different fuel types were evaluated, including TRISO, uranium metal, uranium molybdenum, uranium zirconium, uranium dioxide, uranium nitride, uranium hydride, U-ZrH, and uranium chloride. This selection of fuel types covers a breadth of proposed reactor types, as well as intermediate steps in the production and fabrication of the fuel
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Study of mechanical properties of uranium compounds
Study determines the mechanical properties, including brittleness and ductility of several uranium compounds. These include uranium dioxide, uranium sulfide, and uranium phosphide.
Beals, R. J.↗
Review of industry-proposed in-pile thermionic space reactors. Volume I - General
Diode and reactor design and nuclear fuels including uranium carbide alloys, uranium dioxide and uranium dioxide cermets for industry proposed in-pile thermionic space reactors
THERMIONIC REACTOR↗
Materials Data on UCO5 by Materials Project
UO2CO3 crystallizes in the orthorhombic Imm2 space group. The structure is zero-dimensional and consists of two carbonic acid molecules and two pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules.
36 MATERIALS SCIENCE↗
Materials Data on U(BO3)2 by Materials Project
UO2B2O4 crystallizes in the monoclinic C2/c space group. The structure is one-dimensional and consists of four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules and four BO2 ribbons oriented in the (0, 1, 0) direction. In each BO2 ribbon, B3+ is bonded in a trigonal planar geometry to three O2- atoms. There are a spread of B–O bond distances ranging from 1.34–1.40 Å. There are two inequivalent O2- sites. In the first O2- site, O2- is bonded in a bent 120 degrees geometry to two equivalent B3+ atoms. In the second O2- site, O2- is bonded in a distorted single-bond geometry to one B3+ atom.
36 MATERIALS SCIENCE↗
Materials Data on UCO5 by Materials Project
UO2CO3 crystallizes in the orthorhombic Pmmn space group. The structure is zero-dimensional and consists of two carbonic acid molecules and two pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules.
36 MATERIALS SCIENCE↗
Materials Data on U(HO2)2 by Materials Project
UO2(OH)2 crystallizes in the orthorhombic Cmce space group. The structure is zero-dimensional and consists of four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules and eight water molecules.
36 MATERIALS SCIENCE↗
Materials Data on U(HO2)2 by Materials Project
UO2(OH)2 crystallizes in the orthorhombic Cmc2_1 space group. The structure is zero-dimensional and consists of four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules and eight water molecules.
36 MATERIALS SCIENCE↗
Materials Data on UH14(NO2)4 by Materials Project
UO2(NH2O)2(NH2OH)2(H2O)2 crystallizes in the orthorhombic Pbca space group. The structure is zero-dimensional and consists of eight hydroxyammonium molecules; eight hydroxylamine molecules; four pitchblend, uranium(iv) oxide, uranium(iv) dioxide, uranium dioxide molecules; and eight water molecules.
36 MATERIALS SCIENCE↗
Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination
A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.
Collins, J. F.↗
Direct measurement of 5f delocalization with U XES
Delocalization of the 5f states in the early actinides in general and U metal in particular is significantly important and yet poorly understood. Here, extant spectroscopic techniques have failed to resolve the situation. Here it will be shown that X-Ray Emission Spectroscopy (XES) of the M 4,5 levels can provide the needed information, with a distinct difference between the delocalized U metal and localized uranium dioxide and uranium tetrafluoride cases. A Peak Ratio (PR) model, built upon electric dipole selection rules, has been developed and utilized, with quantitative agreement between experiment and theory. Possible expansion to other types of 5f mixing systems will be discussed.
36 MATERIALS SCIENCE↗
High Level Gap Analysis for Accident Tolerant and Advanced Fuels for Storage and Transportation
This initial gap analysis considers proposed accident tolerant fuel (ATF) options currently being irradiated in commercial reactors, since these are most likely for future batch implementation. Also, advanced fuel (AF) options that may be likely for use in advanced reactors are considered. The cladding technologies considered were chromium-coated zirconium-based alloys, FeCrAl, and both monolithic and matrix composite Silicide carbide (SiC). The fuel technologies considered were chromium-doped uranium dioxide fuel, uranium alloys, uranium nitride, and uranium silicide. Numerous national labs, industry, and countries are performing significant testing and modeling on these proposed technologies to establish performance, but at this time none of the prototypes being irradiated have achieved end-of-life (EOL) burnup. There are some testing results after one burnup cycle to verify in-reactor performance, but little data beyond that. As the ATF prototypes acquire more burnup, data will be produced that is relevant to storage and transportation. The DOE:NE Spent Fuel and Waste Science and Technology (SWFST) Storage and Transportation (ST) Control Account will evaluate the performance data as it becomes available for application to the identified gaps for ST.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Irradiation of refractory fuel compounds, U02 and UC, at high specific power to high burnups - Post-irradiation examination of capsule 1
Irradiation of refractory uranium dioxide and uranium carbide fuels at high specific powers to high burnups
IRRADIATION↗
Calculation of fuel loss from vented nuclear fuel elements for space power reactors
Rate of loss of uranium dioxide or uranium mononitride from vented nuclear fuel element
Gluyas, R. E.↗
Molybdenum-UO2 cermet irradiation at 1145 K.
Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.
Mcdonald, G.↗
Microfluidic Uranium Microspheres Production for TRISO and Advanced Fuel Concepts
The purpose of this project was to expand existing internal gelation sol-gel capabilities at PNNL to explore producing uranium dioxide spheres for potential use as fuel kernels in next generation Tri-isotropic (TRISO) particle fuel. This project expands on previous sol-gel efforts at PNNL by (1) increasing the size regime of sphere production from the micro-fluidic range to the milli-fluidic range, and (2) producing uranium spheres. The approach involved first scaling up the channel size of the fluidic system to the millimeter range, with radiation safety considerations in mind; testing and demonstration on non-radioactive surrogate material, cerium oxide; then transitioning to uranium production and finally optimizing system parameters. Commercially available fluidic chips in the desired size range were could not be found, therefore the project designed and fabricated a T-junction with 1mm channels for droplet production. Because the production process is temperature sensitive, prior efforts have involved performing droplet production in a lab freezer. To reduce radiological waste and footprint, two alternative chilling methods were explored using aluminum thermal beads as a chill bath and a custom aluminum block fit to reagent reservoir sizes. Both were successful in the cerium tests, however the aluminum block design outperformed the thermal bead bath and was further adapted for the radiological test and production run in the Radiochemical Processing Laboratory (RPL). Gelation trials were performed to determine an acceptable range of feed solution parameters for the uranium dioxide gels, characterized by R-values, which is determined by the ratio of uranium nitrate to Hexamethylenetetramine (HMTA)/urea in the feed solution. R-values ranging from 1.6-2 were examined in the gelation trials, with only the 1.8 condition being tested in production. The project was successful in demonstrating a proof of concept design for producing uranium dioxide spheres, however further optimization is needed to dial in production parameters and improve sphere quality and homogeneity.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗