Engineering Papers⌕ Search

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.

At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Hydrochlorination of Uranium Dioxide in a Molten Salt Mixture- Phase 2: Sparged Benchtop Reaction Vessel Experiment

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 Metatomic 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 2 (of 2) experimentation. Though Phase 1 provided evidence of successful hydrochlorination, Phase 2 highlighted the nuances and challenges associated with scaling the reaction. The target UCl 4 species generated by the hydrochlorination of UO 2 generates water, which consumes UCl 4 . Compared to the single datapoint obtained at the end of the Phase 1 experiments, data collected at multiple timepoints during Phase 2 experiments indicated a presumed maximum percent conversion of 23% for this system that is likely influenced by water generation. Engineering improvements might be able to increase the percent conversion, but not likely to the degree needed for successful implementation of the technology. Moisture also presents an issue for the Hastelloy C276 reaction vessel. Eventually, corrosion will compromise the integrity of the reaction vessel, and the corrosion products could potentially affect the hydrochlorination chemistry. The presence of corrosion products, which were not produced in the Phase 1 alumina crucible, can and have created challenges when analyzing samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Analysis and Design for Irradiation of High Power TRISO Fuel Compact Specimens in HFIR

Tristructural isotropic (TRISO) fuel is being proposed for use in several high-temperature advanced reactor concepts because of its structural integrity under high operating temperatures and burnup. One of these advanced reactor concepts is the Kairos Power fluoride salt-cooled high-temperature reactor (KP- FHR) under development by Kairos Power, LLC. Previous TRISO irradiation experiments were focused on qualification for high-temperature gas reactors (HTGRs), which have higher operating temperatures but lower particle powers than the KP-FHR design. To study the performance of TRISO fuel designed for HTGRs under prototypical FHR conditions, a set of experiments was designed using the MiniFuel irradiation vehicle at the Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR). The experiments will irradiate 30 TRISO-containing carbon matrix compacts at inner small vertical experiment facilities in HFIR. Each compact will contain 20 TRISO particles (600 particles total) developed for HTGRs, consisting of either 14% enriched uranium dioxide uranium carbide, naturally enriched uranium dioxide, uranium carbide, or 9.6% enriched uranium dioxide fuel kernels with time- and volume-averaged silicon carbide layer temperatures between 500 and 900°C. This report summarizes the vehicle designs that have been developed, as well as the neutronic and thermal analyses completed for these irradiation experiments. These analyses show that MiniFuel compact irradiation is a versatile experiment that can be used to study a range of TRISO particle powers and fuel types while providing reasonable separation of burnup and temperature effects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A microfluidic electrochemical cell for studying the corrosion of uranium dioxide (UO 2 )

We have developed a specialized microfluidic electrochemical cell that enables in situ investigation of the electrochemical corrosion of microgram quantities of redox active solids. The advantage of downscaling is the reduction of hazards, waste, expense, and greatly expanding data collection for hazardous materials, including radioactive samples. Cyclic voltammetry was used to monitor the oxidation–reduction cycle of minute quantities of micron-size uraninite (UO 2 ) particles, from the formation of hexavalent uranium (U(VI)), U 3 O 7 and reduction to UO 2+x . Reaction progress was also studied in situ with scanning electron microscopy. The electrochemical measurements matched those obtained at the bulk-scale and were consistent with ex situ characterization of the run products by X-ray photoelectron spectroscopy, scanning transmission electron microscopy, and atomic force microscopy; thus, demonstrating the utility of the microfluidic approach for studying radioactive materials.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
NASA NTRS1975

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Bowles, K. J.↗
DOE OSTI2025

Speciation mapping of the oxidation layer on aged uranium dioxide using scanning transmission x-ray spectromicroscopy

In this study, UO 2 was aged in humid air and prepared as a thin section using a focused ion beam (FIB) instrument. The specimen was measured using synchrotron radiation spectromicroscopy techniques at the Beamline 11.0.2 STXM end station of the Advanced Light Source (ALS). Non-negative matrix factorization (NMF) methods were used to identify and map three component x-ray absorption near-edge structure (XANES) spectra in the oxygen K-edge data, revealing a surface layer of U 4 O 9 with a thickness of 206 ± 21 nm, and the bulk of the sample remaining as UO 2 . Uranium N 4,5 -edge XANES spectromicroscopy supports these results. The diffusion-controlled parabolic rate constant for UO 2 oxidation to U 4 O 9 was calculated from the observed layer thickness and compared to literature values. Complementary transmission electron microscopy (TEM) was used to image the sample and identify the phases present in various regions, confirming the STXM results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2020

Criticality Safety and Fuel Performance Considerations for Enrichment Above 5 Weight Percent in the Uranium Dioxide Fuel Cycle

The U.S. Nuclear Regulatory Commission (NRC) is preparing for anticipated licensing applications requesting fuel enrichments of UO 2 ceramic fuels enriched in excess of 5 weight percent (w/o) 235 U for commercial light-water reactors (LWRs). PNNL has been tasked with providing technical assistance to the NRC related to the review and approval of requests for enrichment increases above the current 5 w/o 235 U enrichment limit. This report will provide the agency with technical assistance to enhance the staff’s knowledge base and to identify developmental needs required to support the licensing of greater than 5 w/o UO 2 ceramic fuels. This report provides a background on increasing enrichment of UO 2 LWR fuel beyond 5 w/o 235 U. The primary concern of increasing enrichment is the impacts to criticality safety. This report will focus on criticality safety as it relates to enrichment, conversion of uranium hexafluoride to UO 2 , fuel assembly fabrication, transportation of fresh/spent fuel assemblies, reactor operations, and storage of fresh/spent fuel assemblies. The report provides a general summary of each of the steps in the UO 2 fuel cycle starting from transportation of enriched UF 6 through dry cask storage of spent fuel assemblies. This report also discusses other impacts of increased enrichment during reactor operation and the surrounding fuel cycle activities. Current needs regarding both data and analytical codes to support the increase above 5 w/o 235 U are identified and discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Chromium-doped uranium dioxide fuels: A review

UO 2 doped with parts per million CR 2 O 3 powder is considered a potential near term accident tolerant fuel candidate. Here, the results of decades of industry and academic research into Cr-doped UO 2 are analyzed and their shortcomings are critiqued. Focusing on the incorporation mechanisms of Cr into the fuel matrix, we explore a mechanistic understanding of the characteristic properties of Cr-doped UO 2 , notably, enhanced fission gas retention attributed to enlarged grain sizes following sintering, along with marginal improvements in the thermophysical properties. The findings of recent X-ray Adsorption Near Edge Spectroscopy studies were compared and put into conversation with historic data regarding the incorporation of Cr in UO 2 . On the basis of defect mechanisms, the case is made for the substitutional incorporation of Cr governing the lattice solubility but not the enhanced U diffusivity. Instead, Cr/CR 2 O 3 redox chemistry in a well-defined oxygen potential explains the differences in the U diffusivity and O/M ratio. The primary mechanism of doping enhanced grain growth is found to be liquid assisted sintering due to a CRO (1) eutectic phase at the grain boundaries. The role of inhomogeneities in Cr concentration in UO 2 at various length scales across the materials microstructure is highlighted and connected to promising experimental and modeling work to fill in the gaps in the current understanding of Cr-doped UO 2 . The review considers both the open scientific questions and engineering applications to illustrate the deep connections between the practice and theory in the design of accident tolerant nuclear fuels. In conclusion, the review ends with an outline of future works that combine meticulous irradiation studies and high resolution experiments with next generation modeling and simulations techniques empowered by machine learning advances to accelerate the fabrication and adoption of Cr-doped UO 2 light water reactors.

Cleveland, Mack Wesley [Massachusetts Inst. of Tec↗
DOE OSTI2020

Fracture Strength Determination Methods for Ceramic Materials Applied to Uranium Dioxide

The Advanced Fuels Campaign is currently focusing on development of accident-tolerant fuels that possess a range of property modifications intended to improve fuel performance during accident and transient conditions as well as extend license limits to burnups beyond 62 MWd/kgU. Both of these drivers have identified understanding and mitigating fuel cracking as a key performance criterion. Small-scale cantilever beam testing has been identified as a plausible method by which to collect fracture data for UO 2 as a function of chemical and structural evolutions introduced either during fabrication or irradiation. While this method has been found to be capable of providing data that are in reasonable agreement with literature for unirradiated UO 2 , the inherently small sample volumes that can be sampled limit its ability to capture the statistical nature of mechanisms that govern the fracture of brittle ceramics. A biaxial flexure strength test was developed to be used for unirradiated UO 2 . This method is standard in the community, but no systems presently in use at national laboratories, universities, or private companies are available to be used for nuclear fuel materials. This report describes the operation and benchmarking of this system, which has been validated for a number of common oxides. Future work will extend this system to characterization of both doped UO 2 and other relevant microstructural modifications that can also be measured using small-scale cantilever beam testing. Comparison of datasets collected using both methods will allow the overall applicability of small-scale techniques to be assessed and provide confidence or bounds on their use for irradiated fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

Diianni, D. C.↗
NASA NTRS1973

Theoretical analysis of swelling characteristics of cylindrical uranium dioxide fuel pins with a niobium - 1-percent-zirconium clad

The relations between clad creep strain and fuel volume swelling are shown for cylindrical UO2 fuel pins with a Nb-1Zr clad. These relations were obtained by using the computer code CYGRO-2. These clad-strain - fuel-volume-swelling relations may be used with any fuel-volume-swelling model, provided the fuel volume swelling is isotropic and independent of the clad restraints. The effects of clad temperature (over a range from 118 to 1642 K (2010 to 2960 R)), pin diameter, clad thickness and central hole size in the fuel have been investigated. In all calculations the irradiation time was 500 hours. The burnup rate was varied.

Saltsman, J. F.↗
NASA NTRS1973

Uranium dioxide fuel cladding strain investigation with the use of CYGRO-2 computer program

Previously irradiated UO2 thermionic fuel pins in which gross fuel-cladding strain occurred were modeled with the use of a computer program to define controlling parameters which may contribute to cladding strain. The computed strain was compared with measured strain, and the computer input data were studied in an attempt to get agreement with measured strain. Because of the limitations of the program and uncertainties in input data, good agreement with measured cladding strain was not attained. A discussion of these limitations is presented.

Smith, J. R.↗
NASA NTRS1974

Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

Diianni, D. C.↗