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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Microstructural and Oxidation Effects of Nb Additions to U3Si2

U3Si2 is a long term, accident-tolerant nuclear fuel candidate for light-water reactors because of its superior thermal conductivity and increased uranium density when compared to traditional uranium dioxide (UO2). While reducing internal thermal stresses and increasing efficiency, U3Si2 exhibits energetic oxidation during certain off-normal and accident scenarios, which include coolant or steam exposure. To mitigate this, Nb is investigated as an alloy constituent to enhance corrosion resistance and increase mechanical strength. The work presented investigates the response of Nb-alloyed U3Si2 to steam atmospheres. A thermogravimetric analysis is conducted in flowing steam to T > 1000 °C to assess oxidation resistance. The phase characterization of as-melted, thermally annealed and post-oxidation compositions with up to 12 vol% Nb by powder X-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy is reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Isocyanide adducts of tri- and tetravalent uranium metallocenes supported by tetra(isopropyl)cyclopentadienyl ligands

In this work, reaction of the uranium( III ) metallocenium salt [(Cp iPr4 ) 2 U][B(C 6 F 5 ) 4 ] with tert-butyl isocyanide ( t BuNC) yielded the dicationic uranium( IV ) complex [(Cp iPr4 ) 2 U(CN t Bu) 4 ][B(C 6 F 5 ) 4 ] 2 (1), which displays a linear metallocene geometry. Use of crude mixtures of [(Cp iPr4 ) 2 U][B(C 6 F 5 ) 4 ], which contain a soluble source of iodide, led instead to isolation of the monocationic uranium( IV ) iodide complex [(Cp iPr4 ) 2 U(I)(CN t Bu) 2 ][B(C 6 F 5 ) 4 ] (2). Adduct formation with no change in oxidation state was observed upon addition of t BuNC to the neutral uranium( III ) species (Cp iPr4 ) 2 UI, resulting in isolation of (Cp iPr4 ) 2 U(I)(CN t Bu) (3). X-ray crystallographic and IR spectroscopic studies both showed effects ascribed to the presence of multiple strongly donating isocyanide ligands in 1.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Photodissociation and Infrared Spectroscopy of U + (CO 2 ) n , UO + (CO 2 ) n , and UO 2+ (CO 2 ) n Cation-Molecular Complexes

Laser vaporization of uranium in a pulsed supersonic expansion of carbon dioxide is used to produce complexes of the form U + (CO 2 ) n , UO + (CO 2 ) n , and UO 2+ (CO 2 ) n . These ions are selected in a reflectron time-offlight mass spectrometer and studied with visible laser photodissociation and tunable infrared laser photodissociation spectroscopy in the region of the CO 2 antisymmetric stretch. The dissociation patterns and spectroscopy of these ions indicate that CO 2 ligands are intact molecules. Although reaction products that form oxide-carbonyl or oxalate species are predicted to be stable, there is no direct evidence in the frequency range studied for the formation of these species. There is no clear indication for the coordination numbers for singly charged uranium and its oxide complexes with CO 2 . However, there is strong support in the vibrational patterns for an eight-coordinate complex of the doubly charged UO 2+ species, i.e., UO 2+ (CO 2 ) 8 .

Cluster chemistry↗
DOE OSTI2026

Development of a Chlorine Cathode and Anode Basket Assembly for Production of Uranium Chloride

A chlorine cathode has been developed for in situ chlorination of metals, oxides, and oxychlorides in molten chloride electrolytes that could be used to support synthesis of chloride fuel salts for molten salt reactors. The chlorine cathode is designed to electrochemically reduce chlorine gas to generate chloride ions that, when paired with a metal anode, chlorinate that metal as it is oxidized into the salt. The designed porous carbon electrode effectively distributes Cl 2 to the electrode surface and efficiently generates Clions in the molten chloride electrolyte. This report highlights recent improvements made to the electrode assembly, with a focus on operational control of the anode basket stability. Synthesis tests demonstrated the successful chlorination of uranium metal, resulting in 3.6 wt% uranium generated in the LiCl-KCl base salt in 45 minutes, performed in a bench-scale chlorination apparatus. Higher concentrations could be achieved by applying longer chlorination times or increasing the amount of uranium loaded in the anode basket. Overall, the chlorine cathode can be used with an appropriately designed anode to chlorinate uranium in situ for synthesis of molten salt reactor fuel salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of metallic nuclear material purification process via simultaneous chlorination and volatilization

Here, a purification process involving simultaneous chlorination and volatilization is reported using aluminum, iron, gallium, and uranium separation from cerium metal as the demonstration system. Anhydrous Cl 2 gas was reacted with the metal after it had been converted to small particles of hydride at temperatures ranging from 523 to 973 K. Each of the impurities should form chlorides with relatively high saturated vapor pressures. The objective of the process was to volatilize these chlorides after first achieving near-complete conversion of the metals to hydride and then chlorides. Complete conversion of hydrides to chlorides could not be achieved using relatively low-grade (99.5%) Cl 2 . High decontamination factors for all of the metal impurities required near complete conversion to chloride. It was concluded that oxygen contamination of the feed gas stream likely limits the conversion to chlorides by forming oxides instead, which then limits the volatilization of contaminants. Uranium was particularly challenging to remove in experiments suspected of having high oxygen contamination in the feed gas stream. After switching to ultra-high purity (99.999%) Cl 2 , complete conversion to chlorides within the measurement error was possible for the entire temperature range studied. Decontamination factors for aluminum, iron, gallium, and uranium were found to have average values of 1.9, 9.8, 14, and 5, respectively, at 973 K. While the process was most extensively studied using a once-through gas flow, recycling unreacted Cl 2 gas succeeded at minimizing waste while still maintaining complete conversion and similar decontamination factors.

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

Coated fuel pellets with enhanced water and steam oxidation resistance

Disclosed herein is a method comprising coating a fissile, uranium-containing ceramic material with a water-resistant layer, the layer being non-reactive with the fissile, uranium-containing ceramic material. The coating is applied to a surface of the fissile, uranium-containing ceramic material. Also disclosed is a fuel for use in a nuclear reactor.

Lahoda, Edward J.↗
DOE OSTI2026

Design Rules for Carboborothermic Reduction Synthesis of High Uranium Density UB 4 –UBC Composites

Uranium borides are promising candidate fuel forms for use in advanced nuclear reactors due to their high thermal conductivity and potential for dual use as both fuel and burnable absorber. In this work, uranium tetraboride () and uranium monoboroncarbide (UBC) composite were synthesized by using industrially scalable carboborothermic reduction method. The final uranium boride phase composition is sensitive to the sample holding crucibles ( and graphite) such that graphite supply excess carbon, promoting the formation of a predominant UBC phase. The high‐temperature in situ synchrotron X‐ray diffraction of pristine –UBC show persistence , UBC, and phases while preoxidized –UBC leads to predominant and formation due to progressive oxidation and boron loss at high temperature. The oxidation behavior was further characterized using thermogravimetric analysis, allowing direct comparison with other potential accident tolerant fuels such as , , UC, and UN. The –UBC shows higher uranium loading than monolithic and demonstrates promising oxidation behavior at high temperature, pointing to its potential as an improved uranium boride‐based fuel form.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Thermal Energy Transport in Oxide Nuclear Fuel

To efficiently capture the energy of the nuclear bond, advanced nuclear reactor concepts seek solid fuels that must withstand unprecedented temperature and radiation extremes. In these advanced fuels, thermal energy transport under irradiation is directly related to reactor performance as well as reactor safety. The science of thermal transport in nuclear fuel is a grand challenge as a result of both computational and experimental complexities. Here we provide a comprehensive review of thermal transport research on two actinide oxides: one currently in use in commercial nuclear reactors, uranium dioxide (UO 2 ), and one advanced fuel candidate material, thorium dioxide (ThO 2 ). In both materials, heat is carried by lattice waves or phonons. Crystalline defects caused by fission events effectively scatter phonons and lead to a degradation in fuel performance over time. Bolstered by new computational and experimental tools, researchers are now developing the foundational work necessary to accurately model and ultimately control thermal transport in advanced nuclear fuels. We begin by reviewing research aimed at understanding thermal transport in perfect single crystals. The absence of defects enables studies that focus on the fundamental aspects of phonon transport. Next, we review research that targets defect generation and evolution. Here the focus is on ion irradiation studies used as surrogates for damage caused by fission products. We end this review with a discussion of modeling and experimental efforts directed at predicting and validating mesoscale thermal transport in the presence of irradiation defects. While efforts in these research areas have been robust, challenging work remains in developing holistic tools to capture and predict thermal energy transport across widely varying environmental conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

New reference materials for trace-levels of actinide elements in plutonium

In this report two plutonium oxides were prepared as unique reference materials for measurement of actinide elements present as trace constituents. Each reference material unit is approximately 200 mg of PuO 2 powder in a quartz glass bottle. Characterized attributes of the oxides included mass fractions of plutonium, americium, neptunium, and uranium. Isotope-amount ratios were also determined for plutonium and uranium, but neptunium and americium were observed to be monoisotopic 237 Np and 241 Am. Measurements for characterization and verification of the attributes show that plutonium and trace actinides are homogeneous with the exception of limited heterogeneity for uranium, primarily observed for the 238 U isotope. Model purification ages calculated from measured americium and uranium attribute values are consistent with material histories and indicate that these impurities are predominantly due to the decay of plutonium isotopes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Complete Survey of Fuel Candidates for Microreactor Purposes

This report summarizes various fuel types that may be applicable to several microreactor concepts, which are defined as (1) very high temperature reactors (VHTR), (2) sodium fast reactors (SFR), (3) system for nuclear auxiliary power (SNAP) reactors, (4) gas fast reactors (GFR), and (5) molten salt reactors (MSR). The fuel systems that were assessed include: uranium mononitride (UN), uranium monocarbide (UC), uranium dioxide (UO 2 ), uranium oxycarbide (UCO) Tristructural Isotropic (TRISO), UN TRISO, mixed oxide (MOX), metallic fuels, and metal hydrides. While UCO TRISO has undergone significant testing through the advanced gas reactor (AGR) program, the very high cost necessitates consideration of other fuel types. UN and UC were identified as fuels that should receive further investigation due to their thermophysical and mechanical properties. Minimal irradiation performance data shows that these fuels are good candidate fuels for microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Oxidation of biogenic U(IV) mediated by iron-bearing clay minerals, iron-reducing bacteria, and organic ligands

Bioreduction of hexavalent uranium (U(VI)) to tetravalent uranium (U(IV)) by dissimilatory metal-reducing bacteria (DMRB) is considered an effective strategy for uranium immobilization in contaminated environments. However, U(IV) can be reoxidized to U(VI) under fluctuating redox conditions and remobilized. This work investigates the oxidation behavior of biogenic U(IV) in the presence of bioreduced iron-bearing clay minerals (rNAu-2), iron-reducing bacteria (Shewanella putrefaciens CN32), and organic ligands (ethylenediaminetetraacetic acid (EDTA) and citrate). Results demonstrate that the presence of CN32 significantly inhibits U(IV) oxidation. rNAu-2 exerted a context-dependent influence on U(IV) oxidation: its effect was masked by bicarbonate-promoted U(VI) mobilization in the absence of active CN32, but became detectable when CN32-mediated microbial protection slowed U(IV) oxidation. EDTA and citrate markedly accelerate U(IV) oxidation via formation of soluble U(IV)-ligand complexes, changing U(IV) redox potentials, and by promoting clay mineral dissolution that enhances Fe(II)/Fe(III) redox cycling. Collectively, our findings constrain the roles that clay minerals, iron-reducing bacteria, and organic ligands play in governing U(IV) stability, emphasizing the need to account for these factors in developing robust bioremediation strategies.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Oxidation kinetics of SPS-densified U 3 Si 2 fuels—Microstructure impact

Here, U 3 Si 2 is a potential candidate for accident tolerant fuels because of its high uranium density and excellent thermal conductivity in comparison to UO 2 . However, U 3 Si 2 suffers from oxidation, steam corrosion, and subsequent disintegration/pulverization. The detailed investigation of kinetics that incorporates fundamental treatment of oxidation of U 3 Si 2 is scarcely reported, and the oxidation mechanisms have not been fully elucidated. In this paper, the oxidation behavior of microcrystalline (mc - ) and nanocrystalline (nc - ) U 3 Si 2 have been systematically investigated using a thermogravimetric analysis (TGA) apparatus through a series of isothermal and non-isothermal kinetic studies. The isothermal kinetic study with a model-fitting approach indicates oxidation activation energy of 85 kJ/mol for dense mc - U 3 Si 2 and 96.4 kJ/mol for nc - U 3 Si 2 pellets, while the isoconversional approach leads to an activation energy in the range of 70–85 kJ/mol for mc - U 3 Si 2 and 75–86 kJ/mol for nc - U 3 Si 2 with three most common model-free methods, including Kissinger–Akahira–Sunose, Flynn–Wall–Ozawa, and Friedman methods. The derivation of oxidation activation energies using both isothermal and isoconversional methods highlights the approach to evaluate the oxidation resistance of nuclear materials using TGA quantitatively and makes it possible to compare among various nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The Stability of SiO{sub 2} Fe{sub 3}O{sub 4} Nanoparticles for Uranium Extraction in Acidic Media - 20189

This comprehensive study has shown that the proposed strategy of SiO{sub 2} coating of Fe{sub 3}O{sub 4} nanoparticles (NPs) for the use of uranium (U) extraction in acidic media was successful. The breakdown of Fe{sub 3}O{sub 4} in acidic media has proven to be a challenge when applying to a nuclear waste removal scenario, leading to the dissolution of Fe{sub 3}O{sub 4} and preventing uptake of U ions in the waste. To overcome this a SiO{sub 2} passivating layer was grafted to the Fe{sub 3}O{sub 4} NPs, to which an acid stability test was conducted between pH 3 and pH 4. It has been shown via Transmission Electron Microscopy (TEM) and zeta potential measurements that the uncoated Fe{sub 3}O{sub 4} NPs exhibit rapid coagulation when exposed to acidic conditions, moving from a stable zeta potential of -44 mV to an unstable zeta potential of +6 mV. SiO{sub 2} coated Fe{sub 3}O{sub 4} on the other hand can remain stable for several hours and continue to be stable beyond that. Throughout acid exposure the zeta potential of these NPs remained at ∼ -38 mV, showing great stability. As a functionality is needed on the surface of the NPs for separation and extraction, the phosphate complex used was also tested for its acid stability. It was determined that with a constant zeta potential of ∼ -20 mV, the phosphate functionality was able to remain stable in the acidic conditions simulating the environment of certain nuclear wastes. Overall SiO{sub 2} coated (PO)x-Fe{sub 3}O{sub 4} NPs show great promise for application in separation and removal of U in acid media. (authors)

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

Experimental system for studying temperature gradient-driven fracture of oxide nuclear fuel out of reactor

Temperature gradients in ceramic light water reactor (LWR) uranium dioxide (UO 2 ) nuclear fuel pellets generate thermal stresses that cause fractures in the fuel beginning early in the life of fresh fuel. The combination of heating due to fission and forced convective cooling on the exterior of LWR fuel rods generates a temperature profile that is difficult to replicate outside the reactor environment. In the present study, a state-of-the-art experimental set-up using electrical heating to replicate fission heating was built and surrogate fuel materials such as ceria (CeO 2 ) were used to validate the system. Cracking experiments were conducted on these surrogates by inducing reactivity-initiated-accident (RIA) like temperature gradients in the pellets via induction and direct resistance heating. Induction heating was done using copper coils and molybdenum susceptors which heated the surrogates to a threshold temperature that is sufficiently high for the fuel material to conduct current. Thereafter, direct resistance heating was used by a D.C. power supply to introduce volumetric heating to replicate LWR operating conditions analogous to fission heating. The pellets were held against nickel electrodes and mounted on a boron nitride test-stand. All the tests were carried out in a stainless-steel vacuum chamber. Simultaneous real-time dual imaging of the surrogate pellet surface has been implemented using an optical and infrared camera system which will be mounted along axial and perpendicular directions to the pellet surface respectively. A beam-splitter was used to split the incoming radiation from the sample into two halves. While one of the beams is transmitted from the splitter through a bandpass filter to obtain optical images, the other beam is reflected from the splitter to the thermal camera to capture full field temperature gradients of the as fabricated pellet surface during crack initiation and propagation. In the current series of tests, a 2-color pyrometer was used for recording and comparing the surface and centerline temperatures of the surrogate pellets in lieu of the thermal camera. A LabVIEW data acquisition system has been set up for collecting useful data during experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Actinide oxide dissolution in tributyl phosphate

An alternative to dissolving used nuclear fuel (UNF) in an acidic solution during reprocessing is direct dissolution in an organic solution, which would eliminate an aqueous dissolution step, decrease the amount of nitrate needed, and reduce the facility size. The flowsheet for this potentially less expensive alternative is first to voloxidize the UNF to remove fission product gases and form an oxide. After voloxidation, the UNF is then dissolved in an organic solution containing an extractant mixed with an aliphatic diluent and pre-equilibrated with nitric acid. The organic solution then goes through a solvent extraction process to recover the uranium and/or other desired radionuclides. This work qualitatively studied the dissolution of actinide oxides (UO2, NpO2, and PuO2) in tributyl phosphate using UV-Vis-NIR absorbance spectroscopy to ascertain dissolution behavior. Initial studies included material that is otherwise difficult to dissolve in only nitric acid, specifically CeO2, that is sometimes used as a dissolution surrogate for PuO2. This work confirmed that CeO2, NpO2, and PuO2 are difficult to dissolve in 30 vol% TBP-dodecane pre-equilibrated with 10 M HNO3 and will readily dissolve when co-precipitated with U (i.e., the mixed oxides U-Ce, U-Np, and U-Pu), surrogates for voloxidized nuclear fuel.

Gogolski, Jarrod [Savannah River National Laborato↗
DOE OSTI2024

Enhanced Mechanical Properties and Wear Resistance of FeCrAl Alloys at ~ 300°C and Higher Temperatures

Abstract Worldwide, light water reactors (LWRs) have been using zirconium (Zr)-based alloys for the cladding of the uranium dioxide fuel for more than 6 decades. Zr alloys oxidize rapidly in the presence of water and steam at temperatures > 450°C; therefore, they do not respond well to scenarios of loss of coolant accidents. There is a global effort by nuclear materials technologists to find more robust or stronger cladding materials for LWRs. One option is to use an iron-chromium-aluminum (FeCrAl) alloy since they have excellent resistance to high temperature oxidation and superior mechanical properties at LWR operation temperatures. Results show that (1) FeCrAl alloys have better mechanical properties than Zr alloy and are orders of magnitude more resistant to creep at temperatures higher than LWR normal operation conditions. (2) FeCrAl alloys have better resistance to fretting wear than Zr alloys at the normal operation conditions of LWRs.

Materials Science↗
DOE OSTI2023

Mixed nitrate and metal contamination influences operational speciation of toxic and essential elements

Environmental contamination constrains microbial communities impacting diversity and total metabolic activity. The former S-3 Ponds contamination site at Oak Ridge Reservation (ORR), TN, has elevated concentrations of nitric acid and multiple metals from decades of processing nuclear material. To determine the nature of the metal contamination in the sediment, a three-step sequential chemical extraction (BCR) was performed on sediment segments from a core located upgradient (EB271, non-contaminated) and one downgradient (EB106, contaminated) of the S-3 Ponds. The resulting exchangeable, reducing, and oxidizing fractions were analyzed for 18 different elements. Comparison of the two cores revealed changes in operational speciation for several elements caused by the contamination. Those present from the S-3 Ponds, including Al, U, Co, Cu, Ni, and Cd, were not only elevated in concentration in the EB106 core but were also operationally more available with increased mobility in the acidic environment. Other elements, including Mg, Ca, P, V, As, and Mo, were less operationally available in EB106 having decreased concentrations in the exchangeable fraction. Here, the bioavailability of essential macro nutrients Mg, Ca, and P from the two types of sediment was determined using three metal-tolerant bacteria previously isolated from ORR. Mg and Ca were available from both sediments for all three strains; however, P was not bioavailable from either sediment for any strain. The decreased operational speciation of P in contaminated ORR sediment may increase the dependence of the microbial community on other pools of P or select for microorganisms with increased P scavenging capabilities. Hence, the microbial community at the former S-3 Ponds contamination site may be constrained not only by increased toxic metal concentrations but also by the availability of essential elements, including P.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

SCALE 6.2.4 Validation: Nuclear Criticality Safety

The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). More than 600 cases from this handbook have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID), which is maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the SCALE 6.2.4 code system is assessed using the VALID models of benchmark experiments. A range of nuclear cross section libraries based on Evaluated Nuclear Data File (ENDF)/B-VII.1 in both multigroup (MG) and continuous energy (CE) formats is considered. The critical experiments available to validate the KENO V.a code cover 15 broad categories of systems. These systems use a range of fissile materials, including a range of uranium enrichments, various plutonium isotopic vectors, and some mixed uranium/plutonium oxides. The physical forms of the fissile material also vary and are represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover fast, intermediate, mixed, and thermal spectra. Over 550 of the total cases use the KENO V.a code for the four nuclear data libraries considered in this report.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗