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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Molecular Models of Atomically Dispersed Uranium at MoS 2 Surfaces Reveal Cooperative Mechanism of Water Reduction

Single atoms of uranium supported on molybdenum sulfide surfaces (U@MoS 2 ) have been recently demonstrated to facilitate the hydrogen evolution reaction (HER) through electrocatalysis. Theoretical calculations have predicted uranium hydroxide moieties bound to edge-sulfur atoms of MoS 2 as a proposed transition state involved in the HER process. However, the isolation of relevant intermediates involved in this process remains a challenge, rendering mechanistic hypotheses unverified. The present work describes the isolation and characterization of a uranium-hydroxide intermediate on molybdenum sulfide surfaces using [(Cp* 3 Mo 3 S 4 )UCp*], a molecular model of a reduced uranium center supported at MoS 2 . Mechanistic investigations highlight the metalloligand cooperativity with uranium involved in the water activation pathway. The corresponding uranium-oxo analogue, [(Cp* 3 Mo 3 S 4 )Cp*U(═O)], was also accessed from the hydroxide cluster via hydrogen atom transfer and from [(Cp* 3 Mo 3 S 4 )UCp*] through an alternative direct oxygen atom transfer. These results provide an atomistic perspective on the reactivity of low-valent uranium at molybdenum sulfide surfaces toward water, modeling key intermediates associated with the HER of U@MoS 2 catalysts.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

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

Ligand effects on electronic structure and bonding in U(iii) coordination complexes: a combined MCD, EPR and computational study

The trivalent oxidation state of uranium has been shown to undergo unique reactivity, from its ability to activate a variety of small molecules to its role in the catalytic reduction of ethene to ethane amongst others. Central to this unique reactivity and ability to rationally design ligands for isotope separation is the underlying uranium electronic structure. While electronic structure studies of U(IV), U(V), and U(VI) have been extensive, by comparison, analogous studies of more reduced oxidation states such as U(III) remains underdeveloped. Herein we report a combined MCD and EPR spectroscopic approach along with density functional theory and multireference wavefunction calculations to elucidate the effects of ligand perturbation in three uranium(III) Tp* complexes. Overall, the experimental and computational insight suggests that the change in ligand environment across this series of U(III) complexes resulted in only minor perturbations in the uranium electronic structure. This combined approach was also used to redefine the electronic ground state of a U(III) complex with a redox non-innocent Bipy - ligand. Overall, these studies demonstrate the efficacy of the combined experimental and theoretical approach towards evaluating electronic structure and bonding in U(III) complexes and provide important insight into the challenges in altering ligand environments to modify bonding and reactivity in uranium coordination chemistry.

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

Minimizing thickness variation in monolithic U-10Mo fuel foil and Zr interlayer during hot rolling: A microstructure-based finite element method analysis

Low-enriched uranium alloyed with 10 wt. % molybdenum (U-10Mo) has been identified as a promising alternative to highly enriched uranium fuel for the United States’ high performance research reactors. The monolithic U-10Mo fuel plate consists of a metallic U-10Mo fuel foil with a 25 µm Zr interlayer and a relatively thick cladding of aluminum alloy 6061. The Zr interlayer is typically applied during the hot co-rolling process, and this process dictates the uniformity of the Zr interlayer. Thickness variation observed in the U-10Mo and Zr interlayer has been attributed to several sources: the initial grain size of the U-10Mo castings, can materials, rolling temperature, inhomogeneous molybdenum content, and porosity in the cast U-10Mo. This thickness variation limits the ability to meet the dimensional specification; thus, a better understanding of the factors causing the nonuniform thickness is needed. In this work, we used a novel, microstructure-based finite element method to model the hot rolling process to address these concerns. Grain microstructures in U-10Mo were tessellated and explicitly considered in the finite element model. Each grain was assigned a random material property to mimic the grain strength variations induced by different grain orientations. Simulations were performed using six steel can thicknesses, four grain sizes, and with or without a Zr interlayer to investigate the influences of those variables on the thickness nonuniformity. The simulation results showed that a thinner steel can and finer U-10Mo grain size reduce thickness variations in both the U-10Mo fuel foil and Zr interlayer. The direct findings from the simulations and analysis can be used to optimize the hot rolling schedule, reduce fabrication defects, and meet the dimensional specifications. The proposed microstructure-based finite element model can be also coupled with experimental microstructure characterization data, images, and models to simulate multi-pass hot rolling.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Assessment of Uranium Isotope Impacts

This report examines the benefits and challenges associated with recycling recovered uranium (RU) from the used nuclear fuel (UNF) discharged from a once-through sodium-cooled fast reactor (SFR) with high-assay low-enriched uranium (HALEU) which has a higher residual U-235 content in the RU. However, there are three minor uranium isotopes that influence the value of the RU. There will be trace quantities of U-232 in RU, which can increase dose rates and, in turn, increase the cost of fuel fabrication. Additionally, the U-234 and U-236 isotopes are found in higher concentration in RU than in NU, and act as neutron poisons, requiring higher U-235 concentrations to compensate for the neutron poisoning effects of these isotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Insight into the structural ambiguity of actinide(IV) oxalate sheet structures: a case for alternate coordination geometries

Plutonium(IV) oxalate hexahydrate (Pu(C 2 O 4 ) 2 ∙ 6H 2 O; PuOx) is an important intermediate in the recovery of plutonium from used nuclear fuel. Its formation via precipitation is well studied, yet its crystal structure remains unknown. Instead, the crystal structure of PuOx is assumed to be isostructural with neptunium(IV) oxalate hexahydrate (Np(C 2 O 4 )­ 2 ∙ 6H 2 O; NpOx) and uranium(IV) oxalate hexahydrate (U(C 2 O 4 )­ 2 ∙ 6H 2 O; UOx) despite the high degree of unresolved disorder that exists when determining water positions in the crystal structures of the latter two compounds. Such assumptions regarding the isostructural behavior of the actinide elements have been used to predict the structure of PuOx for use in a wide range of studies. Herein, we report the first crystal structures for PuOx and Th(C 2 O 4 ) 2 · 6H 2 O (ThOx). This data, along with new characterization of UOx and NpOx, has resulted in the full determination of the structures and resolution of the disorder around the water molecules. Specifically, we identify the coordination of two water molecules with each metal center, which necessitates a change in oxalate coordination mode from axial to equatorial that has not been reported in the literature. Here this work exemplifies the need to revisit previous assumptions regarding fundamental actinide chemistry, which are heavily relied upon within the current nuclear field.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Nuclear fuels for transient test reactors

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm 2 ) and are amenable to longer “shaped” transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2 ) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (~2E15 n/cm 2 ). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (~1E15 n/cm 2 ). An evolution of the UO 2 system, termed “ternary ceramic” fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (~7E15 n/cm 2 ), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (~6E15 n/cm 2 ). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. In conclusion, like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Investigating trends in actinide covalency and magnetism with 35/37 Cl SSNMR spectroscopy and first-principles calculations

Solid-state NMR (SSNMR) spectroscopy is a powerful technique for studying actinide chemistry but has been significantly limited due to the complex paramagnetism, and radiological hazards presented by these materials. Lanthanide and actinide salts often feature magnetic ordering and can be paramagnetic, ferromagnetic, or antiferromagnetic depending on temperature and electronic structure. Paramagnetic interactions can manifest in SSNMR both as secular spectral shifts and/or couplings as well as contributions from non-secular relaxation. Both effects can be directly measured with NMR and used to extrapolate rich chemical information such as coordination environments, bonding characteristics, local molecular dynamics, and correlation times. Typically, these studies are carried out on high-γ and highly abundant NMR-active isotopes (e.g., 1 H, 6/7 Li, 19 F, 23 Na, etc.) or on enriched rare isotopes (e.g., 2 H and 17 O), which can be expensive. Herein, we present a facile methodology to measure the 35/37 Cl electric-field gradient (EFG) and paramagnetic shift anisotropy (SA) tensor components using static wideline SSNMR measurements of LaCl 3 , NdCl 3 , UCl 3 , and UCl 4 . The static powder spectra were measured with both 35 Cl and 37 Cl SSNMR to increase the fidelity of the extracted tensor parameters. Variable temperature NMR of a select case confirms the Curie–Weiss paramagnetism. Relaxation measurements of both nuclei further corroborate observations owing to the paramagnetic relaxation enhancement and reveal simultaneous quadrupolar relaxation mechanisms. Density functional theory (DFT) calculations using Hubbard U corrections to the uranium valence orbitals show excellent agreement with experimental EFG tensor parameters and help describe the bonding characteristics in these lanthanide and actinide systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Comparison of excitation temperature of a laser-produced plasma by combining emission and absorption spectroscopy

Measurement of the temporal evolution of laser-produced plasma temperature is very important for many of its applications, and several plasma diagnostic tools are routinely used by researchers. However, it is very challenging to measure the properties of the plasma at the early and late times of its evolution using a single diagnostic tool. In this study, we combined emission and laser absorption spectroscopy to compare the excitation temperatures of a laser-produced uranium plasma system. Several U I transitions in the near-infrared spectral range (775–800 nm) were considered, and the Boltzmann plot method was used to measure the excitation temperatures using both emission and absorption spectroscopy. Emission spectroscopy provided early-time temperature measurements of the plasma up to times 2–20 µs, while absorption spectroscopy provided temperature measurements at late times of plasma evolution (for times 5–80 µs). The emission and absorbance of U I transitions were found to follow the Boltzmann distribution, indicating the plasma is likely in the state of local thermodynamic equilibrium even at late times of its lifetime. The emission and absorption-based time-resolved excitation temperatures demonstrated good agreement at earlier times (≤15 µs) in the overlapped temporal region, while a deviation in the measured values was seen at times (≥15 µs), and potential reasons for such a disagreement are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Uranium Nitride as LWR TRISO Fuel: Thermodynamic Modeling of U-C-N and Thermomechanics

A new concept for light water reactor (LWR) nuclear fuel has recently been proposed that utilizes pellets of tristructural isotropic (TRISO) fuel particles embedded in a SiC or zirconium alloy matrix, replacing sintered urania pellets. The fuel is considered significantly more accident tolerant than urania as the SiC matrix or coated/protected zirconium alloy and TRISO particles would be highly resistant to oxidation and fission product release under beyond-design-basis accident conditions. As is well understood, failed fuel rods release gaseous and volatile fission products, and that is greatly exacerbated as urania oxidizes from UO 2+x to higher oxidations states in failed fuel pin with resulting expansion, loss of integrity, and significant further release of fission products.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

ZPPR-15 Small Reactivity Worth Experiments

In addition to measuring the worth of sodium voiding and the worth of simulated control rods, the ZPPR staff measured the Doppler worth of four standard samples, the worth of axial expansion and the worth of radial bowing in ZPPR-15. Doppler sample worths were measured in ZPPR-15A, ZPPR-15B and ZPPR-15D. The Doppler samples were natural UO 2 , depleted uranium metal, depleted U-10Zr alloy and 33% enriched UO 2 . These samples were placed in a sample capsule in the inner core of ZPPR-15 and heated to various temperatures between 300 K and 1100 K. The resulting changes in reactivity relative to the reference configuration were measured to determine the Doppler worths of the samples. Special segmented drawers were constructed to measure the worth of axial expansion in ZPPR15. The segments were connected to each other, and the back segment was connected to a spring-loaded cable. Pulling the cable introduced small gaps between drawer segments to simulate axial expansion, and releasing the cable eliminated those gaps. The worth of this simulated axial expansion was measured in ZPPR-15A, ZPPR-15B and ZPPR-15D. The worth of radial bowing was measured in ZPPR-15, but the experimenters regarded the result as unsatisfactory. Radial bowing was measured by two methods in ZPPR-17A, so the ZPPR-17A bowing measurements were analyzed in place of the unpublished ZPPR-15 radial bowing measurement. The first ZPPR-17A measurement consisted of rearranging the plates in 96 core drawers at the core-radial blanket boundary to simulate outward motion of the fuel during bowing. The second ZPPR-17A measurement used a new bowing oscillator to move the core plates a small distance vertically in a special drawer. The published uncertainties for the Doppler worth, axial expansion worth and radial bowing worth measurements are the statistical uncertainties in the measurements. In reality, the published uncertainty for each of these measurements is just one component of the total uncertainty. There are additional uncertainties specific to each measurement and a common uncertainty related to conversion from the natural measurement units, cents, to pcm for calculations. A full uncertainty analysis was performed for each measurement. The significant uncertainties were quantified, and a total uncertainty was determined for each measurement. The ZPPR-15 and ZPPR-17A experimental records were used to create detailed as-built Monte Carlo models of the Doppler worth, axial expansion and radial bowing measurement configurations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Oak Ridge National Laboratory Fission Product Hydrochlorination

This work supports the recovery of high-assay low-enriched uranium (HALEU, uranium enriched in 235 U to >5% and <20%) from used nuclear fuel through a process based on the zirconium extraction (ZIRCEX) process. The ZIRCEX process starts with the dry head-end removal of zirconium or aluminum cladding from the nuclear fuel via chlorination or hydrochlorination, which results in the separation of cladding from the fuel as the metal chloride (e.g., aluminum trichloride or zirconium tetrachloride). The hydrochlorination process is included in the Material Recovery Pilot Plant (MRPP), a ZIRCEX-based demonstration plant at Idaho National Laboratory (INL). The experiments outlined in this report assessed fission product (FP) volatility and transport under the hydrochlorination conditions used to separate cladding from used nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Comparative Structural Rigidity Analysis of the NBSR DDE and the NBSR LEU Fuel Element

The National Bureau of Standards Reactor (NBSR) is a 20-MW heavy-water moderated and cooled, enriched-fuel, tank-type reactor located in the NIST Center for Neutron Research laboratory complex. NBSR is one of six U.S. high performance research reactors (USHPRR), including one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to convert from the use of highly enriched uranium fuel to the use of low-enriched uranium (LEU, < 20 wt% U-235) fuel. On the LEU U-10Mo fuel qualification side, mini-plate and large-plate irradiations have been successfully performed, and more experiments are either ongoing or planned. As an additional experimental campaign to support licensing of the LEU fuel for the use in the NBSR, the NBSR Design Demonstration Element (DDE) will be a full-sized test element similar to the NBSR LEU fuel element and irradiated in a test reactor under conditions as prototypic of the NBSR LEU fuel element operating conditions as possible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Analysis of H-Canyon Process Tanks in Preparation of Consolidation and Blending for HALEU Fuels

High-Assay Low- Enriched Uranium (HALEU) fuels are being developed to support the replacement of Highly Enriched Uranium (HEU) fuels used in U.S. High-Performance Research Reactors (USHPRR) as well as advanced nuclear power reactor designs. The projected demand for HALEU far exceeds the supply and studies are underway to assess various options to partially mitigate the potential short supply. The H Canyon facility at the Savannah River Site (SRS) Low Enriched Uranium (LEU) containing 4.95% U-235 from the reprocessing of highly enriched foreign and domestic research reactor fuel for the Tennessee Valley Authority’s (TVA) commercial power reactor market for several decades. The production of LEU at the H-Canyon facility can be readily transitioned to produce 19.75% HALEU solutions from the current separated inventory of purified HEU solutions in H-Canyon storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

High Flux Isotope Reactor Uncertainty Factors

This report provides a brief summary of the uncertainty factors used for High Flux Isotope Reactor (HFIR) steady-state heat transfer analyses. These factors are mainly used in the HFIR Steady-State Heat Transfer Code (HSSHTC) to perform core thermal margin evaluations that determine safe reactor operation. The attempt to classify these factors stems from the assumption that the current approach is characterized by an excess of conservatism, thereby restricting reactor performance. The work documented herein was of a limited scope and pertained mainly to factors’ description and initial grouping based on their functional use. Suggestions are provided for further evaluation for the low-enriched uranium (LEU) to the uranium silicide dispersion fuel (U 3 Si 2 -Al).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

SULI Intern Final Report: Computationally Investigating Hydrogen Thermo-Diffusion in Yttrium Hydride Using Multiscale Methods

The renaissance of nuclear energy has arrived, heralding an age of abundant inexpensive clean energy, and renewed space exploration. In nuclear-powered spacecraft and microreactors, safety and size are of utmost importance. Yttrium Hydride (YHx) is being researched for its utility as a neutron moderator in nuclear reactors; the hydrogen in YHx slows down neutrons, enabling a continuous nuclear reaction in the reactor. This has the benefit of allowing reactors to be more safe, compact, and efficient. The goal of this effort is to computationally predict the coefficient of temperature-dependent hydrogen diffusion within YHx, the Soret coefficient. This parameter is essential for determining the safe operating modes of YHx moderators. Zirconium Hydride (ZrHx) is used in the Training, Research, Isotopes, General Atomics (TRIGA) reactor, is the reference material for these calculations. In this work, nanoscale atomic modeling in the Vienna Ab initio Simulation Package (VASP) is combined with the mesoscale finite element phase-field module in the Multiphysics Object-Oriented Simulation Environment (MOOSE); this culminates in a new multiscale computational method to simulate Soret diffusion of hydrogen in YHx. This data is useful for predicting experimental outcomes. This workflow involves convergence testing followed by static, Nudged Elastic Band (NEB), Quasi-Harmonic Approximation (QHA), and Molecular Dynamics (MD) calculations - linked with phase field simulation. NEB simulates hydrogen migration, while QHA and MD predict temperature-dependent properties. The static calculations align with literature, and preliminary NEB and QHA calculations yield accurate results. Once the atomic calculations are complete, we will incorporate Electron Backscatter Diffraction (EBSD) images and VASP-generated parameters into the phase field module to simulate intra- and intergranular transport of hydrogen in ZrHx and YHx. Future research will extend our approach to fuel-moderator materials systems such as Uranium-Yttrium Hydride (U-YHx). This work contributes to the development of advanced nuclear energy solutions for space travel.

36 - MATERIALS SCIENCE↗