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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Godiva IV Simulated Radiation Field Characterization and Variance Reduction

Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Early Stages of Age Hardening in U-6%Nb: An Energy Barrier Hypothesis for Twin Reorientation

It is well known that age hardening occurs in U-6%Nb (Ref. 1-9). Figure 1 shows that increases in hardness occur at aging temperatures up to ~400°C, and overaging occurs above ~450°C. Overaging occurs by cellular decomposition of the αʺ b martensite, similar to overaging reactions in other uranium alloys. But the mechanism of age hardening remains elusive, particularly in its early stages.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

The Electron Thermal Conductivity of Pu and Zr Substituted Gamma-Uranium

Uranium alloys are attractive recycled nuclear fuels because of their high thermal conductivity (k) and fissile density; however, the effects of alloying elements on k remain unclear. Here, the electron thermal conductivity (k_e) of U-Pu-Zr compositions are calculated using density functional theory. The electronic structure is evaluated to understand the effects of plutonium (Pu) and zirconium (Zr) substitution on the k_e of ?-U. Alloys of up to 37.5 at. % Pu and 37.5 at. % Zr are examined. Two methods are applied to calculate k_e; we find that the accuracy of each method depends on the electronic and mass similarities between the solute and solvent atoms. Specifically, when the solute atom is similar in electronic structure and mass, the method that applies the electron relaxation time of ?-U is best, while if the elements are dissimilar, a mixed method that mixes several parameters associated with k_e from each element in the alloy is best. The introduction of all alloying elements decreases k_e; however, in binary compounds, Pu and Zr have different effects. Pu generally flattens the electronic bands but compensates for this deleterious effect by increasing electron density near the Fermi level. Zr flattens the electronic bands more severely without adding electron density near the Fermi level. Therefore, Zr decreases the k_e more than Pu in binary compounds. In ternary compounds, the difference between Pu and Zr is minimal due to the phononic change from the large mass change of Zr substitution, even at 12.5 at. %. Thus, we predict that higher loadings of Pu, and potentially other actinides, can be added to U-Pu-Zr compositions for faster recycling of spent fuel with without sacrificing k. We also note that these k_e calculation methods can be applied to non-fuel alloys that require k_e predictions, such as cladding, heat exchanger, and structural materials.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Molecular dynamics studies of grain boundary mobility and anisotropy in BCC γ-uranium

Grain morphologies such as grain size and aspect ratio in uranium-based metallic fuels are important microstructural features that can impact various fuel performance properties such as fission-gas-induced swelling, thermal transport, high burnup structure formation, and radiation resistance. Accurate prediction of the fuel grain morphologies requires knowledge of critical grain growth parameters such as grain boundary (GB) mobility and anisotropy. In this work, molecular dynamics simulations were performed to study the GB mobility and its anisotropy in pure body-centered-cubic (BCC) γ uranium. Nine GBs with different combinations of misorientation angles (20°, 30°, 45°) and rotation axes (<100>, <110>, <111>), as well as an additional <111> 38.2° GB were studied using three interatomic potentials. Here, it is found that the GB mobility anisotropy has complex trends, depending on both rotation axis and misorientation. However, in general the <110> rotation axis has the fastest GB mobility at the same misorientation. The results of this work can be used as not only baselines for future studies of GB mobility in uranium-based alloys such as uranium-molybdenum (U-Mo) fuels, but also input for mesoscale modeling of grain growth in uranium-based alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Shield materials recommended for space power nuclear reactors

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

Kaszubinski, L. J.↗
DOE OSTI2021

Mechanical and Thermal Properties for Uranium and U–6Nb Alloy from First-Principles Theory

Elasticity, lattice dynamics, and thermal expansion for uranium and U–6Nb alloy (elastic moduli) are calculated from density functional theory that is extended to include orbital polarization (DFT+OP). Introducing 12.5 at.% of niobium, substitutionally, in uranium softens all the cii elastic moduli, resulting in a significantly softer shear modulus (G). Combined with a nearly invariant bulk modulus (B), the quotient B/G increases dramatically for U–6Nb, suggesting a more ductile material. Lattice dynamics from a harmonic model coupled with a DFT+OP electronic structure is applied for α uranium, and the obtained phonon density of states compares well with inelastic neutron-scattering measurements. The Debye temperature associated with the lattice dynamics falls within the range of experimentally observed Debye temperatures and it also validates our quasi-harmonic (QH) phonon model. The QH Debye–Grüneisen phonon method is combined with a DFT+OP electronic structure and used to explore the anisotropic thermal expansion in α uranium. The anomalous negative thermal expansion (contraction) of the b lattice parameter of the α-phase orthorhombic cell is relatively well reproduced from a free-energy model consisting of QH-phonon and DFT+OP electronic structure contributions.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2024

A Critical Review on the History of Fabricating Monolithic U-Mo Fuel Plates

Fabrication of uranium-molybdenum alloy fuels has been occurring since the early 2000s in support of the development of a high-uranium-density low-enrichment fuel for use in high-performance research and test reactors which operate at relatively low temperatures. The primary fuel form—a thin foil of uranium, alloyed with 10wt% molybdenum, which is coated in a layer of zirconium and then encapsulated with aluminum 6061 as a cladding material—was developed over a number of years.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A critical review of the history of fabricating monolithic U-Mo fuel plates

The fabrication of monolithic uranium-molybdenum alloy fuels, specifically those developed for high-performance research and test reactors, began in the early 2000s. The primary fuel form consists of uranium alloyed with 10-wt% molybdenum in a thin foil coated with zirconium and encapsulated in aluminum-6061 cladding. Over the years, the process has evolved with different types of casting, heat treatments, rolling schedules, and cladding applications. This review examines the history of these fabrication processes and their impact on microstructure and fuel-swelling performance. Even though various fabrication methods were used, we found little correlation between fabrication variation and fuel swelling. This insensitive relationship between fabrication variation and fuel swelling is primarily due to inhomogeneous microstructures that formed during casting and grain refinement that occurred during rolling. We conclude that the fabrication processes we examined produced similar microstructures, indicating that the fuel microstructure is somewhat insensitive to the fabrication parameters evaluated. However, the relatively small amount of historic data, such as those for grain sizes, limited this analysis. More recently fabricated materials, such as those from ongoing irradiation experiment, Mini-Plate-1 and Mini-Plate-2, were also excluded from this analysis and are intended to be reviewed separately. The findings, that fuel microstructure is somewhat insensitive to the fabrication parameters, do not imply that any fabrication method is acceptable, given the uncertainties in data and fuel-swelling observations. For example, only arc melting and vacuum induction melting casting processes were previously explored in the historic fabrication efforts. In conclusion, the findings should not be extrapolated to other casting processes.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Solid-state phase transitions of two quaternary metallic fuel alloys (U-2.5Mo-2.5Ti-5.0Zr and U-1.5Mo-1.5Ti-7.0Zr in wt. %)

This study focuses on the solid-state phase transitions of two quaternary fuel alloys for fast reactors: U-1.5Mo-1.5Ti-7.0Zr (U-MT7Z) and U-2.5Mo-2.5Ti-5.0Zr (U-MT5Z). Here, the phase transitions were determined by differential scanning calorimetry (DSC), X-ray powder diffraction (XRD), and scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS). To identify the high temperature phases, the alloys were annealed at 873K, 948K, 1023K, and 1123K (all under 72 hours). Combining those characterizations, the phase transitions are determined. In U-MT7Z, there is one phase transition observed, and it is ascribed to α + U 2 Ti → γ transition. In U-MT5Z, two transitions are found. The first transition is α → γ, while the second is U 2 Ti → γ. The γ phase onset temperature of U-MT5Z is lower than that of U-MT7Z due to the higher Mo content.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Method to produce salts containing actinide halides

A method of producing uranium halides is disclosed in which chlorine gas is introduced into a liquid uranium-nickel alloy. NaCl salt is surrounding the crucible containing the liquid uranium-nickel alloy, producing a eutectic mixture of NaCl—UCl3. Upon chlorination, the metal halide dissolves in the matrix salt forming a solution. Adding the reactant metal, uranium to the nickel, the alloy is able to remain molten throughout processing. The liquid metal alloy may be removed from the salt bath, while the halogen gas continues to enter the system through the sparge until the desired composition of NaCl—UCl3—UCl4 is achieved. The method and system can be used to produce other metal halide salts such as actinide, lanthanide or transition metal halides contained in a matrix salt consisting of alkali and/or alkaline earth halides.

Williamson, Mark A.↗
DOE OSTI2025

Technical Considerations on MURR Control Blade Design Change and Testing using a New Metal Matrix Composite

The University of Missouri Research Reactor (MURR) is one of six research reactors, including a critical facility, that are pursuing conversion as part of a collaboration with the U.S. Department of Energy National Nuclear Security Administration Material Management and Minimization Office of Reactor Conversion and Uranium Supply, under the U.S. High Performance Research Reactors (USHPRR) conversion project. Five of the six USHPRR are planned to convert from highly enriched uranium (HEU) fuel using a low-enriched uranium (LEU) high assay monolithic alloy of uranium-10 wt% molybdenum (U-10Mo). As part of the conversion safety analysis, it is necessary to demonstrate the safety performance of the proposed core fueled with LEU as compared to the current HEU cores. The MURR reactor is planning to switch to a new control blade design that uses a metal matrix composite of boron carbide (B 4 C) and aluminum as the absorber in place of Boral®. Since MURR is expected to adopt the new metal matrix composite control blade design prior to conversion, the impact of the new blade design on the neutronics characteristics of the MURR cores for conversion are analyzed in this work through updates to incorporate the changes to the blade design in conversion models as they directly impact the LEU conversion safety analysis. The quantitative comparison shows that the neutronics and thermal hydraulic behavior of one metal matrix composite blade replacing a Boral blade is comparable for the two example MURR LEU and HEU cores states considered. Geometrical changes in the metal matrix composite blade design, combined with a 4% increase in areal boron density, showed local heating effects up to 20% higher than the Boral design. As expected, the metal matrix composite showed slightly lower heat depositions and absorber region temperatures for the LEU cases compared to HEU. Although this analysis was comparative for a single blade, maximum control blade temperatures for both Boral, metal matrix composite, and HEU/LEU remained below 100 °C, though additional analysis at a core level could differ. A qualitative irradiation behavior assessment concludes that the mechanisms that may drive swelling and blistering in the current Boral design are eased by the adoption of the metal matrix composite design. The work concludes that the two blade designs are essentially equivalent with regards to neutronics, thermal hydraulics, and expected material behavior under irradiation. However, due to the geometrical changes to the blades including redesigned and thinner cladding, new testing and increased surveillance for distortion and swelling are recommended to confirm the performance of the metal matrix composite control blade design. Where testing is completed prior to conversion, the only anticipated impacts on conversion to LEU U-10Mo fuel would be the need for models and safety analysis incorporating the metal matrix composite control blades.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1975

Creep behavior of uranium carbide-based alloys

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

Seltzer, M. S.↗
DOE OSTI2023

Estimate of Gamma Dose Rates from Arrays of Fermi-1 Blanket Elements During the MEDE Process

The Enrico Fermi Atomic Power Plant Unit 1 (Fermi-1) was a sodium-cooled fast breeder reactor located in Monroe County, Michigan. The reactor was powered by a core of enriched uranium metal alloy driver fuel, which was enveloped by an axial and radial blanket material consisting of depleted uranium metal alloyed with 3 wt.% molybdenum. There are 406 axial and 559 radial irradiated sodium-bonded Fermi-1 blanket assemblies in storage at INL, totaling 34 metric tons of heavy metal. Disposal of the Fermi-1 blanket material directly into a geological repository is prohibited due to the reactive characteristic of its bond sodium. A melt drain evaporate (MEDE) process can effectively remove bond sodium from Fermi-1 blanket material. Consequently, planning is underway to apply a MEDE process to treat the 34 metric tons of heavy metal of irradiated Fermi-1 blanket material. Given the irradiated Fermi-1 blanket material’s relatively low power history and 50+ years of decay, it can be treated in a shielded glovebox. To assess the requisite shielding, the dose rates of the Fermi-1 blanket elements in various process configurations are needed. The method to perform the dose rate calculations is to first generate an average source term and associated photon source spectra for the Fermi-1 blanket material in SCALE and then use the associated photon spectra to calculate dose rates using MCNP6.2 in seven representative geometries. Dose rates in rem/h were calculated on contact (1 cm from outer geometry surface), 30 cm away from outer geometry surface, and 1 m from outer geometry surface at axial heights spanning the length of the blanket material within the geometry. The maximum average dose rate for the single Fermi-1 radial blanket element is ~0.60 rem/h, for the Fermi-1 radial blanket assembly ~1.10 rem/h, for the single Fermi-1 axial blanket element ~0.5 rem/h, for the Fermi-1 axial blanket assembly ~1.01 rem/h, for the MEDE can ~ 0.98 rem/h, for the Fermi Storage Canister ~ 0.83 rem/h, and for the MEDE cans in the DOE Standard Canister ~0.51 rem/h. Based on these maximum average values, the bounding dose rates are assumed to be 3× the average dose rates calculated for the axial blanket material and 7.6× the average dose rates calculated for the radial blanket material, based on the distribution of Fermi-1 blanket material assembly burnup.

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

Transition Core Analysis for HEU to LEU Fuel Conversion at the University of Missouri Research Reactor

A new type of low-enriched uranium (LEU) fuel based on an alloy of uranium and molybdenum is expected to allow the conversion of U.S. domestic high-performance research and test reactors requiring high density fuel from highly enriched uranium (HEU) to LEU. The University of Missouri Research Reactor (MURR (R)) has undergone design and performance calculations for conversion to this LEU fuel. Presented in this paper is the analysis of a crucial step in the conversion process: the sequence of MURR transition cores from all fresh to equilibrium burnup LEU operations. During the initial conversion from HEU to LEU fuel, MURR will operate atypically due to the lack of burned LEU elements. Given the constraints of MURR operation and experiments, a proposed transition scheme minimizes the time MURR operates atypically compared to the prototypic cycles currently run with HEU fuel and moves quickly to the same sort of equilibrium cycles for the LEU fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Capability to Process and Characterize Uranium-Zirconium (U-Zr) and Uranium-Zirconium-Plutonium (U-Zr-Pu) Alloys

This work investigated co-reduction of anhydrous compounds of uranium, zirconium, and plutonium to produce uranium rich ternary nuclear fuel alloys. Metallothermic co-reduction is a novel method to produce all-metal nuclear fuels. Metal fuels offer thermal, fissility, compatibility, and security benefits over oxide fuels. Alloys of uranium–10% zirconium with plutonium contents of 0%, 2.5%, 5%, and 10% were produced, with yields of 60–85% of theoretical values in a traditional calciothermic bomb reduction apparatus. Microstructural analysis indicated transformation of uranium phase from gamma to beta and then alpha, in alternating lamellar plates typical of alpha phase uranium and delta phase uranium-zirconium, with zirconium and carbides at prior grain boundaries. Some of the analyses were inconclusive in their results and therefore require additional testing. Successful ternary co-reduction would simplify production of homogeneous feedstock and thereby streamline manufacture of homogeneous ternary metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗