Uranium Silicide and Uranium Silicide-Nitride Fuels: Assessing Phase Behavior for Fabrication and In-Reactor Behavior
Explore the source record for details and available documents.
L. Wilson, Tashiema↗
SEARCH · Engineering Papers
Results for “URANIUM SILICIDES”
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.
A modified Embedded-Atom Method interatomic potential for uranium-silicide
Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). U-Si fuels benefit from higher thermal conductivity and higher fissile density compared to UO2. In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical Modified Embedded Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential accurately describes not only the primary phase of interest (U3Si2), but also a variety of U-Si phases across the composition spectrum.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nano-scale mechanical behavior of uranium silicide compounds
Uranium silicide compounds exhibit a range of attractive material properties for nuclear fuel applications, but little is known regarding their mechanical behavior. The structure of nuclear fuels at high burnup can prove to be especially difficult given the significant changes in both chemistry and microstructure caused by fission products. Nano-indentation appears to be a valuable addition to post irradiation examination. The present focus is assessment of the room temperature modulus and hardness of U{sub 3}Si{sub 2}, U{sub 3}Si{sub 5} and USi materials. Each of the compositions under investigation was synthesized via arc melting of the pure metals in an Ar atmosphere using a titanium getter to remove O{sub 2} contamination. The resulting pellets following milling and sintering of the arc melted material were between 96 and 98% of density. Coupling different techniques (SEM and Chemical Analysis), microstructural investigations of the samples has been performed. Preliminary nano-mechanical response of U-Si fuel materials has been demonstrated, on a Hysitron Tribo-indenter which will impact the ability to design advanced fuel materials for enhanced high temperature performance.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
First-principles study of surface properties of uranium silicides
We report that uranium silicides are currently under investigation as accident tolerant fuels for light water reactors because of its high uranium density and high thermal conductivity. Surface energy as an important material property is required for modeling of gas bubble behavior in nuclear fuels using mesoscale approaches, such as phase field and rate theory methods. However, there is no such information available for uranium silicides from either experiment or theory. To this end, we study the surface properties of two uranium silicide compounds U 3 Si 2 and U 3 Si using first-principles calculations. Of the low-index facets of tetragonal U 3 Si 2 and U 3 Si, we study a total of 13 surfaces up to a maximum Miller index of 3. From the calculated surface energies, the equilibrium single crystal shapes of U 3 Si 2 and U 3 Si are obtained using Wulff construction. The dominant surface orientation, surface area weighted surface energy and surface anisotropy are predicted. Finally, the obtained surface properties of U 3 Si 2 and U 3 Si can be used for an accurate description of the morphology of fission gas bubbles in uranium silicide fuels in the future.
36 MATERIALS SCIENCE↗
Development of molecular dynamics potential for uranium silicide fuels
Use of uranium–silicide (U-Si) in place of uranium dioxide (UO2) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. This is due to a higher thermal conductivity than UO2 that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. For example, silicide compounds (U3Si2 and U3Si) are known to become amorphous under irradiation. There was clear independent experimental evidence to support a crystalline to amorphous transformation in those compounds. However, it is still not well understood how the amorphous transformation will affect on fuel behavior. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In order to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters and increase the predictive capabilities of the developed model, inputs from atomistic simulations are essential. First-principles based density functional theory (DFT) calculations will provide the most reliable information. However, it is probably not possible to obtain kinetic information such as amorphization under irradiation directly from DFT simulations due to size and time limitations. Thus, a more feasible way may be to employ molecular dynamics (MD) simulation. Unfortunately, so far no MD potential is available for U-Si to discover the underlying mechanisms. Here, we will present our recent progress in developing a U-Si potential from ab initio data. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of molecular dynamics potential for uranium silicide fuels
The use of uranium-silicide (U-Si) in place of uranium dioxide (UO{sub 2}) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels for light water reactors. First-principles based density functional theory (DFT) electronic structure calculations may provide the most reliable information. However, it is probably not possible to obtain kinetic information such as amorphization under irradiation directly from DFT simulations due to size and time limitations. Thus, a more feasible way may be to employ molecular dynamics (MD) simulation. Unfortunately, so far no MD potential is available for U-Si to discover the underlying mechanisms. Here, we present our recent progress in developing a U-Si potential. The development is based on the Tersoff type potentials for single element U and Si. The Si potential is taken from the literature and a Tersoff type U potential is developed in this project. With the primary focus on the U{sub 3}Si{sub 2} phase, some other U-Si systems such as U{sub 3}Si are also included as a test of the transferability of the potentials for binary U-Si phases. Based on the potentials for unary U and Si, two sets of parameters for the binary U-Si system are developed using the Tersoff mixing rules and the cross-term fitting, respectively. The cross-term potential is found to give better results on the enthalpy of formation, lattice constants and elastic constants than those produced by the Tersoff mixing potential, with the reference data taken from either experiments or DFT calculations. In particular, the results on the formation enthalpy and lattice constants for the U{sub 3}Si{sub 2} phase and lattice constants for the high temperature U{sub 3}Si (h-U{sub 3}Si) phase generated by the cross-term potential agree well with experimental data. Reasonable agreements are also reached on the elastic constants of U{sub 3}Si{sub 2}, on the formation enthalpy for the low temperature U{sub 3}Si (m-U{sub 3}Si) and h-U{sub 3}Si phases, and on the lattice constants of m-U{sub 3}Si phase. All these phases are predicted to be mechanically stable. The unary U potential is also tested for three metallic U phases (α, β, γ). The potential is found capable to predict the cohesive energies well against experimental data for all three phases. It matches reasonably with previous experiments on the lattice constants and elastic constants of αU. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Evaluation of Thermal Neutron Scattering Cross Section of Uranium Silicide with Ab Initio Lattice Dynamics
Uranium silicide (U 3 Si 2 ) is a candidate material for the high-density nuclear fuel in commercial light water reactors [1], [2]. Its higher uranium density, 11.3 g-U/cm3, compared to that of uranium dioxide (UO 2 ), 9.7 g-U/cm3, can improve the performance of a nuclear reactor while using low enriched uranium (LEU) and diversify the choice of cladding materials [1]–[3]. It also has a higher thermal conductivity than UO 2 , which can reduce the thermal stress on the material caused by a temperature gradient across the fuel pellet and provide a larger margin for some postulated accidents [1], [2], [4]. Furthermore, compared to U3Si, another high-density fuel candidate, it has better resistance to in-pile swelling due to less irradiation-induced rapid amorphization [1], [3]. Corresponding to its importance in nuclear engineering, many previous studies have reported the properties of U3Si2. Experiments showed that U 3 Si 2 is a paramagnetic (PM) metal, where a slight linear increase in magnetic susceptibility was measured with increasing temperature [5], [6]. In addition, thermodynamic quantities such as thermal expansion coefficient, heat capacity, and thermal conductivity were experimentally determined over a wide temperature range [1], [7], [8]. In several computational studies, ab initio atomistic simulations based on density functional theory (DFT) were performed to calculate various properties including elastic constants, electronic density of states (DOS), and phonon dispersion curves [9]–[12]. Nevertheless, thermal neutron scattering cross sections, which are critical to the prediction of the parameters in reactor physics that are ultimately related to reactor criticality, have not yet been evaluated for U3Si2. The scattering cross section can be calculated from the phonon DOS, or the energy spectrum of lattice vibrations, of the crystalline system [13], [14]. However, there is also no experimental data available for the phonon DOS of U 3 Si 2 . While some computational studies reported the phonon DOS and/or dispersion curves from ab initio simulations [9]–[12], the accuracy cannot be guaranteed because it is unclear whether the spin-polarization behavior of PM U 3 Si 2 was properly described. In the present study, the thermal neutron scattering cross section for U 3 Si 2 is evaluated for the first time by calculating the phonon DOS for U3Si2 from ab initio lattice dynamics (AILD) simulations based on DFT. First, U 3 Si 2 is modeled based on the experimental structure, and AILD simulations are performed on the modeled U3Si2 to optimize the structure. Next, AILD simulations are performed for supercells with atomic displacement to calculate Hellmann-Feynman forces. Based on the calculated forces, partial phonon DOSs for U and Si are obtained, and the thermal neutron scattering law (TSL) for U 3 Si 2 is finally evaluated. To verify the accuracy of the calculations in the present study, the calculation results are compared with experimental data on the structure and heat capacity of U3Si2 [1], [7], [8], [15].
Geometry Optimization↗
Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation
In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U 3 Si 2 ) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U 3 Si 2 has been optimized and high phase purity U 3 Si 2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ±0.06 g/cm 3 . Additional characterization of the pellets by scaning electron microscopy and X-ray diffraction has also been performed. As a result, pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Method to produce uranium silicides
The method described herein may be characterized as reacting uranium dioxide with carbon to produce uranium carbide, and, reacting the uranium carbide with a silane, a silicon halide, a siloxane, or combinations thereof, and excess hydrogen to produce uranium silicide.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium silicide
We report the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breech of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium silicide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UC and UB 2 . Presented here are the literature for the U3Si2 degradation modes, thermodynamics, and oxidation performance of the pure compound and its reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation/corrosion behavior for high uranium density fuels for use in LWRs.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Current Plans for Irradiation Testing of Low Enriched Uranium Silicide Fuel in Support of High Flux Isotope Reactor (HFIR) Conversion
Current Plans for Irradiation Testing of Low Enriched Uranium Silicide Fuel in Support of High Flux Isotope Reactor (HFIR) Conversion
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Radiation Re-solution Calculation in Uranium-Silicide Fuels
The release of fission gas from nuclear fuels is of primary concern for safe operation of nuclear power plants. Although the production of fission gas atoms can be easily calculated from the fission rate in the fuel and the average yield of fission gas, the actual diffusion, behavior, and ultimate escape of fission gas from nuclear fuel depends on many other variables. As fission gas diffuses through the fuel grain, it tends to collect into intra-granular bubbles, as portrayed in Figure 1.1. These bubbles continue to grow due to absorption of single gas atoms. Simultaneously, passing fission fragments can cause collisions in the bubble that result in gas atoms being knocked back into the grain. This so called “re-solution” event results in a transient equilibrium of single gas atoms within the grain. As single gas atoms progress through the grain, they will eventually collect along grain boundaries, creating inter-granular bubbles. As the inter-granular bubbles grow over time, they will interconnect with other grain-face bubbles until a pathway is created to the outside of the fuel surface, at which point the highly pressurized inter-granular bubbles will expel their contents into the fuel plenum. This last process is the primary cause of fission gas release. From the simple description above, it is clear there are several parameters that ultimately affect fission gas release, including the diffusivity of single gas atoms, the absorption and knockout rate of single gas atoms in intra-granular bubbles, and the growth and interlinkage of intergranular bubbles. Of these, the knockout, or re-solution rate has an particularly important role in determining the transient concentration of single gas atoms in the grain. The re-solution rate will be explored in the following sections with regards to uranium-silicide fuels in order to support future models of fission gas bubble behavior.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Mechanical Properties of Uranium Silicides by Nanoindentation and Finite Elements Modeling
Three methods were used to measure the mechanical properties of U 3 Si, U 3 Si 2 , and USi. Quasi-static and continuous stiffness measurement nanoindentation were used to determine hardness and Young’s modulus, and microindentation was used to evaluate the bulk hardness. Hardness and Young’s modulus of the three U-Si compounds were both observed to increase with Si content. In conclusion, finite elements modelling was used to validate the nanoindentation data calculated for U 3 Si 2 and estimate its yield strength.
36 MATERIALS SCIENCE↗
Milestone report on MD potential development for uranium silicide
This report summarizes the progress on the interatomic potential development of triuranium-disilicide (U 3 Si 2 ) for molecular dynamics (MD) simulations. The development is based on the Tersoff type potentials for single element U and Si. The Si potential is taken from the literature and a Tersoff type U potential is developed in this project. With the primary focus on the U 3 Si 2 phase, some other U-Si systems such as U 3 Si are also included as a test of the transferability of the potentials for binary U-Si phases. Based on the potentials for unary U and Si, two sets of parameters for the binary U-Si system are developed using the Tersoff mixing rules and the cross-term fitting, respectively. The cross-term potential is found to give better results on the enthalpy of formation, lattice constants and elastic constants than those produced by the Tersoff mixing potential, with the reference data taken from either experiments or density functional theory (DFT) calculations. In particular, the results on the formation enthalpy and lattice constants for the U 3 Si 2 phase and lattice constants for the high temperature U 3 Si (h-U 3 Si) phase generated by the cross-term potential agree well with experimental data. Reasonable agreements are also reached on the elastic constants of U 3 Si 2 , on the formation enthalpy for the low temperature U 3 Si (m-U 3 Si) and h-U 3 Si phases, and on the lattice constants of m-U 3 Si phase. All these phases are predicted to be mechanically stable. The unary U potential is tested for three metallic U phases (α, β, γ). The potential is found capable to predict the cohesive energies well against experimental data for all three phases. It matches reasonably with previous experiments on the lattice constants and elastic constants of αU.
36 MATERIALS SCIENCE↗
Conceptual Fuel Element Design Candidates for Conversion of High Flux Isotope Reactor with Low-Enriched Uranium Silicide Dispersion Fuel
Engineering design studies are underway to assess the feasibility of converting the High Flux Isotope Reactor (HFIR) to operate with low-enriched uranium (LEU) silicide dispersion (U3Si2-Al) fuel. These studies are supported by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. A systematic approach employing neutronic and thermal-hydraulic analyses have been performed with the ORNL Shift and HFIR Steady State Heat Transfer Code tools, respectively, to predict reactor performance and thermal safety margins for proposed LEU3Si2-Al fuel designs. The design process was initiated by generating an optimized design with fabrication features identified from previous studies that result in excellent performance and safety metrics. The approach continued by substituting a single fabrication feature anticipated to be difficult to manufacture with another feature expected to perform an analogous function to that of the removed feature. Four conceptual fuel element design candidates, with various fabrication features, for conversion of HFIR with 4.8 gU/cm3 LEU3Si2-Al fuel have been generated and shown to meet pre-defined performance and safety metrics. Results to date indicate that HFIR could convert with the subject fuel system and meet performance and safety requirements if, among other considerations, fabrication of the specific design features are demonstrated and qualification of the fuel is complete under HFIR-specific conditions.
Chandler, David↗
Neutronic and thermal-hydraulic feasibility studies for High Flux Isotope Reactor conversion to low-enriched uranium silicide dispersion fuel
An iterative design process involving neutronic and thermal-hydraulic modeling and simulation has been employed to assess the feasibility of converting the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) silicide dispersion fuel. ORNL is funded by the National Nuclear Security Administration to evaluate HFIR conversion. Previous HFIR conversion studies focused on U-10Mo monolithic fuel; however, due to potential fabrication issues with the complex HFIR U-10Mo fuel design, ORNL is evaluating U 3 Si 2 -Al dispersion fuel as an alternative LEU fuel system.Fueled by 10.1 kg of HEU and operated at 85 MW, HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world. Retrofitting a compact, high-power density, HEU-based core with LEU is a challenging problem to solve, especially when considering the conversion requirements. Neutronic and thermal-hydraulic analyses were performed with Shift and HSSHTC, respectively, to predict reactor performance and thermal safety margins. A number of designs were proposed and evaluated using an iterative approach in an effort to show that reactor performance could match that obtained using HEU fuel and that thermal safety margins were adequate. Furthermore this study concludes that conversion of HFIR with U 3 Si 2 -Al LEU fuel is feasible if, among other requirements, the fuel meat region is centered and symmetric about the fuel plate thickness centerline, the active fuel zone length is increased from 50.80 cm to 55.88 cm, the proposed fabrication tolerances can be met, and the fuel can be qualified for HFIR conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Transient testing of uranium silicide fuel in zircaloy and silicon carbide cladding
Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U 3 Si 2 fuel pellets were tested under reactivity-initiated accident conditions in dry helium capsules. These tests were performed with total energy depositions sufficient to induce melting of the U 3 Si 2 fuel pellets based on available thermophysical property data. The instrumented experiments utilized in-pile thermocouples and infrared pyrometers to measure the cladding temperature during the transient, which are in reasonable agreement with predicted peak cladding temperatures using a simple BISON fuel performance model. This same model is then used to predict expected cladding stresses during the transients in areas where pellet cladding interaction is expected. It was shown that for transient energy depositions up to 528J/g rod like geometry of the test articles was maintained. While it is acknowledged that more confirmatory testing is required in more prototypic environments, these first of their kind experiments have demonstrated that unirradiated U 3 Si 2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528J/g.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Automated reactor physics analysis framework of High Flux Isotope Reactor low-enriched uranium silicide dispersion fuel designs
The High Flux Isotope Reactor (HFIR) is a versatile research reactor that provides one of the highest steady-state neutron fluxes of any reactor in the world. The HFIR reactor physics team investigated the conversion of the current 93 wt% highly enriched uranium U 3 O 8 -Al dispersion fuel to a 19.75% low-enriched uranium (LEU) U 3 Si 2 -Al dispersion fuel. The team continuously develops a Python module to streamline the analysis steps required for an LEU core design to ensure reproducible and agile design iteration. The Python module automates the data processing between analysis steps and automates the input perturbation for branch calculations and design changes. The automated framework has proven to significantly increase the efficiency and reproducibility of the reactor physics team to design High Flux Isotope Reactor (HFIR) LEU cores and thoroughly analyze performance metrics, safety metrics, and thermal safety margins. Consequently, the team can now respond rapidly to fuel fabrication engineer and thermal-hydraulic-structural analyst requests. Numerous combinations of LEU fuel designs are explored, of which two LEU fuel designs are presented here in this paper: a low density silicide design, and a high-density silicide design. Results show that both designs meet or exceed safety and performance metrics with exception for minor differences caused by the hardened spectrum from LEU.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗