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Condensates, systems composed of bosons that “crystallize” on the basis of the weak interactions between the wave functions of their constituent, ground state particles, are fundamental to the modern physics of emergent phenomena.
An experiment designed to compare neutron noise analysis techniques on subcritical assemblies using differing neutron detector designs was developed as a collaborative effort between the Institut de radioprotection et de surete nucleaire (IRSN), Los Alamos (LANL), Lawrence Livermore (LLNL), and Sandia National Laboratories. This experiment has the final goal of characterizing neutron noise analysis accuracy in determining multiplication in light water systems for various neutron detection systems. This experiment utilized the Seven Percent Critical Experiment (7uPCX) facility at Sandia National Laboratories. The 7uPCX is an assembly designed to investigate light water systems with fuel enrichment above 5% and is able to use square or hex grid plates with varying pitches.
Here in this study, we investigated the dislocation loop types formed in the proton-irradiated uranium-nitrogen-oxygen (U-N-O) system, which involves uranium mononitride (UN), uranium sesquinitride (a-U2N3), and uranium dioxide (UO2) phases. The dislocation loop formation is examined using specimens irradiated at 400°C and 710°C. Based on the detailed transmission-based electron microscopy characterization with i) the morphology-based on-zone and ii) the invisibility-criterion based two-beam condition imaging techniques, only a single type of dislocation loop in each phase is found: a/2?110?, a/2?111?, or a/3?111? dislocation loops in UN, a-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in UN phases indicate a critical loop size of ~ 6 nm, above which perfect loops are energetically favorable. This could explain the absence of faulted loops in the experimental observation of the irradiated UN phase at two temperatures. This work will enhance the understanding of irradiation induced microstructural evolution for uranium nitrides as advanced nuclear fuels for the next-generation nuclear reactors.
Here, the use of fluorination as a rapid separation technique for quantification of trace-mass, short-lived isotopes is illustrated. Single crystal 238 UO 2 was exposed to a 14 MeV neutron source to produce mixed fission and activation products. Selective removal of trace volatile fluorides from non-volatile ones produced two results of interest. A decrease in background produced better detection and quantification of select analytes across the gamma spectrum. Second, at the very low burnup condition described, the volatility behaviors of common isotopes were altered relative to their well-established behaviors in higher burnup metal and metal oxide fuels.
The Transformational Challenge Reactor (TCR) program aims to demonstrate a revolutionary design approach enabled by advanced manufacturing and data analytics in the design of nuclear reactors. This article discusses scoping analyses of preconceptual designs to inform TCR design decisions and the evaluation of sensitivities and uncertainties on postulated transient scenarios. The applicability of the systems codes TRACE and RELAP5-3D to TCR transient analysis are examined, and RELAP5-3D models are used to examine the transient response of two candidate core designs at multiple power levels. Then, the uncertainty quantification code RAVEN is used to quantify the effect of several design parameters on reactivity-initiated accident (RIA) progression at hot zero power (HZP) and hot full power (HFP) as well as to assess the impact of uncertainties in transient parameters for the pressurized loss of forced cooling (PLOFC).When results were compared, TRACE and RELAP5-3D showed good agreement in their ability to predict system behavior, but RELAP5-3D calculations were closer to analytical predictions for the RIA. Furthermore, models for a PLOFC accident in two designs (a UO2 and TRISO core) at multiple power levels showed greater temperature margins for the TRISO core at all power levels. Using this information, along with other scoping analyses and constraints, the TCR design team selected a power level of 3 MWth and a TRISO-based core design. For this design RAVEN was applied to vary RIA parameters in RELAP5-3D models at HZP and HFP to understand the effect on figures of merit such as peak power, fuel and coolant temperature, and energy deposition. This sensitivity study found that the inserted reactivity worth was the most important parameter controlling all figures of merit, but for insertion up to 1.5$ no failure of TCR fuel is anticipated. For constant reactivity insertions, the magnitude of the fuel temperature coefficient was found to have the greatest effect on all figures of merit under most circumstances. These results not only demonstrate the anticipated robust safety of the proposed TCR fuel form but also provide a reference for future metal-hydride moderated systems to understand RIA behavior. In the PLOFC, the impact of heat transfer enhancement due to wavy flow channel effects dominated the variance in peak temperatures, and variations in heat exchanger elevation provided the greatest control on natural circulation flow rate. Additionally, no fuel particle failure is anticipated in the PLOFC.
Here we present crystallographic analyses of high-energy X-ray diffraction data on polycrystalline UO2 up to the melting temperature. The Rietveld refinements of our X-ray data are in agreement with previous measurements, but are systematically located around the upper bound of their uncertainty, indicating a slightly steeper trend of thermal expansion compared to established values. This observation is consistent with recent first principles calculations.
Uranium dioxide (UO2) pellets with controlled microstructures were densified up to 93.4% of their theoretical density in less than 25 minutes at a furnace temperature of 873 K, utilizing controlled current-rate alternating current (AC) flash sintering (FS). Using this AC-FS method it was possible to control the sintering rate and thermal gradients, resulting in dense pellets with no appreciable hourglassing and good mechanical integrity. Moreover, the apparent sintering activation energy for FS and for conventionally sintered samples was estimated to be 108 kJ mol -1 and 380 kJ mol -1 , respectively using the master sintering curve method. The apparent activation energy for FS was remarkably close to those reported in the literature for spark plasma sintering of UO 2 . Furthermore, both these field assisted sintering methods utilize fast heating rates and electrical effects that are likely enhancing the grain boundary diffusion mechanism. Finally controlled current-rate AC-FS has been demonstrated as a technological advancement, capable of producing ceramic nuclear fuels in a fraction of the conventional processing time.
A hierarchical multi-scale approach is pursued in this work to investigate the influence of porosity, pore and grain size on the intergranular brittle fracture in UO2. In this approach, molecular dynamics simulations are performed to obtain the fracture properties for different grain boundary types. A phase-field model is then utilized to perform intergranular fracture simulations of representative microstructures with different porosities, pore and grain sizes. In these simulations the grain boundary fracture properties obtained from molecular dynamics simulations are used. The responses from the phase-field fracture simulations are then fitted with a stress-based brittle fracture model usable at the engineering scale. This approach encapsulates three different length and time scales, and allows the development of microstructurally informed engineering scale model from properties evaluated at the atomistic scale.
Uranium Ore Concentrate (UOC, often called yellowcake) is a generic term that describes the initial product resulting from the mining and subsequent milling of uranium ores en route to production of the U-compounds used in the fuel cycle. Depending on the mine, the ore, the chemical process, and the treatment parameters, UOC composition can vary greatly. With the recent advent of handheld spectrometers, we have chosen to investigate whether either commercial off-the-shelf (COTS) handheld devices or laboratory-grade Raman instruments might be able to i) identify UOC materials, and ii) differentiate UOC samples based on chemical composition and thus suggest the mining or milling process. Twenty-eight UOC samples were analyzed via FT-Raman spectroscopy using both 1064 nm and 785 nm excitation wavelengths. These data were also compared with results from a newly developed handheld COTS Raman spectrometer using a technique that lowers background fluorescence signal. Initial chemometric analysis was able to differentiate UOC samples based on mine location. Additional compositional information was obtained from the samples by performing XRD analysis on a subset of samples. The compositional information was integrated with chemometric analysis of the spectroscopic dataset allowing confirmation that class identification is possible based on compositional differences between the UOC samples, typically involving species such as U3O8, α-UO2(OH)2, UO4•2H2O (metastudtite), K(UO2)2O3, etc. While there are clearly excitation λ sensitivities, especially for dark samples, Raman analysis coupled with chemometric data treatment can nicely differentiate UOC samples based on composition and even mine origin.
Abstract Proton (H + ) irradiation effects in polycrystalline UO 2 have been studied. The irradiation was carried out using three ion energies and two different ion fluxes at 600 °C. Scanning electron microscopy (SEM) investigations showed that significant surface flaking took place. Focused ion beam (FIB) milling in SEM was successfully applied for extracting lamellas from uneven blistered surfaces for transmission electron microscopy (TEM) investigations allowing detailed investigations for the degradation mechanisms. High-resolution TEM for the flaked UO 2 surfaces revealed that the implanted H + formed sharp two-dimensional cavities at the peak ion-stopping region instead of diffusing to the matrix. The resulting lateral stress likely caused UO 2 surface deterioration in good agreement with previous blistering and flaking studies on crystalline materials. Graphical abstract
Here, this work aims to simulate the restructuring behavior observed in different regions of high burnup fuel, providing a first-of-its-kind restructuring model for the dark zone and rim region of high-burnup UO 2 fuel. We employed a grand-potential-based phase-field model to concurrently evaluate subgrain formation and the growth of fission gas bubbles within the fuel. An energy-based subgrain formation criterion was introduced to simulate the restructuring process. The effects of different initial conditions and different modeling parameters were systematically studied to capture how each of these parameters influences the characteristics of the restructured fuel. Subgrain formation was observed to begin around existing fission gas bubbles and proceed toward triple junctions, grain boundaries, and grain interiors. Restructuring was demonstrated to be influenced by a combination of initial dislocation densities, burnup rate, subgrain formation rate, and temperature. Under a given subgrain formation rate, the rate of restructuring increases with rising fuel temperature. A restructuring bias was observed within the microstructure, due to the variation in defect accumulation when comparing different grains. Microstructures corresponding to the dark zone and rim region can be obtained by parameterizing the model with the appropriate defect production rate, as determined based on the burnup rate and temperature. Furthermore, bubble size and distribution do not significantly affect the rate of restructuring. The predicted microstructures are consistent with experimental observations of the restructured regions. Finally, we present a correlation demonstrating the evolution of the restructuring volume fraction as a function of local burnup.
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Techniques were developed to measure properties at sub-grain scales using depleted Uranium Oxide (d-UO2) samples heat-treated to obtain different grain sizes and oxygen stoichiometries, through three main tasks: 1) sample processing and characterization, 2) microscale and conventional testing and 3) modeling. Grain size and crystallography were characterized using Scanning Electron Microscopy (SEM), in conjunction with Electron Backscattering Diffraction (EBSD) and Electron Channeling Contrast Imaging (ECCI). Grains were then carefully selected based on their crystallographic orientations to perform ex-situ micromechanical tests with samples machined via Focused Ion Beam (FIB), with emphasis on micro-cantilever bending. These experiments were performed under controlled atmospheres, to insure stoichiometry control, at temperatures up to 700 °C and allowed measurements involving elastic (effective Young’s modulus), plastic (critical resolved shear stresses) and creep (creep strain rates) behavior. Conventional compression experiments were performed simultaneously to compare with the ex-situ measurements and study potential size effects. Modeling was implemented using anisotropic elasticity and inelastic constitutive relations for plasticity and creep based on kinematics and kinetics of dislocation glide that account for the effects of crystal orientation, and stress. The models will be calibrated and validated using the experimental data. This project provided insight on correlations among stoichiometry, crystallography and mechanical behavior in advanced oxide fuels, provided valuable experimental data to validate and calibrate mesoscale fuel performance codes and also a framework to measure sub-grain scale mechanical properties that should be suitable for use with irradiated samples due to small volumes required. The goals and metrics of the ongoing study of thermo-mechanical behavior in depleted uranium dioxide (d-UO 2 ) outlined in this project have been concluded successfully, resulting in: 1) the successful fabrication, processing, and characterization of large-grained samples with various orientations (up to and including single crystals) having stoichiometric and hyper-stoichiometric O/U ratios; 2) formulation, calibration, and validation of a crystal plasticity constitutive model to describe the creep deformation of UO 2 at the sub-grain length scale (single crystal level) at intermediate temperatures; 3) the successful calibration of a crystal plasticity constitutive model to describe the elasto-plastic deformation of microcantilever beams, also at moderate temperatures. Samples were prepared from natural uranium oxide powder of production-quality provided by Areva. The powder was pressed in a die to a pressure of 100 MPa to produce green pellets with no sintering aids, lubricants, or any other additives. The green pellets were then heated up to 1700 °C under ultra-high purity argon atmosphere (~1 ppm O2). The atmosphere was then changed to 79% Argon, 21% O 2 and the temperature was held at 1700 °C for 2 hours to sinter the pellets under oxidative conditions [1] that are known to increase grain growth kinetics in UO 2 [2]. Samples were then cooled down under Ar-4%H2 atmosphere to reduce the samples back to stoichiometric UO 2 . For macro-scale procedures, testing of UO 2 samples with large grains was performed at 1200 °C using a modified load frame capable of applying dead-weight loads to ensure constant stress conditions, while displacement of the sample produced by the applied load was measured with high precision micrometers to obtain strains. Stress steps were used during testing and the strains were monitored to measured creep strain rates under steady state for each level of stress used, so that stress exponents could be obtained. The results of the mechanical testing, along with sample geometry and crystal orientation of the grains in the samples, as well as post-test sample characterization were used to formulate a viscoplastic model to account for steady state (stage II) creep behavior, along with basic assumptions from crystal plasticity and kinematic constraints due to testing fixtures. In the micro-scale, testing of microcantilever beams at temperatures ranging from 25 to 570 °C was performed in-situ with a scanning electron microscope with a special attachment to apply load and measure displacement while the samples were at temperature. The load-displacement curves obtained showed linear behavior before fracture for all temperatures attempted except 570 °C, where clear deviations from non-linearity were observed before fracture. These deviations were consistently observed for all samples tested for a given orientation. A viscoplastic model was used to account for the presence of inelastic strain, along with basic assumptions from crystal plasticity and beam theory. These models were kept as simple as possible, and results from tests performed in a set of samples with a given crystal orientation were used to calibrate the material constants for the model, while results from a different sample set were then used for validation, thus satisfying the conditions of all main tasks within the parameters of this project. Details of these efforts are outlined in this report.
This report describes the salt chemistry and sedimentation studies that were completed in FY23 as part of the Pyrochemical Fuel Cycles – ANL project. The primary objective of these activities was to employ electrochemical methods in order to quantify and gain insights into the reaction mechanisms associated with the interaction of O2 and moisture impurities with the molten chloride salts used for the pyrochemical processing of nuclear materials. If not monitored and controlled, the ingression of these atmospheric impurities can lead to changes in the salt redox conditions and initiate the formation of oxide particles, potentially resulting in operational and safeguards challenges. By elucidating the mechanisms underlying the formation of oxide particles, we can effectively monitor the conditions of the salt and design systems to mitigate the influence of atmospheric O2 and H2O. To achieve these goals, we designed a test system that allowed for precisely controlled gas ingressions into the molten salt vessel. This system included on-line monitoring provided by electrochemical probes that enable near real-time measurements of the salt conditions throughout the course of the experiment. Using this system, we systematically varied experimental conditions, including O2/Ar flow rates and O2 concentrations, to comprehensively understand their impact on solid particle formation. We also used a particle size and shape analyzer to characterize the particles that were generated. Tests in FY23 concentrated on the formation of CeO2 particles, but we also prepared a separate apparatus targeting UO2 particles for use in FY24. We additionally conducted extensive modeling activities in support of this work. This included multiphysics simulations of the gas ingressions along with the development of a machine learning workflow to enable accelerated molecular dynamics modeling of the molten salt chemistry of LiCl-KCl-UCl3. The combination of experimental and modeling tools has allowed us to begin to get a more complete understanding of the reactions that occur when atmospheric ingressions occur in pyroprocessing systems.
FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.
Dense nano-sized UO 2+x pellets are synthesized by spark plasma sintering with controlled stoichiometries (UO 2.03 and UO 2.11 ) and grain sizes (~100 nm), and subsequently isothermally annealed to study their effects on grain growth kinetics and microstructure stability. The grain growth kinetics is determined and analyzed focusing on the interaction between grain boundary migration, pore growth and coalescence. Grains grow much bigger in nano-sized UO 2.11 than UO 2.03 upon thermal annealing, consistent with the fact that hyper-stoichiometric UO 2+x is beneficial for sintering due to enhanced U ion diffusion from excessive O ion interstitials. The activation energies of the grain growth for UO 2.03 and UO 2.11 are determined as ~1.0 and 1.3~2.0 eV, respectively. As compared with the micron-sized UO 2 in which volumetric diffusion dominates the grain coarsening with an activation energy of ~3.0 eV, the enhanced grain growth kinetics in nano-sized UO 2+x suggests that grain boundary diffusion controls grain growth. Lastly, the higher activation energy of more hyper-stoichiometric nano-sized UO 2.11 may be attributed to the excessive O interstitials pinning grain boundary migration.
In this study, the intergranular fracture behavior of UO2 was studied by molecular dynamics simulations using the Basak potential. In addition, the constitutive traction-separation law was derived from atomistic data using the cohesive-zone model. In the simulations a bicrystal model with the (100) symmetric tilt E5 grain boundaries was utilized. Uniaxial tension along the grain boundary normal was applied to simulate Mode-I fracture. The fracture was observed to propagate along the grain boundary by micro-pore nucleation and coalescence, giving an overall intergranular fracture behavior. Phase transformations from the Fluorite to the Rutile and Scrutinyite phases were identified at the propagating crack tips. These new phases are metastable and they transformed back to the Fluorite phase at the wake of crack tips as the local stress concentration was relieved by complete cracking. Such transient behavior observed at atomistic scale was found to substantially increase the energy release rate for fracture. Insertion of Xe gas into the initial notch showed minor effect on the overall fracture behavior.
In this report, results from 1800°C safety testing and post-safety-test PIE of AGR-2 UCO Compact 2-3-2 and 1700°C safety testing of AGR-2 UO2 Compact 3-4-1 are discussed and compared to previously reported results from 1800°C safety testing of AGR-2 UCO Compact 5-4-1 and 1600°C safety testing of compact “twins”: AGR-2 UCO Compact 2-3-1 and AGR-2 UO 2 Compact 3-4-2. These compact pairs are called “twins” because their symmetric locations in the irradiation test train resulted in similar irradiation conditions. Table 1 shows the calculated burnup in percent fissions per initial metal atom (FIMA), the fast neutron fluence (neutron energies > 0.18 MeV), and the average compact temperatures during irradiation for these compacts. Capsule 2 Compacts 2-3-2 and 2-3-1 were intentionally irradiated at notably higher temperature than compacts in the other AGR-2 capsules or those previously studied in the AGR-1 irradiation test. The Capsule 2 compacts provide insight into the irradiation and post-irradiation safety test performance at higher temperatures than typical HTGR operating temperatures.