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Measuring the Electrochemical Corrosion of 233U-doped UO2 in a Microfluidic Device
We have developed a microfluidic, ‘particle-attached microfluidic electrochemical’ device, or PAMEC, that uses microgram quantities of radioactive materials, such as uranium dioxide (UO2), rather than bulk specimens to both observe electrochemical changes and measure spectroscopic properties. The PAMEC device reduces the need for running experiments in a shielded facility and allows for in-operando, real-time investigations of electrochemical properties and morphological changes of micrograms of particles. This novel design offers a promising platform to investigate spent nuclear fuel (SNF) and other nuclear materials under various controlled condition, and to provide insights to nuclear waste management, and fuel performance
Effect of Added Gadolinium Oxide on the Thermal Air Oxidation of Uranium Dioxide
To develop a more reliable and stable UO2-based nuclear fuel, the Pacific Northwest National Laboratory (PNNL) investigated modifying fuel with several soluble lanthanides and zirconium. This article provides the results of these studies investigating gadolinium doping at levels up to 10 mass%. The authors characterized and compared commercially- and PNNL-prepared gadolinium-doped UO2 to determine the oxygen-to-metal ratio, elemental distribution, chemical composition, physical appearance, lattice parameters, and grain structure using atomic force microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy, and X-ray diffractometry. After confirming PNNL-prepared UO2 and (Gd,U)O2 were similar to commercially prepared UO2 and (Gd,U)O2, we measured the thermal behavior of these gadolinium-doped UO2 materials to air oxidation using differential scanning calorimetry and thermogravimetric analysis. Addition of gadolinium stabilized the first oxidation product U4¬O9/U3O7 and slowed the subsequent oxidation to U3O8. Comparison of our measured two-step oxidation of UO2 to U3O8 at 325°C to common gas/solid kinetic reaction models found that each oxidation step is best described as a convolution of kinetic behaviors; the gadolinium insertion into the UO2 lattice enforces a significant alteration in mechanism and oxidation rate. Noticeable changes in mechanism become apparent between 1% and 3% gadolinium content.
UO2 microstructural evolutions induced by Ni, Mo, and W dopants for intentional forensics
The concept of tagging nuclear fuel with a chemical barcode to enable forensics analysis across the nuclear fuel cycle is an area of active investigation, particularly to ensure fabrication viability without disrupting current fuel performance. This study explored the feasibility of using Ni, Mo, and W isotopic double-spikes as dopants in UO2 fuel from the perspective of fuel fabrication. Doped UO2 pellets were produced using conventional fuel fabrication processes, including powder mixing, sieving, pressing, and sintering in a reductive atmosphere. Two composition levels, 100 and 1000 ppm, were evaluated for each dopant element with isotopic double-spike configurations. For the Ni system, additional dopant concentrations of 250 and 500 ppm were produced with nonperturbed isotopic ratios. The results demonstrated that successful incorporation of Ni, Mo, and W double-spikes into UO2 pellets occurred with minimal shift in final density or dopant loss during pellet fabrication. Isotopic analysis confirmed the presence of the double-spike signature even when diluted with natural isotopic material in ratio of 1:5 in the fabrication process. Microstructural examinations revealed different impacts on grain size compared with undoped UO2. This study showed that Ni incorporation up to ∼500 ppm promoted moderate grain growth, whereas the Mo and W systems caused grain size reduction at all concentrations. Changes in the UO2 lattice parameter as a function of composition were detected exclusively for Ni up to 500 ppm, indicating that the Ni solid solution was the main factor for the observed grain growth. Insoluble (Mo and W) or supersaturated (Ni > 500 ppm) conditions produced grain size reduction. The Ni-doped pellets in the solution range resulted in a final microstructure within fuel specifications, demonstrating its potential benefits of employing complex dopant systems for potential nuclear forensic applications.
An evaluation of tri-valent oxide (Cr 2 O 3 ) as a grain enlarging dopant for UO 2 nuclear fuels fabricated under reducing environment
A study was performed to evaluate the microstructure and crystallography of nominally 500–2000 Cr 2 O 3 -doped UO 2 fabricated in a temperature range of 1150–1750°C under reducing experimental conditions. We observed an increase in grain size of the samples with the increase in heat treating temperature as expected. For a given sintering temperature (1700–1750°C), an increase in the grain size was also observed with the increase in Cr 2 O 3 concentration up to a value of ~1000–1200 wppm. A decrease in fission gas release as a function of grain size was estimated for the Cr 2 O 3 -doped UO 2 samples assuming specified post-irradiation annealing conditions. A nearly linear decrease was obtained in the lattice parameter of the Cr 2 O 3 -doped UO 2 fcc phase with the increase in Cr 2 O 3 concentration, especially up to a nominal value of 1000 wppm. The lattice parameter decrease was also persistent with the increase in the average grain size as a result of addition of Cr 2 O 3 into the UO 2 lattice. An increase in the crystallite size and a decrease in the microstrain of the $fcc$ phase were observed with the increase in the average grain size of the samples, indicating a higher crystallinity of the Cr 2 O 3 -doped samples than that of the undoped UO 2 sample.
Microstructural and crystallographic effects of sol-gel synthesized Ti-doped UO 2 sintered under reducing conditions
Titanium (Ti)-doped UO 2 microspheres of three different Ti concentrations (1000, 2000, and 4000 wppm) were synthesized using an internal gelation process. The microspheres were pressed into pellets, and a two-step heat treatment was applied to form monolithic cylindrical pellets with high densities (≥95%TD). Microstructure of these samples consisted of equiaxed grains with >300% increase in average grain size compared to the undoped UO 2 pellets. Secondary Ti-rich chemical phases corresponding to a liquid eutectic formed during sintering were observed at grain boundaries of UO 2 for samples doped with 4000 wppm Ti. Furthermore, these Ti-rich chemical phases were not observed in 1000 or 2000 wppm Ti samples at microscale using electron microscopy investigations. The 0.02–0.04% lower lattice parameter values for the Ti-doped UO 2 samples compared to the undoped UO 2 confirms the incorporation of Ti into the UO 2 lattice.
Diffusion in undoped and Cr-doped amorphous UO 2
UO 2 fuel pellets are often doped with chromium oxide to obtain favourable properties such as higher density, improved thermal stability, large grain sizes, improved pellet-clad interaction margins, and increased fission gas retention during transients. Chromium has a low solubility limit in UO 2 , with past experimental work reporting solubility limits ranging between 0.004 to 0.06 wt.% Cr. Due to its low solubility, segregation of Cr ions to the grain boundary may occur. Further, the complexity of these boundaries may be high as observed in other ceramics resulting in disordered or amorphous regions along the boundary, affecting a range of material and operational properties of the fuel pellet. To assess these disordered regions, in this work we study amorphous undoped and Cr doped UO 2 systems (containing 10–50 at.% Cr 3+ ) that have been modelled using classical molecular dynamics methods incorporating Cr 3+ into the well-used CRG potential library. Diffusion coefficients, pre-exponential factors, and activation energies for diffusion were computed for oxygen ions, assessing the impact of structure and extrinsic species on migration. Oxygen diffusion was observed to be much faster in the undoped amorphous system compared to its crystalline counterpart. Oxygen diffusion in doped systems decreased with increasing Cr concentration, highlighting the importance of additives to retain fission products and other migratory species.
Statistical fracture behavior of doped UO 2 using a ball-on-ring equibiaxial flexure test method
Metal oxide dopants, such as titanium and chromium oxides, have garnered considerable attention for their potential to increase grain size (≥ 30 µm) in UO 2 fuel, purportedly enhancing fission gas retention during reactor operation. Fuel performance is significantly impacted by fuel fracture behavior, so it is important to understand the effects of enhanced grain size and dopant content on UO 2 fuel fracture. UO 2 pellets were doped with 0.1 wt% TiO 2 and 0.3 wt% Cr 2 O 3 to alter density and grain size. Inductively coupled plasma mass spectroscopy measured dopant levels pre- and post-sintering. X-ray diffraction revealed lattice changes and microstrain via Rietveld refinement. Field emission scanning electron microscopy determined grain sizes of approximately 30 µm for TiO 2 doping and 7 µm for Cr 2 O 3 doping. Transverse rupture strength tests were performed on over 30 samples per dataset to obtain characteristic strength and Weibull modulus. Results indicate no statistical difference in fracture strength between 0.1 wt% TiO 2 doped UO 2 and undoped UO 2 , while 0.3 wt% Cr 2 O 3 doped UO 2 exhibited a 20% decrease in fracture strength. Doped UO 2 samples also showed reduced Weibull modulus compared to undoped UO 2 , suggesting increased scatter in fracture strength. This study's findings suggest that titanium and chromium oxide doping in UO 2 , regardless of grain size, induce residual stresses, decreasing fracture strength and increasing variability in fracture behavior.
The influence of iron incorporation on the microstructure and crystal chemistry of UO 2 fuel
The incorporation of dopants into UO 2 is of interest for both doped advanced fuel and nuclear forensics research. Suitable forensics dopants, or taggants, for nuclear fuels are ones that do not adversely affect the performance of the parent material while still being traceable during key stages of the fuel’s life cycle. In this paper, we present our efforts to understand the effects of doping UO 2 with Fe 2 O 3 under multiple parameters, such as Fe concentration (0–3000 wppm), sintering temperature (1673 K, 1773 K, and 1973 K), sintering time (2–12 h), and atmosphere (reducing and inert). Here, we investigated densification behavior, microstructure evolution, and lattice defects based on the geometric densities, scanning electron microscopy (SEM), X-ray diffractograms, and Raman spectra. The effect of Fe on the microstructure and crystal chemistry of UO 2 is very complex, especially at the intermediate temperature (1773 K), where we found that densification kinetics are enhanced and that lattice contraction occurs for doping concentrations greater than 1000 wppm Fe.
Sensitivity of UO 2 fuel performance to microstructural evolutions driven by dilute additives
Use of dilute additives to nuclear fuel is being considered to increase the security of commercial fuel management through traceability of fabricated fuel elements. Taggants, as additives are denoted when included for traceability purposes, may also improve fuel performance, as demonstrated in Cr-containing uranium dioxide as described in the literature, and they may also improve fuel safety. In fact, studies have shown that some additives affect fuel material properties such as grain size and density after sintering. Given the possible range of elements that could be used as additives, the impact of such fuel property variations on the fuel’s thermomechanical behavior becomes relevant. These effects can be evaluated through a sensitivity study of standard fuel models to analyze changes in these properties using a fuel performance code. In this work, the BISON code is being used to investigate these effects through a 2D axisymmetric model of smeared UO 2 fuel pellets and ZIRLO® cladding under realistic pressurized water reactor core irradiation conditions. Here, randomly sampled densities and grain sizes within specified ranges are used as input parameters in the simulations, and several fuel model-related outputs are evaluated. The thermomechanical response of the cladding is also addressed in this study. The simultaneous variation of both input parameters offers a more comprehensive path to identify key sensitivities. Outputs explored include temperature, fission gas release, creep, and radial stress. Results show that although most of these outputs are sensitive to grain size to a certain extent, density mainly affects fuel temperature and elastic strain. Furthermore, sensitivities can vary depending on the radial position within the fuel pellet.
BISON Development and Validation for Priority LWR-ATF concepts
Over the years, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) (2015-2018, 2020) and Consortium for Advanced Simulation of Light Water Reactors (CASL) (2019) programs have provided support for development of Accident Tolerant Fuel (ATF) material models in the BISON fuel performance code. Since the beginning, the goal has been to utilize a multiscale modeling approach to gain a physical understanding of the fuel concepts of interest and to develop mechanistic models in the absence of a large amount of experimental data. This work builds upon that of previous years. In particular we present newly updated fission gas release models for both gas behavior in Cr 2 O 3 -doped UO 2 and U 3 Si 2 fuels, and a new creep model for U 3 Si 2 . The validation exercises completed last year are revisited with the latest models and the results updated. A brief summary of recent modeling activities for FeCrAl cladding is also provided.
Multiscale Modeling of Cr 2 O 3 -doped UO 2 Creep and Fracture
The pursuit of UO 2 fuel with various dopants (e.g., Cr 2 O 3 , Al 2 O 3 ) for improved accident tolerance due to larger grain sizes and suspected reductions in fission gas release and creep requires advanced modeling and simulation tools. These tools enable rapid multiscale development and understanding of material behavior in regimes where experimental measurements may be lacking. This work builds upon previous multiscale modeling efforts for fission gas behavior of Cr 2 O 3 -doped UO 2 to begin looking at the effects of dopants on mechanical properties with a particular focus on creep and fracture. The results indicate that dislocation climb is the most dominate creep mechanism at almost all temperatures with sensitivity analyses further providing evidence in support of that claim. Limited experimental data from the Advanced Fuels Campaign (AFC) program in the U. S. Department of Energy on tensile strength of Cr 2 O 3 -doped UO 2 indicates that fracture may be less severe in doped specimens. The results on creep and fracture are preliminary given the large uncertainty associated with the models.
Mesoscale modeling to inform Bison models of accident tolerant fuel concepts
U 3 Si 2 and doped UO 2 have been under investigation in recent years as potential accident-tolerant fuel concepts. In this report, lower-length scale studies of these fuel concepts carried out during Fiscal Year 2020 are detailed. U 3 Si 2 is a potential accident-tolerant fuel that shows promise due to its high thermal conductivity and higher uranium density relative to UO 2 . However, its swelling and fission gas release behavior in light water reactor (LWR) conditions is relatively unknown. To provide mechanistic insight and determine parameters for engineering-scale fuel performance modeling of pellet-form U 3 Si 2 , phase-field simulations of the growth, interconnection, and venting of intergranular fission gas bubbles were performed. The fractional coverage of the grain boundary and the fraction of bubble area that is vented were calculated as a function of time. From the simulation data, the fractional grain boundary coverage at saturation, an important parameter needed in engineering-scale modeling of swelling and fission gas release, was determined. Multiple simulations were run to determine the uncertainty in the calculated value. The effect of model assumptions and input parameters that are not well known was evaluated. Simulation results are compared to related theoretical and computational work. Based on the simulation results, a value of 0.60 for the fractional grain boundary coverage at saturation is recommended for U 3 Si 2 fuel.
Compare Mechanistic Predictions for Doped UO 2 Mechanical Response and Other Properties with Empirical Models and Experimental Measurements
The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program develops predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity, high resolution fuel performance tool at the engineering scale. As part of its development, additional modeling capabilities and improvements have been developed for relevant fuel forms. In this work, a fuel creep deformation model for Cr-doped fuel has been implemented into BISON, along with improvements to the empirical UO 2 fuel creep model based on experimental data and improvements to the radial power factor calculation for doped fuels. This work allows for more accurate simulation analyses for both UO 2 and doped-UO 2 fuels.
Multiscale Modeling of Cr 2 O 3 -doped UO 2 Creep and Fracture (Rev.1)
The pursuit of UO 2 fuel with various dopants (e.g., Cr 2 O 3 and Al 2 O 3 ) for improved accident tolerance due to larger grain sizes and suspected reductions in fission gas release and creep requires advanced modeling and simulation tools. Such tools enable rapid multiscale development and enhanced understanding of material behavior in regimes for which experimental measurements may be lacking. This work builds on previous multiscale modeling efforts targeting the fission gas behavior of Cr 2 O 3 -doped UO 2 , in order to begin exploring the effects of dopants on mechanical properties, with a particular focus on creep and fracture. The results indicate that dislocation climb is the most dominate creep mechanism at almost all temperatures, and sensitivity analyses provide evidence in further support of this claim. The limited experimental data generated by the U.S. Department of Energy’s Advanced Fuels Campaign (AFC) on the tensile strength of Cr 2 O 3 -doped UO 2 indicate that fracture may be less severe in doped specimens. The results on creep and fracture are preliminary, given the large degree of uncertainty associated with the models.
Application of physics-based ATF Material Models in Engineering Scale Assessment [Slides]
Presentation for a meeting on accident tolerant fuels. Brings together information from already cleared reports.
Mn-doped oxide nuclear fuel
A nuclear fuel includes uranium(IV) oxide (UO2) and manganese (Mn) as a dopant. The Mn dopant may be present in the fuel in an amount up to the solubility limit for Mn under a given set of conditions, for example, about 0.01 wt % to about 1 wt %. The nuclear fuel is substantially free of aluminum (Al). The nuclear fuel exhibits enhanced grain size development during sintering temperatures as low at 1400 K due to an increase in uranium sub-lattice vacancies induced by dissolution of the Mn dopant at interstitial defect sites. The Mn-doped nuclear fuel exhibits improved grain sizes at lower temperatures compared to Cr-, Al-, and undoped UO2, and therefore desirably exhibits lower fission gas release and higher plasticity, reducing the chances of fuel rod failure.
The effect of cerium, neodymium, and ytterbium doping on UO 2 dissolution
Here, the dissolution rate of spent nuclear fuel has been studied heavily to understand the impacts of a failed waste package scenario on potential radionuclide release from a geologic repository. Multiple countries have evaluated these scenarios in oxidizing and reducing environments respective to their relevant repository conditions. The repository environment (e.g., water chemistry, temperature, oxygen content) and the fuel itself (e.g., chemical content of the fuel) both heavily impact the dissolution rate of the spent fuel. This work examined the impact that rare earth element dopants (Ce, Nd, Yb) have on the fuel dissolution under repository relevant conditions with decreasing oxidizing conditions using a single pass flowthrough system. UO 2 samples were doped with Ce, Nd, or Yb with concentrations between 1 and 5 at%. The addition of dopants to the samples reduced the dissolution rates on most samples relative to pure UO 2 samples. Scoping experiments that occurred in a less oxidizing environment showed a reduction in dissolution rate compared to fully oxidizing conditions. The results within highlight the importance of dopant behavior in used fuel dissolution modeling.