Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “UO”

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.

At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Evaluation of BISON's Transient Fission Gas Release Model on High-Burnup UO 2 MiniFuel

BISON's fission gas release (FGR) model has been evaluated and tested over several hypothetical thermal transient cases at high burnup and validated against FGR experimental data from an annealing test at high burnup. Diffusion-controlled FGR is accounted for in the model by including several physical mechanisms such as the diffusion of fission gas to grain faces, grain boundary sweeping, grain growth, and intergranular bubble growth. Under temperature variations, FGR is increased due to the development of micro-cracks at grain faces. This existing transient capability, however, was developed based on transient fuel behavior during power ramps and the associated micro-cracking process. Nevertheless, a fraction of the FGR during temperature transients such as LOCA can ensue from different mechanisms (in particular, fuel fragmentation) and present an important dependency on local burnup and high burnup structure formation. It follows that the performance and necessary developments of the existing BISON fission gas model for the analysis of LOCA-type transients need to be investigated through a systematic assessment study. Such a study is performed in this work. The model shows an overall good qualitative agreement with experimental observations on FGR during annealing tests with different temperature conditions. It also accounts well for microstructural effects on FGR. When quantitatively compared to FGR data from previously irradiated 103 MWd/kgU UO 2 discs under thermal annealing, the model shows a less satisfactory agreement with the experimental data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Updates on UO 2 -BeO Experiment (IER 523) [Slides]

This presentation titled "Updates on UO 2 -BeO Experiment (IER 523)" covers experiment status, experiment motivation, CED-1 summary, current efforts (CED-2), and includes a concluding summary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

UO 2 -Liquid Metal Suspension Fuel Concept: Alloy Selection and Depletion Calculations

A novel nuclear fuel was recently proposed that consists of UO 2 particles suspended in a liquid metal mixture. Two liquid metal candidates were identified: BiPbSn and PbSn. Scoping simulations were performed on an over-simplified reactor core that showed a potential limit on achievable burnup. Those findings led to two questions that needed to be addressed: What is the corrosion behavior of the liquid metals toward Zircaloy-4 (Zr4)? And what level of burnup can the fuel be expected to achieve? This report documents efforts to answer both of those questions. First, corrosion experiments were performed between the liquid metal candidates and Zr4 coupons. Second, the previous scoping simulations were expanded to include depletion analyses so that maximum burnups could be estimated. At 400°C, neither PbSn nor BiPbSn caused significant corrosion to the Zr4 coupons. At 600°C, BiPbSn did not cause significant corrosion, but PbSn did. In addition, pure bismuth and pure tin were both highly corrosive, whereas pure lead was non-corrosive. The depletion simulations found that low-enriched fuel could maintain light-water reactor operating temperatures but achieve a burnup of only about one-third of existing reactors. Higher enrichments could achieve burnups equal to or higher than those of existing reactors but at the expense of higher operating temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Deployment of BISON models of fuel restructuring at high burnup and related fission gas behavior in UO 2

This milestone report details the advancements made in fiscal year 2024 under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to improve the modeling of fission gas behavior in high burnup UO 2 nuclear fuel in the BISON fuel performance code. As nuclear fuel is pushed to higher burnups, significant microstructural changes occur within the fuel, including the formation of a high burnup structure (HBS) on the pellet rim and a dark zone deeper within the pellet. These regions, characterized by subgrain formation and increased pore densities, have critical implications for fission gas behavior and release, which are not well understood. The modeling capabilities in BISON did not adequately predict these phenomena, leading to an underestimation of fuel restructuring and - potentially - of fission gas release. To address these gaps, this milestone focused on three key objectives: (1) reviewing and assessing Sifgrs's capabilities for low burnup fuel, on which high burnup capabilities rely, (2) validating and expanding HBS fission gas modeling capabilities, including investigating mechanisms for fission gas release from HBS, and (3) expanding Sifgrs to enable modeling of dark zone formation and its effects on fission gas behavior. These objectives were achieved and are described herein. The achievements of this NEAMS milestone are significant for the industry's goal of burnup extension. The improved predictive modeling capabilities for both low- and high-burnup conditions enhance our understanding of fuel performance under both normal operations and transient scenarios. Although goals were reached, future work is necessary to validate these models against experimental data and quantify their accuracy in different conditions. In parallel, mechanistic modeling efforts should continue to extend and refine these capabilities to increase accuracy while reducing reliance on empirical models. This will ensure robust performance across a broader range of conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A New Reduced Order Model For The Mechanistic Creep Behavior Of UO 2

This manuscript describes an ongoing NEAMS effort to better determine the performance of advanced nuclear fuels, in particular the creep behavior of doped UO$_2$ for light water reactors. In our previous work, we outlined a method to utilize data generated from lower length scale simulations and implement it into the engineering scale fuel performance analysis. This process has been further refined, and in addition, new data has been used to train the surrogate model which has also been substantially improved since the previous iteration. The new model is compared against the current empirical model used in BISON using both scoping calculations to define the performance over the parameter space and using integral instrumented fuel assessment cases to determine the impact of these models on the overall fuel performance. Suggestions and guidance for future improvements to this method are provided to ensure the model covers relevant parameter space and phenomena.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mesoscale Modeling for Restructuring and Fragmentation in High Burnup UO 2

This report summarizes the mesoscale modeling conducted in fiscal year 2025 under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, focusing on the microstructural evolution and restructuring in high burnup UO 2 nuclear fuel and its impact on fuel fragmentation. We developed a pioneering phase-field model to simulate restructuring behavior across different regions of high burnup fuel, including the dark zone and rim region. A grand-potential-based phase-field model is employed 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 temperature and burnup rate were studied to capture how each of these parameters influences the characteristics of the restructured fuel. Subgrain formation was observed to initiate around existing fission gas bubbles and proceed toward triple junctions, grain boundaries, and grain interiors. Under a given subgrain formation rate, the rate of restructuring increases with rising fuel temperature. The restructuring occurs faster with higher burnup rate. 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. The predicted microstructures are consistent with experimental observations of the restructured regions. Based on the mesoscale simulations, a mechanistic model for restructuring and grain size evolution was implemented in BISON. Thus, this work provides a first-of-its-kind restructuring model for different regions of high-burnup fuel to BISON. It was shown that the model predicts appropriate grain size evolution along fuel radius as those observed in experiments. This model enables BISON to predict the effect of restructuring on the fission gas release. Finally, the phase-field fracture simulation with dark zone specific microstructures were presented to provide the fragmentation criteria for the dark zone. This work introduces a first-of-its-kind restructuring model for high burnup fuel, significantly enhancing BISON's predictive capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Transition Metal Taggants in UO 2 from First Principles

The incorporation of transition metals into nuclear fuel has gained attention both to improve fuel properties and as a possible nuclear forensics tool. Recent experimental studies by Ulrich et al. and Adorno Lopes et al. have investigated Ni and Fe as candidate transition metal dopants for potential nuclear forensics purposes and found that there is minimal alteration to key UO 2 fuel properties. In the present work, we performed density functional theory (DFT) investigations into possible defect structures for Ni and Fe incorporation. The dynamic stability of these defect structures was validated by calculating the phonon density of states. We found that transition metal incorporation likely occurs via substitution at U sites in the fluorite crystal structure with a nearby O vacancy for charge-balancing, which agrees with experimentally proposed structures. Additionally, this defect structure does not cause long-range alterations to the crystal parameters—an important consideration for use as a nuclear fuel taggant. Future needed work involves computational investigation of additional defect concentrations using larger supercells, additional transition metal charge states, and thermal effects. Such investigations can be carried forward into sintering models for a more complete understanding of the suitability of Ni and Fe as fuel taggants.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of grain size and porosity on cladding failure in high-burnup UO 2 : A sensitivity and uncertainty study

Isotopic taggants are being studied to aid in the provenance assessment of nuclear materials. However, these taggants must be selected such that they do not adversely affect fuel performance during normal operation or accident scenarios. Taggants are known to affect the fuel’s grain size and porosity. In the work described in this paper, the BISON fuel performance code was used to assess the potential effects of taggants (i.e., grain size and porosity) on fuel rod behavior and cladding failure during a high-burnup, large-break loss-of-coolant accident. Here, 281 individual fuel rods from the same reactor core were modeled for a sensitivity study, a parametric study, and uncertainty quantification. The cladding failure predictions often exhibited stochastic behavior. After additional study, it was found that the cladding failure model is highly sensitive to residual error inherent to numerical approximation solvers. Some strategies to mitigate this sensitivity are discussed. The study found no relationship between known taggant effects and cladding failure status. However, taggants were found to affect the time and location of failure in certain rods. In conclusion, future work to continue investigating and validating these findings is briefly discussed.

Doped UO 2↗
DOE OSTI2020

Materials Data on UO by Materials Project

UO1 is Halite, Rock Salt structured and crystallizes in the cubic Fm-3m space group. The structure is three-dimensional. U is bonded to six equivalent O atoms to form a mixture of corner and edge-sharing UO6 octahedra. The corner-sharing octahedral tilt angles are 0°. All U–O bond lengths are 2.43 Å. O is bonded to six equivalent U atoms to form a mixture of corner and edge-sharing OU6 octahedra. The corner-sharing octahedral tilt angles are 0°.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A phase field study of the thermal migration of gas bubbles in UO 2 nuclear fuel under temperature gradient

Phase field models are developed to study the gas bubble migration in uranium dioxide nuclear fuel in which a large temperature gradient exists during the operation. In this work, thermal diffusion mechanism for nanosized gas bubbles and vapor transport process for micron-sized gas bubbles are considered, respectively. In both cases, gas bubbles migrate to the high-temperature area. Due to the velocity difference between leading and trailing edges of the gas bubbles, nanosized gas bubbles are elongated along the temperature gradient direction when thermal diffusion is dominated. Micron-sized gas bubbles are either compressed along temperature gradient direction to form lenticular shape bubbles or elongated along temperature gradient direction, depending on the location of the gas bubbles within the fuel pellet. Initial gas bubble radius has no significant effect on the gas bubble migration velocity for both thermal diffusion and vapor transport mechanisms. We notice that the shape change of the gas bubble due to vapor transport mechanism has no significant effect on the migration velocity. Furthermore, the center cavity formation is also captured by our model which is due to the migration and accumulation of lenticular gas bubbles at the center of the fuel pellet. The modeling results compare well with experimental observations and theoretical analysis in the literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal-shock experiments for separate-effects validation of UO 2 fuel fracture models

Because of the important role that fracture plays in the behavior of ceramic UO2 fuel in a nuclear reactor environment, fracture models are a major component of fuel performance codes. As with any aspect of fuel performance, it is crucial to validate these fracture models against experimental data; however, obtaining well-controlled data for conditions representative of a reactor environment is difficult. Quenching is proposed here as a relatively simple approach for using a laboratory environment to achieve conditions that approximate those of a reactor environment. In this paper, an experimental apparatus containing a single instrumented fuel pellet in a sealed section of copper tubing is developed. It is then applied to a series of seven experiments in which the apparatus is first heated to a high temperature (580-680°C) by immersion in a molten salt bath, then quenched in a cold bath (-10-4 °C). Development of these experiments was guided by numerical simulations, and post-test simulations were performed to aid in understanding the experimental behavior and assessing the accuracy of the predictions of fracture initiation and propagation. In addition, similar experiments were performed on solid copper rods to provide temperature-dependent heat transfer coefficients for use in simulations of these experiments. Here, this study demonstrates that quenching is a viable approach for generating thermal gradients representative of those in prototypical light-water reactor conditions at powers of about 5–10 kW/m. Fracture is expected to begin at these power levels, and moderate amounts of radial and axial cracking was observed in these quenching tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Short Communication: Observation of Initial Burst Release of Fission Gas from High-Burnup UO 2 Nuclear Fuel During Thermal Transient

A system was developed and tested to provide a deeper understanding of the fission gas release kinetics from nuclear fuel during thermal transients. Pressure, temperature, spectral data, and optical images were simultaneously collected during resistive sample heating, and all released gases were collected in liquid nitrogen-cooled charcoal traps. Standup testing was performed with a high-burnup nuclear fuel segment from rod section with an average local burnup of 77 GWd/tU. During heating, the segment released approximately 6 ± 2% of generated fission gas inventory after ramp heating to 460°C. Here, heating ceased when the sample ejected from the holder, as observed by constant imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Molybdenum and Tungsten isotope compositions of UO 2 fuel pellets: Implications for isotopically enriched taggants

The addition of isotopically enriched taggants to material at the front end of the nuclear fuel cycle could be a powerful tool used to assist law enforcement authorities should material outside of regulatory control be found. Two potential candidates for this purpose are molybdenum (Mo) and tungsten (W) as both elements have five or more stable isotopes and are trace elements contained within nuclear fuel. So there is a concern that Mo and W could undergo isotope fractionation during processes like uranium enrichment and to date, it is unknown if nuclear fuels have natural Mo and W isotope compositions. If Mo and W isotopic variability is present in nuclear fuels, this would hinder the use of these elements as isotopic taggants because it would be difficult to discern the original taggant isotope composition with high confidence. Therefore, a set of 16 low enriched uranium (LEU) fuel pellets from US commercial producers was analyzed using multi collector-inductively coupled plasma mass spectrometry (MC-ICPMS) to determine Mo and W isotope compositions (i.e., 94 Mo/ 92 Mo, 95 Mo/ 92 Mo, 96 Mo/ 92 Mo, 97 Mo/ 92 Mo, 98 Mo/ 92 Mo, 183 W/ 182 W, 184 W/ 182 W, and 186 W/ 182 W). Relative to terrestrial standards, LEU fuel pellets have variable Mo and W isotope compositions, thereby complicating the use of these elements as isotopically enriched taggants. As such, this work demonstrates that the isotope composition of any potential taggant must be well characterized in the base nuclear fuel prior to any taggant addition. Furthermore, these results suggest that Mo and W are not ideal candidates for isotopically enriched taggants.

58 GEOSCIENCES↗
DOE OSTI2024

Mesoscale Modeling of the Effects of Accelerated Burnup on UO2 Microstructural Evolution

Accelerating the nuclear fuel qualification process will rely on some combination of advanced modeling and simulation techniques with accelerated irradiation testing and separate effects experiments to enable the development of new fuel concepts in a shorter time frame. One of the key challenges to successfully leveraging accelerated irradiation tests will be understanding the artifacts that may be introduced with accelerated accumulation of dose and/or burnup. This work presents phase field (MARMOT) simulations of the evolution of representative 2D UO2 microstructures up to 40 MWd/kgU. Simulations were performed under both commercial light water reactor fuel conditions as well as those that would be expected for highly accelerated (~10x) burnup conditions similar to those used in the MiniFuel irradiations in Oak Ridge National Laboratory’s High Flux Isotope Reactor. The phase field model was coupled with a discrete nucleation algorithm to model re- structuring at high burnup. The effect of the different fission rates in both microstructures was investigated at two temperatures: 650?C and 800?C. The lower temperature simulations both showed an onset of restructuring at nearly 60 MWd/kgU. More extensive restructuring was obtained in the MiniFuel microstructure compared with that of the PWR fuel. At 800?C, no restructuring was obtained as a result of the thermally activated diffusion of Xe atoms and U vacancies to fission gas bubbles, which reduces the nucleation driving force. These results highlight the importance of using modeling and simulation tools to inform the environmental conditions during targeted accelerated irradiation tests to extract the most useful fuel performance data.

accelerated fuel qualification, Phase Field, Restr↗