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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.

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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Data transfers for full core heterogeneous reactor high- fidelity multiphysics studies

Multiphysics simulations for nuclear reactor analysis are usually performed by resorting to operator splitting and fixed point iterations between single-physics solvers. This enables the separate solution of each physics, such as neutronics, fuel performance, and thermal hydraulics, on meshes tailored to the requirements of the respective numerical discretizations of the equations. As the equations are coupled, several fields must be transferred between single-physics solves. Projecting fields between meshes while preserving order of accuracy, conservation properties, and mapping non-overlapping geometries is a complex endeavor. This conference paper will present the transfers as implemented in MOOSE, which can handle arbitrary meshes, arbitrary mappings, conservation of integral quantities, and are made to scale with distributed simulations on both ends of the transfers. Their adequacy for advanced nuclear reactor multiphysics coupling is shown through examples and numerical studies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Bison Analyses of Fuel Forms used in Small Modular Reactors

Small Modular Reactors (SMRs) are actively being pursued by the international nuclear industry. The Bison fuel performance code developed at Idaho National Laboratory is a tool capable of analyzing the thermo-mechanical response and species diffusion for a wide variety of fuel types under both normal and accident conditions. Two of the fuel types considered for a wide range of SMRs are TRISO fuel particles and metallic fuel. This paper provides an overview of the capabilities in Bison for these fuel types with a few example studies relevant for the SMR industry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Recent High Burnup LOCA Testing at Oak Ridge National Laboratory

To improve fuel cycle economics, increasing the fuel burnup limit in light-water reactors requires a solid technical foundation. Observations from experiments at Halden and Studsvik highlighted severe fuel fragmentation during loss-of-coolant-accident (LOCA) conditions, indicating a need for further technical considerations. These experiments suggest that the fragmentation threshold for high-burnup fuel might be influenced by pre-transient power levels. Consequently, additional LOCA test data are essential to complement existing findings and to deepen our understanding. Oak Ridge National Laboratory’s Severe Accident Test Station has been instrumental in advancing knowledge about high-burnup fuel fragmentation, relocation, and dispersal. This milestone report details two high-burnup LOCA tests designed to evaluate the effects of terminal temperature and grid spacers (or cladding restraints) on fuel fragmentation and relocation susceptibility. In addition, BISON fuel performance modeling and out-of-cell benchmark testing were conducted to better interpret the in-cell test results. These tests were developed in collaboration with fuel vendors to ensure that the results provide valuable data for topical reports and support the Nuclear Regulatory Commission’s review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A High-Fidelity Model of the Peach Bottom 2 Turbine-Trip Benchmark Using VERA

This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type.

BWR↗
DOE OSTI2024

Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles

Tristructural isotropic (TRISO) particles are under consideration for use in several proposed advanced nuclear reactor concepts. The silicon carbide (SiC) layer in TRISO acts as a barrier to prevent the release of the fission products. However, despite remarkable retention, silver (Ag) release has been observed from intact particles, which requires investigation since the Ag isotope ( 110m Ag) has a long half-life. Previous work focused on developing a multiscale, mechanistic model for Ag diffusion accounting for temperature and microstructure effect and has been successfully validated. In this work, we expand the previous model to account for irradiation-enhanced Ag diffusivity in SiC and improve its accuracy over a wider grain size and temperature ranges relevant for advanced reactor conditions. A temperature, grain size, and flux dependent diffusivity is therefore derived using the mesoscale code MARMOT and implemented in the fuel performance code BISON. The irradiation-enhanced Ag diffusivity in SiC is compared against experimental data and validated using BISON against Ag release measurements from the Advanced Gas Reactor Fuel Development and Qualification Program (AGR-1 and AGR-2). Herein, we quantify the impact of SiC grain size, irradiation, and temperature on Ag release. In agreement with previous studies, we find accounting for SiC grain size improves agreement between BISON predictions and experimental observations for most cases. In conclusion, we also find that accounting for irradiation improves agreement for cases where Ag release was underestimated, but the impact was less significant than accounting for microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Engineering of a Coupled Nanocomposite as a High-Performance Protonic Ceramic Fuel Cell Cathode

The lack of high-performance cathode catalysts is a salient issue that bedeviled the commercialization of protonic ceramic fuel cells (PCFCs). Here, in this work, we report a remarkable electrocatalytic activity and stability enhancement of cathode electrodes by engineering a coupled nanocomposite. The as-prepared Pr 0.3 (Ba 0.5 Sr 0.5 ) 0.7 Co 0.8 Fe 0.2 O 3−δ nanocomposite possesses a bulk cubic phase on which homogeneous and intimate orthorhombic PrCo 0.5 Fe 0.5 O 3−δ nanoparticles are uniformly decorated. X-ray diffraction and Raman spectroscopy reveal the excellent thermal stability of the nanocomposite. It achieves a high peak power density of 1.02 W cm –2 based on protonic electrolytes at 600 °C. No noticeable structural degradation is observed over ∼210 h at 550 °C according to scanning electron microscopy analysis. This work demonstrates an effective strategy to boost the performance of perovskite oxides for PCFCs via nanocomposite engineering. It may apply to other catalyst designs and discoveries, such as for batteries, electrolyzers, and membrane reactors.

08 - HYDROGEN↗
DOE OSTI2024

Mini-99 Experiment Design and Target Assembly

Qualification of nuclear fuels requires an understanding of a myriad of fuel performance variables, which requires time-consuming irradiations in test reactors. Accelerating burnup accumulation to reduce the irradiation time has historically been accomplished by increasing the 235 U enrichment in integral fuel tests, which inherently couples the fuel’s temperature and fission heat generation. Separate effects irradiation testing of nuclear fuel, using the MiniFuel target in the Oak Ridge National Laboratory’s (ORNL’s) High Flux Isotope Reactor (HFIR), offers an approach for decoupling fuel temperature and fission rate by reducing the quantity of the fuel and relying primarily on gamma heating in the surrounding components to drive fuel temperatures. However, if the fission rate is high enough, the fuel temperature will still vary over time as the fission heating changes, primarily due to poor heat conduction between the fuel and the surrounding components. This work details the design, analysis, and fabrication of a MiniFuel target which utilizes sodium bonding to improve the heat rejection from the fuel specimen, enabling the testing of higher 235 U enrichments (~8wt.%) to further accelerate burnup accumulation without prohibitively large temperature variations throughout the course of the irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

BISON Simulated and Experimental Fission Product Release Comparisons from Reradiated AGR-3/4 Compacts During High Temperature Heating Tests

The fuel performance modeling code BISON was used to predict the release of fission products iodine-131 (131I), xenon-133 (133Xe), and krypton-85 (85Kr) from four re-irradiated AGR-3/4 fuel compacts containing tristructural isotropic (TRISO) coated particles during high-temperature isothermal heating tests. The AGR-3/4 fuel compacts were irradiated in the Advanced Test Reactor (ATR) as part of the third and fourth series of planned experiments to support the Advanced Gas Reactor (AGR) Program. They were subsequently stored and re-irradiated in the Neutron Radiography (NRAD) reactor for approximately five days and then stored for another five to eight days before being subjected to isothermal heating tests in the Fuel Accident Condition Simulation (FACS furnace) for 200 to 300 hours at temperatures between 1000°C and 1600°C to evaluate fission product release at elevated temperatures. New nuclide-specific fission product source term models for the three nuclides of interest were developed using the reactor multiphysics code Griffin and implemented into BISON to support this work. The new source term models were incorporated into coupled compact- and particle-scale BISON simulations, which predict spatially- and temporally-resolved radionuclide generation, radioactive decay, transport, and release throughout the entire irradiation history, including the initial ATR irradiation, NRAD re-irradiations, FACS heating tests, and intermediate periods spent in storage. The experimentally measured fission product release from the heating tests were compared to modeling release predictions calculated by BISON to evaluate how well the code compares to experimental results. Overall, the experimental measured and BISON predicted comparative release results varied but generally agreed to within 5 particle equivalents. Comparative release results identified general observations to take into consideration to help refine future models and reduce uncertainties associated with both the measurement results and predictive results. This includes developing new uranium oxycarbide (UCO) specific kernel diffusivities for the three isotopes examined to more accurately reflect the material properties of the fuel form. Deriving new diffusivities will aid in producing a more informed BISON model

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Sustainable Aviation Fuel Blending and Logistics

Worldwide, aviation accounts for 2% of all manmade carbon dioxide emissions and 12% of all transportation CO2 emissions. In 2023, the U.S. accounted for 27% of the world jet fuel consumption. The aviation industry has set sustainability goals, and mandates. Sustainable aviation fuel (SAF), made from nonpetroleum feedstocks, significantly reduces aviation emissions. SAF must be blended with petroleum-based jet fuel prior to its use in aircraft. Jet fuel quality standards and certification documents are essential to fuel performance, operability, and safety and are the primary driver in determining locations for blending. Several locations were evaluated including terminals, airports, refineries, and greenfield/brownfield sites. While all the locations evaluated are technically capable of blending fuel, there are practical considerations that make terminals the optimal location for blending. The report includes background on jet fuel and SAF use, fuel quality standards, modes of transport for both conventional jet fuel and SAF, and terminal information.

09 BIOMASS FUELS↗
DOE OSTI2026

AGR-5/6/7 Ceramography Report

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established in 2002 to conduct research and development on tri-structural isotropic (TRISO)-coated particle fuel for High Temperature Gas-cooled Reactors (HTGRs). All AGR irradiations have been completed: AGR-1 (Collin 2015), AGR-2 (Collin 2014), AGR-3/4 (Collin 2016) and AGR-5/6/7 (Pham 2021). The objectives of each program are presented in Figure 1.1. The purpose of the AGR-5/6/7 program was to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials, safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product (FP) transport (Sharp 2020). The AGR-5/6/7 work continues as a final irradiation program of the Advanced Reactor Technologies (ART) program, that began in February 2018 and ended in July 2020. The AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 effective full power days (EFPDs) (Stempien 2023), resulting in final burn-up values, on a per-compact basis, ranging from 5.66% to 15.26% fissions per initial heavy metal atom (FIMA), and fast fluence values ranging from 1.62×1025 n/m2 to 5.55×1025 n/m2 (E >0.18 MeV) (Stempien 2022)

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A Refined BISON-Griffin Workflow For Designing Targeted TRISO Transient Experiments In The Transient Reactor Test Facility

U.S. industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In our previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs,(2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. Predicted tangential stresses in particle silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. Use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Implementing Fuel Cladding Bonding and Assessing its Impact on Axial Gas Communication

Fuel rods irradiated in light-water reactors to burnup values above 45 MWd/kgU are subject to the formation of a chemical and mechanical bond between the fuel and cladding upon gap closure. The formation of the bond subsequently inhibits the ability of fission gases released from the fuel to flow freely to the plenum of the rod. The flowing of gases within fuel rods is referred to as axial gas communication. During transients, such as loss of coolant accidents, the bond may influence cladding deformation prior to breaking. Upon bond breakage, gases are able to more freely communicate to the lower pressure regions of the rod. The impact of bonding on gas communication and the ballooning behavior of the cladding during a loss of coolant accident is of interest to the nuclear industry in support of burnup extension for the existing light-water reactor fleet. In this report, a model to capture the effects of fuel cladding bonding in the BISON fuel performance code is presented. The theory of the model along with implementation testing is provided. A summary of a previously developed axial gas communication is given to set the stage for how the two models may be coupled together. A full-length pressurized-water reactor fuel rod demonstration is highlighted to evaluate the impact of including bonding on axial gas communication calculations using a preliminary coupling methodology. An overview of the next steps regarding the modeling of bonding, axial gas communication, and a more tightly coupled framework is also provided.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Radiation induced athermal diffusivity in uranium mononitride

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuel considered for light water reactors and advanced reactor designs. Properties like self- and fission gas diffusivity need to be better understood, given that they influence key fuel performance phenomena such as fission gas swelling and release. In particular, the radiation induced athermal (D 3 ) diffusivity remains challenging to accurately predict and has only been sparsely characterized in UN, despite its importance as it likely governs diffusion at the low temperatures this high-thermal-conductivity fuel form may operate. Molecular Dynamics simulations are used to estimate the mean square displacement induced by a primary knock-on atom (PKA) with a given kinetic energy. These results are combined with the PKA energy distributions obtained from binary collision approximation calculations to obtain the displacement due to a particular fission fragment. Finally, this is combined with experimental fission fragment yields to determine the displacement due to an average fission event and, thus, express the athermal diffusivity as a function of the fission rate density. These results are in excellent agreement with available experimental data. In conclusion, a particular importance is given to the understanding and the quantification of the variability of these results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Boiling Water Capsule Irradiation Test Device Design Study Report

RELAP5 calculations were conducted to support an experimental design, referred to as water capsule, which can facilitate fuel performance testing within ATR. The main objective is to provide useful insights into the experimental design and operating strategies that can mimic LWR operating conditions within the spatial constraints of the ATR position under consideration. This report summarizes the RELAP5 analysis results based on the most recent water capsule design derived through various tests. To discuss a design under conservative conditions, fuel power is consistently assumed to be 400 W/m in all analyses.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗
DOE OSTI2025

Role of Accelerated Burnup Irradiation Testing in Support of Accelerated Fuel Qualification

Accelerated fuel qualification has gained attention as a means to reduce the time needed to realize new nuclear fuel concepts and expand the operating windows of existing fuel forms. A key component of this approach is accelerated burnup irradiation testing. Although the concept of accelerated burnup has been familiar to the community for many decades, the specifics about how the increasing fission rate may be used as a qualification tool have not yet been elucidated. The present work provides a vision of how accelerated fission rate testing can enable accelerated fuel qualification. Technology readiness levels (TRLs) are reintroduced to demarcate the stages of traditional fuel qualification, and accelerated fuel qualification is presented in this context. The critical steps needed to achieve each TRL are reframed within the context of modern nuclear materials research and development, as revolutionary fuel concepts are more common than previous eras. The practical impacts of accelerated fuel qualification approaches as applied to contemporary fuel qualification efforts are illustrated. Examples are given to illustrate how accelerated burnup irradiations are being used currently and could be applied in the future to support qualification and licensure. Finally, outstanding challenges in the application of accelerated burnup methods to nuclear fuel qualification are summarized, with priority placed on understanding how fission rate impacts diffusion, microstructure evolution, and other critical mechanisms that dictate fuel performance.

Accelerated fuel qualification↗
DOE OSTI2025

In-situ ion irradiation of a spent UO2 fuel: evolution of fission products and nanograins with radiation dose

This study investigates the behavior of fission gas bubbles, five metal precipitates (5MPs) (Mo, Ru, Rh, Tc, Pd) and nanograins in spent uranium dioxide (UO2) fuel under various ion irradiation doses, temperatures, and flux conditions. Utilizing in-situ ion irradiation and advanced transmission electron microscopy, we analyzed the evolution of fission gas bubbles, 5MPs and nanograins in UO2 samples from the Belgium Reactor 3 (BR-3). Our findings reveal significant shrinkage of fission gas bubbles and 5MPs with increasing irradiation dose, accompanied by a decrease in pair density. The study also reveals nanograin coarsening under irradiation. We demonstrate that ion irradiation induces a homogeneous re-solution process where individual atoms are ejected from bubbles and precipitates, leading to their dissolution and subsequent re-precipitation in the matrix. This study provides critical insights into the dynamic behavior of fission products and nanograins under irradiation, facilitating the development of predictive models and contributing to the optimization of nuclear fuel performance and safety.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A realistic TRACE PWR LOCA model with application to high-burnup analysis

This paper presents a TRACE model of a postulated large-break loss-of-coolant accident (LBLOCA) event in a four-loop pressurized water reactor (PWR). The input files have been made publicly available and serve as a starting point for researchers to improve upon or apply to their purposes. The model, based on a legacy PWR model, contains numerous improvements and modifications consistent with US Nuclear Regulatory Commission LBLOCA analysis guidelines. The model is applied to analyze high-burnup (>62 GWD/MTU) fuel behavior during LBLOCA, improving on previous work to provide realistic thermal hydraulic predictions for future fuel performance analyses and experiments related to fission fragment, relocation, and dispersal (FFRD). Sensitivity studies are presented to quantify the impact of the vessel-modeling approach, core radial discretization, and other phenomena. In conclusion, the model is intended for research purposes and contains sufficient plant geometric, operational, and safety system details to represent the main physical phenomena pertaining to LBLOCA.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗