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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO 2 ), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Thermo-Mechanical Phase-Field Modeling of Fracture in High-Burnup UO2 Fuels Under Transient Conditions

This study presents a novel multiphysics phase-field fracture model to analyze high-burnup uranium dioxide (UO2) fuel behavior under transient reactor conditions. Fracture is treated as a stochastic phase transition, which inherently accounts for the random microstructural effects that lead to variations in the value of fracture strength. Moreover, the model takes into consideration the effects of temperature and burnup on thermal conductivity. Therefore, the model is able to predict crack initiation, propagation, and complex morphologies in response to thermal gradients and stress distributions. Several simulations were conducted to investigate the effects of operational and transient conditions on fracture behavior and the resulting cracking patterns. High-burnup fuels exhibit reduced thermal conductivity, elevating temperature gradients and resulting in extensive radial and circumferential cracks. Transient heating rates and temperatures significantly affect fracture patterns, with higher heating rates generating steeper gradients and more irregular crack trajectories. This approach provides critical insights into fuel integrity during accident scenarios and supports the safety evaluation of extended burnup limits.

Chemistry↗
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

Thermal conductivity suppression in uranium-doped thorium dioxide due to phonon-spin interactions

In this work, impact of low level of uranium (U) atom substitution on thermal conductivity of thorium dioxide (ThO 2 ) is investigated. ThO 2 is an electronic insulator with a wide optical band-gap and no unpaired electrons whose thermal transport is governed by phonons. U-substitution introduces unpaired f-electrons resulting in paramagnetic behavior of U-ThO 2 at room temperature, which significantly suppresses its thermal conductivity. A single crystal of U-ThO 2 with graded composition of U is grown using a hydrothermal synthesis method, and thermal conductivity measurements are performed in regions with uniform composition of U at levels of 0%, 6%, 9% and 16%. Measured thermal conductivity profiles over 77–300 K temperature range are analyzed using an analytical expression for phonon-mediated thermal transport based on Klemens-Callaway model. Temperature dependent thermal conductivity is found to deviate significantly from the Rayleigh scattering trend expected for a simple substitutional point defect with a small perturbation to mass and interatomic forces. With the resonant scattering term, observed large suppression of thermal conductivity at low temperatures can be closely reproduced. Additionally, the extracted phonon-spin coupling constants imply a nonlinear relation of phonon-spin interaction intensity with respect to U doping percentage. Our study reveals how phonon-spin scattering contributed by unpaired f-electrons in U atoms influences thermal transport in the U-ThO 2 system.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Impact of dynamic Jahn-Teller effect on magnetic excitations, lattice vibration, and thermal conductivity in U 𝑥 ⁢T⁢h 1−𝑥 ⁡O 2 system

Vibrational and magnetic properties of single-crystal uranium-thorium dioxide (U 𝑥 ⁢T⁢h 1−𝑥⁡ O 2 ) with a full range of 0 < 𝑥 < 1 are investigated. Thorium dioxide is a diamagnet whose thermal properties are governed by lattice vibration. The addition of paramagnetic uranium ions leads to the emergence of magnetic effects that alter the thermophysical properties noticeably even at room temperature. The interaction of phonons with magnetic moments of uranium 5⁢𝑓 electrons mediated by magnetoelastic coupling results in an anomalous low-temperature thermal conductivity profile. Analysis of the magnetic susceptibility measurements indicates a departure from the Curie-Weiss relationship characteristic of noninteracting paramagnetic ions, previously associated with the dynamic Jahn-Teller (DJT) effect characterized by coupling between spin and the oxygen sublattice. The T 2⁢g Raman peak position follows a nonlinear trend as a function of uranium concentration and hints that these Raman active optical modes play a role in either DJT or in mediating quadrupole-quadrupole interactions. A first-principle-based thermal transport model is implemented to explain the low-temperature transport measurements, where the anomalous reduction is attributed to phonon-spin resonant scattering. The interplay between spins and phonons is also captured using high-resolution inelastic x-ray scattering (IXS) measurements of phonon linewidths. Our results provide insights into the phonon interactions with the magnetic excitations governing DJT effect and impacting the low-temperature thermal transport processes in this material system. Furthermore, these findings have implications for understanding low-temperature thermal transport and magnetic properties in advanced materials for information processing and energy applications.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Investigation of UO 2 doped with Fe 2 O 3 sintered under a reducing atmosphere

Nuclear fuel produced with trace amounts of transition metal additives is of potential interest for introducing intentional signatures for accelerating nuclear forensics. In this work, we investigate the effects on microstructure, grain size, crystal structure, and stoichiometry when trace amounts of Fe in the form of Fe 2 O 3 are added to UO 2 . Sintering of compacts with different concentrations, from 250 to 3000 ppmw, was performed under a reducing atmosphere at 1773 K. The persistence of the taggant during the fuel fabrication process as well as its impact on grain size, crystal lattice, and first-neighbor chemistry was evaluated using inductively coupled plasma–optical emission spectroscopy, powder X-ray diffraction (pXRD), Raman spectroscopy, scanning electron microscopy–backscatter electron spectroscopy, energy dispersive spectroscopy, and thermogravimetric analysis. We observed that a negligible amount of Fe was lost during sintering. Our results indicate that the feedstock, Fe 2 O 3 transforms into Fe and FeO under the test sintering conditions. For all compositions, metallic Fe precipitate was found in grain boundaries as a secondary phase. In conclusion, the potential incorporation of ionic Fe into the UO 2 unit cell was determined by pXRD and Raman spectra.

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 OSTI2025

Tantalum Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.02 cm Pitch)

The United States (US) Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories (SNL). The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water moderated and — reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM 096, LEU- COMP-THERM-097, LEU-COMP-THERM-101, LEU-COMP-THERM-102, and LEU-COMP-THERM-111. Applied interests in neutron-absorbing elements at epithermal and intermediate neutron energy ranges at DOE’s Hanford Site and Savannah River Site were identified through DOE Office of Environmental Management (EM) workshops and needs surveys performed in conjunction with the National Nuclear Security Administration/Nuclear Criticality Safety Program. In response, a team lead by Oak Ridge National Laboratory (ORNL) performed a study evaluating the use of 7uPCX for testing the epithermal cross sections of materials at SNL. This marked the inception of the set of experiments described here. The purpose of these experiments was to measure the effects of tantalum in nearly-critical systems. The tantalum was introduced into the fuel arrays as experiment rods within a central test region. The central test region was designed to target the epithermal neutron energy range by providing a dry cylindrical cavity constructed of aluminum and lined with 0.04 in (0.1016 cm) thick cadmium sheet. The central test region has an outer diameter of 3.75 in (9.525 cm) length of 31 in (78.74 cm) and can hold 85 tantalum rods nominally 0.25 in (0.635 cm) outside diameter and 31.25 in (79.375 cm) long. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 2016 fuel rod positions on a pitch of 0.4 in (1.016 cm) in a series of 21 hexagonal rings positioned around the central test region. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The eight cases in this experiment series were performed in 2024 in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility by varying the number of fuel rods in the critical assembly. The personnel involved in conducting these experiments included David Ames, Elijah Lutz, Beth Hanson, Patrick Ward, Jason Soares, and Augie Chapa from SNL, as well as Mathieu Dupont and B.J. Marshall from ORNL. Case 1 had no tantalum experiment rods. Case 2 had seven tantalum rods in the center of the central test region. Case 3 had 18 tantalum rods in the third hexagonal ring of the central test region. Case 4 had 19 tantalum rods in the center of the central test region. Case 5 had 30 tantalum rods in the outer most positions of the central test region. Case 6 had 37 tantalum rods in the center of the central test region. Case 7 had 61 tantalum rods in the center of the central test region. Case 8 had 85 tantalum rods filling all positions in the central test region. All eight critical experiments are judged to be acceptable as benchmark experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

In situ irradiation of spent nuclear fuels

To improve the economics of commercial nuclear reactors, nuclear vendors and utilities are seeking approval for increased burnup usage of the existing nuclear fleet. This is especially critical for meeting the clean energy initiative by the United States Government, Department of Energy (DOE). However, one of the key challenges the nuclear industry faces in this regard is that the regions exposed to high burnup and low temperatures exhibit a fine-grained microstructure with large bubbles known as high-burnup structure (HBS) [1]. The formation of HBS has been correlated to the diminished performance of the reactor, as well as fuel fragmentation and pulverization during transient and accidental conditions [2]. Therefore, it is paramount to understand the mechanisms for HBS formation along with its impact on the properties and performance of nuclear fuels. While existing programs, such as Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Advanced Fuel Campaign (AFC) are focusing on evaluating the performance impact of HBS, the physical mechanisms contributing to HBS formation are still not fully understood. In addition, having predictive capabilities and sound understanding of the microstructural evolution of nuclear fuel is essential for accelerated development, qualification, and deployment of new nuclear materials and novel reactor designs for advanced nuclear reactors. There is a lack of consensus among researchers regarding the mechanisms leading to such restructuring observed in HBS. Grain subdivision due to polygonization versus recrystallization, continuous versus discrete recrystallization occurring in tandem or conjunction, etc., have been proposed and debated. In general, it is hypothesized that defect accumulation and dislocation interaction within the grains cause the realignment of dislocations into grain boundaries, leading to the new subgrain formation, which over time transforms into new grains. However, due to the lack of transient data, the importance of fission rate, irradiation, thermal, and stress history of the fuel on the restructuring could not be assessed. In situ microstructural evolution under various irradiation conditions is desired to bridge this gap. Alternatively, phase-field-based models have been developed to capture HBS formation via discrete recrystallization utilizing the classical nucleation approach [3–5]. However, in these models, the grain nucleation criteria are often defined based on empirical relations for burnup and fission gas density leading to dislocation density change. A mechanistic approach to capture the dislocation interaction with the microstructural features leading to grain subdivision is lacking.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Photodissociation and Infrared Spectroscopy of U + (CO 2 ) n , UO + (CO 2 ) n , and UO 2+ (CO 2 ) n Cation-Molecular Complexes

Laser vaporization of uranium in a pulsed supersonic expansion of carbon dioxide is used to produce complexes of the form U + (CO 2 ) n , UO + (CO 2 ) n , and UO 2+ (CO 2 ) n . These ions are selected in a reflectron time-offlight mass spectrometer and studied with visible laser photodissociation and tunable infrared laser photodissociation spectroscopy in the region of the CO 2 antisymmetric stretch. The dissociation patterns and spectroscopy of these ions indicate that CO 2 ligands are intact molecules. Although reaction products that form oxide-carbonyl or oxalate species are predicted to be stable, there is no direct evidence in the frequency range studied for the formation of these species. There is no clear indication for the coordination numbers for singly charged uranium and its oxide complexes with CO 2 . However, there is strong support in the vibrational patterns for an eight-coordinate complex of the doubly charged UO 2+ species, i.e., UO 2+ (CO 2 ) 8 .

Cluster chemistry↗
DOE OSTI2024

Low Pressure Oxygen-Assisted Acidic Dissolution of U Metal Foils

Traditional methods of dissolving uranium in nitric acid involve the generation of nitrogen oxide gases (NO x ), which not only are hazardous but also contribute to increased pressure within the dissolver. When uranium metal foils are dissolved in nitric acid, a series of redox reactions occurs, leading to the formation of NO x , such as nitrogen dioxide (NO 2 ) and nitric oxide (NO).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Uranyl Capture and Activation with Lewis Acids and Macrocyclic Hosts (Final Technical Report)

This document is the final report for the project with the title, “Uranyl Capture with Lewis Acids and Macrocyclic Hosts.” Nuclear power is attractive for meeting the current and future energy needs of society, in that it does not release carbon dioxide or other pollutants into the atmosphere during routine use. This motivated the project, because preparation of nuclear fuel, recovery of useful components from used fuel, and handling of waste materials remain significant impediments to further deployment of important nuclear technologies. In part, these problems arise from limited availability of chemical reactions that can reliably interconvert forms of uranium and other heavy elements during preparation and processing. For example, harsh and expensive chemicals are often required for making and breaking chemical bonds to uranium, and the reactions involved are inefficient. The overall objective of this research was to harness knowledge of chemical structure and bonding to develop a useful and predictive und

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Recommendations for an Applicant to Calculate Activity Data for Greenhouse Gases Estimates

In 2009, the U.S. Nuclear Regulatory Commission (NRC) directed the NRC staff to address climate change issues and consider the impacts of the emissions of carbon dioxide (CO 2 ) and other greenhouse gases (GHGs) in its environmental reviews for major licensing actions (NRC 2009b). To implement this direction from the Commission, the staff issued guidance in 2011 and updated guidance in 2014 in Attachment 1 to Interim Staff Guidance COL/ESP-ISG-026 (NRC 2011; NRC 2014). This guidance provides a simpler method than the method described in RG 4.2 Rev. 3, that an applicant can use to meet the plant parameter envelope (PPE) value from the Generic Environmental Impact Statement for Licensing of New Nuclear Reactors (NR GEIS). NRC staff estimated the 97-year lifecycle GHG emissions from a reference 1000 megawatt electrical (MWe) light-water reactor (LWR) for various activities associated with construction, operation (including uranium fuel cycle), and decommissioning of nuclear power plants and presented the results in Appendix H of the NR GEIS. Appendix H of the NR GEIS includes estimates of direct emissions from construction equipment and emergency diesel engines in a nuclear facility and indirect emissions from workforce vehicular traffic, fuel transportation and the uranium fuel cycle. The NR GEIS Section 3.3 extended the estimates in Appendix H for the installation of two 1000 MWe nuclear reactors on the same site. Scaling factors were used to extrapolate the GHG emissions of a reference 1000 MWe reactor to a two-unit nuclear reactor plant (each reactor unit generating 1000 MWe). GHG emission estimates for building, operation, decommissioning and safe storage (SAFSTOR) for a two-unit nuclear reactor plant would be based on the plant’s physical size, and therefore estimates for these source categories were assumed to be twice the value of the reference 1000 MWe reactor. However, GHG emissions from the fuel cycle (including fuel transportation) were scaled upward by a factor of 3, based on plant efficiencies greater than the 80 percent assumption in Appendix H. Table 1 below shows the PPE emissions for two 1000 MWe nuclear reactors as provided in NR GEIS. The total GHG emissions for two 1000 MWe reactors were calculated as 2,534,000 metric tons (MT) of CO 2 equivalent (CO 2 (e)) based on a 97 year GHG life cycle period. The GHG emissions lifetime of 97 years for a reference nuclear reactor includes a 7-year building phase, 40 years of operation, 10 years of active decommissioning, and 40 years of SAFSTOR operations (NRC 2024). Construction equipment and vehicular traffic from workers commute would contribute to the GHG emissions during a 7-year building phase. Uranium fuel cycle, vehicular traffic, fuel and waste transportation, and testing of standby diesel generators would contribute to GHG emissions during the 40-year operations phase. While NRC’s regulations allow up to 60 years of reactor facility decommissioning, Appendix H estimated that most of the GHGs would occur over an estimated 10-year period during which to the licensee would engage in significant demolition and earth-moving activities, as discussed in Supplement 1 to NUREG-0586 (NRC 2002). Vehicular traffic by the workforce during a 40-year SAFSTOR period would additionally contribute GHG emissions. The carbon footprint for a 40-year SAFSTOR period was separately analyzed from the decommissioning activities as provided in Table YYYY-2 of the staff issued guidance in 2011 (NRC 2011).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗