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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Multiphysics Meshfree Degradation Modeling of Energy Storage Materials with Kernel Enrichment

Energy storage materials exhibit strong electro-chemo-mechanical coupling and highly anisotropic material properties, contributing to the formation and propagation of micro-cracking during charge/discharge cycling and ultimately diminishing performance and service life. With microstructural images supplied by the National Renewable Energy Laboratory (NREL), pixel-based meshfree model construction by the reproducing kernel particle method (RKPM) is used to represent the complex material microstructures that dictate the coupled physics of these systems. Traditional electro-chemo-mechanical models rely on mesh-based finite element methods, which can lead to difficulties in meshing such complex geometries and capturing crack propagation due to mesh dependency. The first kernel enrichment discussed will be the interface modified reproducing kernel (IM-RK) [1, 2], constructed by scaling a smooth kernel function with an interface-distance function to achieve strategic discontinuity types (i.e. weak discontinuities for strain discontinuities and strong discontinuities for cracks) and alleviate Gibbs oscillations near these transition zones. The IM-RK is especially useful for areas in which a known discontinuity-type is expected a priori. The second kernel enrichment to be discussed is a neural network-enhanced reproducing kernel (NN-RK) [3, 4], which is introduced to effectively model non-obvious damage and crack propagation in the material microstructures; the location, orientation, and solution transition near a localization are automatically captured by superimposed block-level NN optimizations. This NN enrichment approach allows for effective modeling of localizations via a fixed background discretization, relieving tedious efforts for adaptive refinement in traditional mesh-based methods. Applications to the heterogeneous microstructures of Li-ion battery cathodes will be presented to demonstrate the effectiveness of the proposed methods. NN-RK is additionally used to inform how crack opening and closure in turn affect the electro-chemo-mechanical responses in the material microstructure. Reference: [1] Wang, Y., Baek, J., Tang, Y. et al. "Support vector machine guided reproducing kernel particle method for image-based modeling of microstructures," Comput Mech 73, 907-942 (2024). https://doi.org/10.1007/s00466-023-02394-9. [2] Susuki, K., Allen, J. & Chen, J. S.. "Image-based modeling of coupled electro-chemo-mechanical behavior of Li-ion battery cathode using an interface-modified reproducing kernel particle method," Engineering with Computers (2024). https://doi.org/10.1007/s00366-024-02016-9. [3] Baek, J., Chen, J. S., Susuki, K., "Neural Network enhanced Reproducing Kernel Particle Method for Modeling Localizations," International Journal for Numerical Methods in Engineering, Vol. 123, 4422-4454 (2022). https://doi.org/10.1002/nme.7040.

25 ENERGY STORAGE↗
DOE OSTI2026

Kernel Enriched Meshfree Multiphysics Degradation Modeling of Energy Storage Materials

Energy storage materials exhibit strong electro-chemo-mechanical coupling and highly anisotropic material properties, contributing to the formation and propagation of micro-cracking during charge/discharge cycling and ultimately diminishing performance and service life. With microstructural images supplied by the National Laboratory of the Rockies (NLR), pixel-based meshfree model construction by the reproducing kernel particle method (RKPM) is used to represent the complex material microstructures that dictate the coupled physics of these systems. Traditional electro-chemo-mechanical models rely on mesh-based finite element methods, which can lead to difficulties in meshing such complex geometries and capturing crack propagation due to mesh dependency. The first kernel enrichment discussed will be the interface modified reproducing kernel (IM-RK) [1, 2], constructed by scaling a smooth kernel function with an interface-distance function to achieve strategic discontinuity types (i.e. weak discontinuities for strain discontinuities and strong discontinuities for cracks) and alleviate Gibbs oscillations near these transition zones. The IM-RK is especially useful for areas in which a known discontinuity-type is expected a priori. The second kernel enrichment to be discussed is a neural network-enhanced reproducing kernel (NN-RK) [3, 4], which is introduced to effectively model non-obvious damage and crack propagation in the material microstructures; the location, orientation, and solution transition near a localization are automatically captured by superimposed block-level NN optimizations. This NN enrichment approach allows for effective modeling of localizations via a fixed background discretization, relieving tedious efforts for adaptive refinement in traditional mesh-based methods. Applications to the heterogeneous microstructures of Li-ion battery cathodes will be presented to demonstrate the effectiveness of the proposed methods. NN-RK is additionally used to inform how crack opening and closure in turn affect the electro-chemo-mechanical responses in the material microstructure. References: [1] Wang, Y., Baek, J., Tang, Y. et al. "Support vector machine guided reproducing kernel particle method for image-based modeling of microstructures," Comput Mech 73, 907-942 (2024). https://doi.org/10.1007/s00466-023-02394-9. [2] Susuki, K., Allen, J. & Chen, J. S.. "Image-based modeling of coupled electro-chemo-mechanical behavior of Li-ion battery cathode using an interface-modified reproducing kernel particle method," Engineering with Computers (2024). https://doi.org/10.1007/s00366-024-02016-9. [3] Baek, J., Chen, J. S., Susuki, K., "Neural Network enhanced Reproducing Kernel Particle Method for Modeling Localizations," International Journal for Numerical Methods in Engineering, Vol. 123, 4422-4454 (2022). https://doi.org/10.1002/nme.7040.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

AGR-1 Depletion Benchmark

The AGR-1 TRISO particle irradiation test was the first Advanced Gas Reactor test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The TRISO particle fuel was 350 μm diameter uranium oxy-carbide (UCO) kernels enriched to 19.75 wt% U-235. The particles were embedded in cylindrical graphite compacts, placed in test capsules, and irradiated over a period of three years or 13 ATR power cycles. The fuel compacts achieved burnups approaching 20% FIMA and accumulated fast fluence up to 4.25E+25 n/m 2 (E n >0.18 MeV). This report describes the AGR-1 test train, capsules, compacts, and TRISO particle fuel in sufficient detail for a physics analyst to model the AGR-1 experiment as part of a larger ATR neutron transport model and perform a depletion calculation. The intent of the data herein is to define a practical numerical TRISO particle depletion benchmark problem for the AGR 1 irradiation test.

42 ENGINEERING↗
DOE OSTI2020

Destructive PIE and Safety Testing of Six AGR-2 UO 2 Capsule 3 Compacts

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2) was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from June 2010 to October 2013 (Collin 2014). The fuel compacts in this experiment held either tristructural isotropic (TRISO)-coated spherical kernels of uranium oxide (UO2) or TRISO-coated kernels containing both uranium carbide and uranium oxide phases (UCO). There were six separately monitored and controlled capsules in the AGR-2 test train. Capsule 3 held twelve compacts containing UO2-TRISO particles fabricated by BWX Technologies Nuclear Operations Group. The AGR-2 TRISO particles were fabricated in a pilot-scale fluidized-bed chemical vapor deposition (FB-CVD) furnace with a coating chamber inner diameter of 150 mm (Phillips, Barnes, and Hunn 2010), which was a change from the first irradiation experiment (AGR-1) particles that had been coated in a lab-scale FB-CVD coating system with a chamber inner diameter of 50 mm (Lowden 2006). The TRISO particles were overcoated with resinated graphite flake at Oak Ridge National Laboratory (ORNL), and the overcoated particles were pressed into one-inch-long, half-inch-diameter cylinders (Hunn, Montgomery, and Pappano 2010). Each cylindrical compact held an average of 1,543 TRISO particles with 9.6% enriched UO2 kernels that had a nominal diameter of 500 μm (Hunn, Savage, and Silva 2012). Capsule 3 compacts were irradiated to average calculated burnups of 9.01–10.69% fissions per initial metal atom (FIMA), and the average calculated fluences of fast neutrons with energies E n > 0.18 MeV were 3.05–3.53×10 25 n/m 2 (Sterbentz 2014). The calculated time-average, volume-average Capsule 3 compact temperatures were 996–1,062°C. However, Capsule 3 compact temperatures varied several hundred degrees across each compact, and the timeaverage minimum (TA min ) and time-average maximum (TA max ) temperatures were between 889–999°C and 1,072–1,105°C, respectively (Hawkes 2014). After irradiation, the AGR-2 test train was transferred from ATR to the INL Materials and Fuels Complex for inspection and disassembly (Ploger, Demkowicz, and Harp 2015). The initial inspection included dimensional metrology of the compacts and graphite fuel holders. Like all the AGR-2 compacts, the compacts in Capsule 3 shrank slightly during irradiation, as expected, with an average length reduction of 1.07–1.24% and an average diameter reduction of 0.13–0.41%. Post-irradiation examination (PIE) of the capsule components was completed to measure select fission products ( 90 Sr, 110m Ag, 134 Cs, 137 Cs, 144 Ce, and 154 Eu) outside the compacts (Stempien and Demkowicz 2020). This involved gamma counting of the graphite and graphoil spacers at the top and bottom of each capsule, acid leaching for radiochemical analysis of fission products on the metallic capsule components, and burn-leach analysis of the graphite holders. The total amount of 110mAg measured on the Capsule 3 components was 13% of the calculated capsule inventory. This was significantly lower than the amount of 110m Ag measured on the three UCO capsule components, which ranged from 32–70%. The lower 110m Ag release in Capsule 3 was likely due to lower peak temperatures compared with the UCO fuel capsules (Hawkes 2014). Measured inventories of the other select fission products were also lower in Capsule 3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Microstructural and Micro-Chemical Evolutions in the Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles

AGR-1 and AGR-2 TRISO fuel particles were fabricated with slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. To correlate those differences with the fuel kernel responses to neutron irradiations in terms of irradiated fuel microstructure, fission products chemical and physical states, and fission gas bubble evolutions, extensive microstructural and analytical characterizations were conducted. The studies used a state of art transmission electron microscopy (TEM) equipped with Energy-dispersive X-ray spectroscopy (EDS) of four silicon solid-state detectors which have super sensitivity and fast speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen to represent a representative irradiation conditions with a fuel burnup within the range from 10.8 to 18.6% FIMA, and the time-average volume-average temperatures vary from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% for the AGR1 fuel kernels and 14.03 wt.% for the AGR-2 fuel kernels. The TEM results show that there were significant microstructural reconstructions in the irradiated fuel kernels for both the AGR-1 and AGR-2 fuels. There are four major phases including fuel matrix of UO2 and UC, U2RuC2, and UMoC2 in the irradiated AGR2 fuel kernel. Zr and Nb form solid solution in the UC phase. UMoC2 phase often shows a detectable concentration of Tc. Pd was found to mainly locate in the buffer layer or to be associated with fission gas bubble within the UMoC2 phase. The EDS maps qualitatively show that the rare-earth fission products (Nb, et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there is a high number of rod-shape precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The difference of microstructural and micro-chemical evolutions in irradiated fuel kernels between the AGR-1 and AGR-2 TRISO fuel particle may result from a combined factor of irradiation temperature, fuel geometry and chemical composition. However, the irradiation temperature probably play a more deterministic role. Limited electron energy loss spectroscopy (EELS) characterizations on the AGR2 fuel kernel show that there is nearly no carbon in the UO2 phase while a small fraction of oxygen was detected in the UC/UMoC2 phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

UCO TRISO Minifuel FY23 NSUF-Kairos Power Post-Irradiation Examination Status Report

Irradiation of miniature tristructural isotropic (TRISO)–coated particle fuel compacts at high-power particle was performed in the Oak Ridge National Laboratory’s (ORNL’s) High Flux Isotope Reactor (HFIR) using the MiniFuel irradiation capability. Each compact comprised 20 TRISO particles with a low-enriched uranium carbide uranium oxide (UCO), natural UCO, or low-enriched UO 2 kernel within a graphitic matrix. After irradiation, the MiniFuel targets and subcapsules were disassembled to recover the irradiated fuel specimens and pursue post-irradiation examination (PIE) to inform Kairos Power on the fuel specimen performance. This report describes the PIE results collected to date, including dilatometry on the passive thermometry to confirm the irradiation temperature, fission gas release measurements, and gamma counting. This work was funded by the Nuclear Science User Facilities program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Disposition Plan for AGR-Related Equipment and Materials

The Advanced Gas Reactor (AGR) program has purchased equipment, chemicals, and acquired uranium feedstocks for the purpose of developing fuel fabrication processes, fabricating fuel for irradiation experiments, and growing domestic expertise. During the process of fuel development and qualification activities, various quantities of unused chemicals, uranium feedstocks, intermediate materials, scrap, and waste have accumulated at BWX Technologies Nuclear Operations Group (BWXT-NOG) and Oak Ridge National Laboratory (ORNL) that must be properly dispositioned. More expensive standalone equipment will be shipped to and repurposed at the Idaho National Laboratory(INL) and ORNL. Less expensive standalone equipment will be either abandoned in place for use by BWXT-NOG or discarded. Embedded equipment, that effects the functionality of BWXT-NOG systems, will be either abandoned in place or dismantled for disposal. Residual chemicals and non-fuel materials used for fuel fabrication and characterization will be repurposed when economically viable or discarded. High-purity uranium oxides, crystals, and metal will be repurposed for fuel fabrication. The majority of the high-enriched uranium will be included in an upcoming Central Scrap Management Office down blending operation. Archival materials will be collected of low-enriched uranium fuel kernels, tristructural isotropic (TRISO) fuel particles, and fuel compact that are residues of AGR fuel fabrication activities. Archival materials will be stored at either INL or ORNL. Some quality, unenriched uranium materials, that may be of interest to other parties investigating TRISO fuels, will also be archived. Uranium materials that are not archived and which cannot be readily repurposed will be discarded.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Analysis and Design for Irradiation of High Power TRISO Fuel Compact Specimens in HFIR

Tristructural isotropic (TRISO) fuel is being proposed for use in several high-temperature advanced reactor concepts because of its structural integrity under high operating temperatures and burnup. One of these advanced reactor concepts is the Kairos Power fluoride salt-cooled high-temperature reactor (KP- FHR) under development by Kairos Power, LLC. Previous TRISO irradiation experiments were focused on qualification for high-temperature gas reactors (HTGRs), which have higher operating temperatures but lower particle powers than the KP-FHR design. To study the performance of TRISO fuel designed for HTGRs under prototypical FHR conditions, a set of experiments was designed using the MiniFuel irradiation vehicle at the Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR). The experiments will irradiate 30 TRISO-containing carbon matrix compacts at inner small vertical experiment facilities in HFIR. Each compact will contain 20 TRISO particles (600 particles total) developed for HTGRs, consisting of either 14% enriched uranium dioxide uranium carbide, naturally enriched uranium dioxide, uranium carbide, or 9.6% enriched uranium dioxide fuel kernels with time- and volume-averaged silicon carbide layer temperatures between 500 and 900°C. This report summarizes the vehicle designs that have been developed, as well as the neutronic and thermal analyses completed for these irradiation experiments. These analyses show that MiniFuel compact irradiation is a versatile experiment that can be used to study a range of TRISO particle powers and fuel types while providing reasonable separation of burnup and temperature effects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

AGR-5/6/7 Irradiation Results and Reporting

Final fuel qualification irradiation (AGR-5/6) and performance margin test (AGR-7) 425 µm diameter kernels with 15.5% 235U enrichment Target time-average peak fuel temperatures ~1500°C Target peak burnup 18% FIMA 194 UCO fuel compacts (~570,000 particles) in five capsules Irradiation started Feb 2018 in ATR Northeast Flux Trap

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Investigation of sol-gel feedstock additions and process variables on the density and microstructure of UN microspheres

The kernel of fully ceramic microencapsulated (FCM) fuel requires a material with high fissile density. For that reason, among others, UN is the appropriate chemical state for low enriched U fuel. The UN kernels are spheres ~800 μm in diameter made using a sol-gel process. The effect of additives on the chemistry and density of the UN microspheres are investigated in this work. Gadolinium nitrate hexahydrate, Gd 2 O 3 , B, SiO 2 and SiC were incorporated into the sol-gel broth in varying concentrations. It was found that Gd can serve as both a sintering aid and burnable poison when added to the sol-gel broth as a Gd(NO 3 ) 3 6H 2 O. However, even with the increased theoretical density of the UN microspheres, the U density was still too low for the FCM design that replaces UO 2 pellets in a commercial light water reactor. Silicon carbide and B were also successfully added but produced a lower density final product. Other sol-gel processing variables were investigated. In conclusion, the pour density of the sol-gel feedstock was found to influence the final converted UN kernel density.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Microstructure and fission products in the UCO kernel of an AGR-1 TRISO fuel particle after post irradiation safety testing

Tristructural isotropic (TRISO) coated particle fuel as a key fuel concept for the High Temperature Gas-cooled Reactors (HTGRs) and a candidate accident tolerant fuel (ATF) has been currently investigated under the US-DOE Advanced Gas Reactor (AGR) fuel development and qualification program. Over the past decades, extensive studies have been conducted to evaluate the fission product release, diffusion on Ag, Pd and Cd in the SiC layer and the TRISO coating system, and post-safety-test performance. However, to date, there are limited reported results on the fuel kernels’ response to irradiation with or without post irradiation high temperature annealing. To incrementally fulfill this knowledge gap extensive studies using transmission electron microscopes (TEM) and atom probe tomography (APT) were conducted on a TRISO fuel particle kernel with a 19.74% 235U enrichment, irradiated to 18.63% FIMA and subsequently subjected to post-safety testing at 1600ºC for 300 hours.. Microstructure characterizations, elemental analysis, and phase identification were conducted using conventional TEM and scanning TEM (STEM) imaging, energy-dispersive X-ray spectroscopy (EDS), selected-area electron diffraction (SAED) and atom probe tomography (APT). The following findings were made: (1) significant reconstructions and phase evolutions occurred in the irradiated and post-safety annealed fuel kernel, and its microstructure consists of two primary phases, namely a “high-Z (atomic mass)” UC phase and a “low-Z” UO phase, (2) no fission gas bubbles are identified within the fuel kernel, that can be attributed to the high temperature post safety annealing, (3) fission products Zr, Nb, Mo, Ru, Tc and Rh were found to preferentially segregate into UC phase or to form metallic precipitates, while the lanthanide fission products tend to stay in the solution of UO phase, (4) Pd was detected in the rod-shaped precipitates, (5) APT data reveals segregation of Te, Pd, and Ag at the UO and UC phase boundary, and (6) Pd is detected in both the UO and UC phases but with an enrichment in the former one.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Additional Confirmatory LBL Analysis of AGR-5/6/7 Compacts and Over-Coated Particles

Fuel compacts for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) were fabricated by BWX Technologies Nuclear Operations Group (BWXT NOG) located in Lynchburg, Virginia. Two compact packing fractions (PFs) were produced - nominally, 40% PF and 25% PF - in which the tristructural-isotropic (TRISO)-coated particle volume was targeted to be approximately 40% and 25% of the total compact volume, respectively. The TRISO coatings were deposited using a 150 mm diameter fluidized-bed chemical vapor deposition (CVD) furnace on spherical kernels that were nominally 425 μm in diameter. The kernels were from kernel composite Lot J52R-16-69317, which contained low-enriched uranium (15.5% 235 U) in a mixture of uranium carbide and uranium oxide (UCO). Kernels were coated with four concentric CVD layers: a porous carbon buffer layer that was nominally 100 μm thick, an inner pyrolytic carbon (IPyC) layer that was nominally 40 μm thick, a silicon carbide (SiC) layer that was nominally 35 μm thick, and an outer pyrolytic carbon (OPyC) layer that was nominally 40-μm-thick. Coated particle composite J52R-16-98005 was overcoated (OC) with a graphite/resin blend, and these OC TRISO particles were pressed to form cylindrical compacts that were nominally a half inch in diameter and one inch long.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Determination of Average Burnup in AGR-2 Compacts 3-3-1, 3-3-2, 2-2-2, and 6-4-2

Burnup measurements have been performed on irradiated tristructural-isotropic (TRISO) particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2). The AGR-2 irradiation experiment used TRISO particles fabricated by BWX Technologies (Barnes and Marshall 2009) and formed into compacts in the Oak Ridge National Laboratory (ORNL) Coated Particle Fuel Development Laboratory. Two compact lots were fabricated that differed in kernel type. The compact lot designated by LEU09-OP2-Z contained kernels made up of uranium carbide and uranium oxide (UCO) that were nominally 425 µm in diameter and 14% enriched (Hunn, Montgomery, and Pappano 2010a). The compact lot designated by LEU11-OP2-Z contained kernels made of only uranium dioxide (UO 2 ) that were nominally 500 µm in diameter and 9.6% enriched (Hunn, Montgomery, and Pappano 2010b). The addition of carbide to the kernel provides a getter for excess oxygen generated by the fission of UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Multiphysics Degradation Modeling of Energy Storage Materials via RKPM with a Neural Network-Enhancement

In energy storage materials, strong electrochemical-mechanical coupling and highly anisotropic material properties contribute to the formation and propagation of micro-cracking during charge/discharge cycling, resulting in reduced performance and service life. A coupled electro-chemo-mechanical reproducing kernel particle method (RKPM) formulation is developed, and a patch-test is formulated to certify optimal convergence of the proposed RKPM method for the coupled physics system. With microstructural images supplied by the National Renewable Energy Laboratory (NREL), pixel-based model construction by RKPM is then used to represent the complex material microstructures for modeling the coupled physics of these systems. Further, a neural network-enhanced reproducing kernel particle method (NN-RKPM) [1, 2] is introduced to effectively model damage and crack propagation in the material microstructures; the location, orientation, and solution transition near a localization are automatically captured by superimposed block-level NN optimizations. This NN enrichment approach allows for effective modeling of localizations via a fixed background discretization, relieving tedious efforts for adaptive refinement in traditional mesh-based methods. Applications to the heterogeneous microstructures of Li-ion battery cathodes will be presented to demonstrate the effectiveness of the proposed methods. Reference: [1] Baek, J., Chen, J. S., Susuki, K., "Neural Network enhanced Reproducing Kernel Particle Method for Modeling Localizations," International Journal for Numerical Methods in Engineering, Vol. 123, pp 4422-4454, https://doi.org/10.1002/nme.7040, 2022. [2] Baek, J., Chen, J. S., "A Neural Network-Based Enrichment of Reproducing Kernel Approximation for Modeling Brittle Fracture", Computer Methods in Applied Mechanics and Engineering Vol. 410, 116590, 2024.

electro-chemo-mechanical coupling↗
DOE OSTI2022

Simultaneous dissection of grain carotenoid levels and kernel color in biparental maize populations with yellow-to-orange grain

Maize enriched in provitamin A carotenoids could be key in combatting vitamin A deficiency in human populations relying on maize as a food staple. Consumer studies indicate that orange maize may be regarded as novel and preferred. This study identifies genes of relevance for grain carotenoid concentrations and kernel color, through simultaneous dissection of these traits in 10 families of the US maize nested association mapping panel that have yellow to orange grain. Quantitative trait loci were identified via joint-linkage analysis, with phenotypic variation explained for individual kernel color quantitative trait loci ranging from 2.4% to 17.5%. These quantitative trait loci were cross-analyzed with significant marker-trait associations in a genome-wide association study that utilized ~27 million variants. Nine genes were identified: four encoding activities upstream of the core carotenoid pathway, one at the pathway branchpoint, three within the α- or β-pathway branches, and one encoding a carotenoid cleavage dioxygenase. Of these, three exhibited significant pleiotropy between kernel color and one or more carotenoid traits. Kernel color exhibited moderate positive correlations with β-branch and total carotenoids and negligible correlations with α-branch carotenoids. These findings can be leveraged to simultaneously achieve desirable kernel color phenotypes and increase concentrations of provitamin A and other priority carotenoids.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2024

Leveraging a Neural Network-Enhanced Reproducing Kernel Particle Method for Multiphysics Degradation Modeling of Energy Storage Materials

Energy storage materials exhibit strong electro-chemo-mechanical coupling and highly anisotropic material properties, contributing to the formation and propagation of micro-cracking during charge/discharge cycling and resulting in reduced performance and service life. A coupled electro-chemo-mechanical reproducing kernel particle method (RKPM) formulation has been developed to analyze this system. With microstructural images supplied by the National Renewable Energy Laboratory (NREL), pixel-based model construction by RKPM is used to represent the complex material microstructures that dictate the coupled physics of these systems. Traditional electro-chemo-mechanical models rely on mesh-based finite element methods, which can lead to difficulties in meshing such complex geometries and capturing crack propagation due to mesh dependency. Here, a neural network-enhanced reproducing kernel particle method (NN-RKPM) [1, 2] is introduced to effectively model damage and crack propagation in the material microstructures; the location, orientation, and solution transition near a localization are automatically captured by superimposed block-level NN optimizations. This NN enrichment approach allows for effective modeling of localizations via a fixed background discretization, relieving tedious efforts for adaptive refinement in traditional mesh-based methods. Applications to the heterogeneous microstructures of Li-ion battery cathodes will be presented to demonstrate the effectiveness of the proposed methods. NN-RKPM is additionally used to inform how crack opening and closure in turn affect the coupled chemical equations and material microstructure. Reference: [1] Baek, J., Chen, J. S., Susuki, K., "Neural Network enhanced Reproducing Kernel Particle Method for Modeling Localizations," International Journal for Numerical Methods in Engineering, Vol. 123, pp 4422-4454, https://doi.org/10.1002/nme.7040, 2022. [2] Baek, J., Chen, J. S., "A Neural Network-Based Enrichment of Reproducing Kernel Approximation for Modeling Brittle Fracture", Computer Methods in Applied Mechanics and Engineering Vol. 410, 116590, 2024.

degradation↗
DOE OSTI2018

Microstructure and Fission Product Distribution Examination in the UCO kernel of TRISO Fuel Particles

The effect of irradiation and fabrication process on the UCO fuel kernel, kernel-buffer interlayer and fission products, has not been fully studied for UCO fuel kernels of TRISO coated particles. Specifically, the potential correlation(s) with fission product retention, composition and distribution within these structures for the tristructural isotropic (TRISO) fuel performance at normal and accident conditions, are not well reported in literature. Structural and chemical information in the irradiated UCO fuel kernel, e.g., the stoichiometry variations, may further contribute to fuel performance modeling under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The microstructural and micro-chemical evolution of selected irradiated UCO fuel kernels, kernel buffer interfaces and recoil zones were characterized using Scanning Transmission Electron Microscopy (STEM) and Energy Dispersive Spectroscopy (EDS) elemental mapping. The fuel kernel of particles from Advanced Gas Reactor (AGR)-2 irradiated to 10.8% fissions per initial metal atom (FIMA), were examined with the following main findings: (1) large fission gas bubbles were mainly identified in the zirconium and molybdenum enriched but oxygen depleted phases in the center of fuel kernel, (2) significant inter-diffusion between the UCO fuel kernel and pyro-carbon buffer layer was observed as a high density of sub-micron sized carbon inclusions was identified at near the interface in the kernel side and some isolated UCO particles were located in the buffer layer, (3) pores filled with free carbon were also observed in the recoil zone specimen. Fuel kernels from two AGR-1 compacts irradiated to 11.4 and 17.4% FIMA respectively, were examined with the following main findings: (1) two primary phases consisting of a “high-Z (atomic mass)” UC(O) phase (rocksalt; a=0.496 nm) that also contained Zr, Mo, and/or Ru and a “low-Z” UO2(C) phase (fluorite structure; a = 0.547nm) in which trace amounts of I, Nd, Pr, and/or Er could be detected. This latter phase contained many precipitates of the “high-Z” phase. Conversely, this latter phase did not contain precipitates of the UO2-type material, (2) Xe bubbles were found almost exclusively in the “high-Z” phase and in the particular example shown, the Zr-containing variant. Further work in this area should address whether this potential Xe preference for just one of the high-Z phases is indeed a real effect.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 (LEUO2), 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗