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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Automated Segmentation of Twin Boundaries in TRISO Silicon Carbide Using Deep Neural Networks

Coated particle fuels, such as the tristructural isotropic (TRISO) fuel particle, are essential for high-temperature gas reactor (HTGR) applications due to their efficiency and stability under normal and off-normal conditions. However, widespread commercialization and deployment of this technology for next-generation nuclear applications require robust quality assurance and quality control (QA/QC) methods linking fabrication, properties, and performance. Of the many important metrics for TRISO QA/QC, quantification of the silicon carbide (SiC) microstructure is critical because it correlates with fission product retention during irradiation. Previous work has shown extensive twinning of the SiC microstructure, which strongly affects microstructural metrics; however, twin grain boundaries are not expected play a significant role in fission product diffusion. This report summarizes the initial development, training, and testing of a machine learning image processing algorithm to detect twin grain boundaries in a backscattered electron image, which can be removed so that microstructural metrics can be recalculated for legacy data. Further development and deployment of this model will provide automated, scalable improvement of potential QA/QC methods for the SiC layer of TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Pros and Cons Analysis of HALEU Utilization in Example Fuel Cycles

The Systems Analysis and Integration campaign assessed the pros and cons of high-assay low-enriched uranium (HALEU) utilization in advanced reactors and associated fuel cycles. The assessment was done for three example fuel cycles at equilibrium states: once-through, limited recycle, and continuous recycle (CR) starting with HALEU. Front- and back-end fuel cycle parameters and the Levelized Cost of Fuel (LCF), which is the Levelized Cost of Electricity excluding reactor cost, of the three example fuel cycles were calculated using a single Analysis Example Reactor. The pros and cons of HALEU utilization were assessed by normalizing the fuel cycle parameters and LCF to a unit of electricity generation (GWe-year) and comparing them with a Basis of Comparison. In this study, a sodium-cooled reactor with sodium-bonded metallic fuel having a burnup of ~100 GWd/t was used as the Analysis Example Reactor because its technology readiness level is high, and the burnup and fuel enrichment are in the middle of those ranges of advanced reactor concepts that are under development. The current once-through Light Water Reactors (OT-LWRs) with <5% low-enriched uranium and 50 GWd/t burnup were used as the Basis of Comparison. In addition, a series of sensitivity analyses was conducted by varying burnup, enrichment, fuel forms, and reactor types to capture the design variations in two once-through Advanced Reactor Demonstration Program (ARDP) reactors, Natrium with sodium-free metallic fuel having a burnup of ~150 GW/t and Xe-100 with Tristructural-Isotropic (TRISO) pebble fuel having a burnup of ~168 GWd/t.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-3/4 Compact 10-4 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 10 4 taken from the Advanced Gas Reactor (AGR) experiment, AGR 3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Micromechanical Properties of the SiC and Pyrolytic Carbon Layers in Tristructural-Isotropic Coated Particles

Tristructural isotropic (TRISO) coated particle fuel was initially developed for high temperature gas cooled reactors (HTGR) and has been proposed for several other advanced reactor concepts. The design of TRISO particles focuses on preventing the release of fission products in normal and off-normal reactor conditions. The particle design features an actinide bearing fuel kernel that is surrounded by three pyrolytic carbon (PyC) layers and a silicon carbide layer (SiC). The mechanical stability of the particle and fission product retention for both metallic and gaseous fission products depend on the SiC layer. Post irradiation examination (PIE) of TRISO fuel from the first two US DOE Advanced Gas Reactor Fuel Development and Qualification Program irradiation campaigns, AGR-1 and AGR-2, had identified a low rate of particles exhibiting cracking in the SiC layer that did not propagate across the SiC layer. While cracking in the SiC is rare for test conditions and particles associated with the AGR program, understanding the stress state and mechanical properties of the SiC and PyC layers related to particle architecture can aid predicting thermomechanical response of TRISO fuel under the prescribed operation envelope and beyond as well as aiding in the development of similar fuel concepts for other advanced reactors. The presented investigation shows the relationship of the mechanical properties and mechanical response (e.g., understanding crack propagation) of the SiC and PyC layers relative to position within the particle. Testing was conducted on the inner and outer PyC layers of TRISO particles to quantify differences in mechanical behavior.

Montoya, Katherine [ORNL] (ORCID:0000000326955086)↗
DOE OSTI2021

Understanding TRISO Coated Particle Neutron Irradiation Behavior:Evolution of Advanced Micro Analysis and Electron Microscopy Approaches

Objectives and motivation• Provide data for fuel qualification in support of reactor licensing• Establish a domestic commercial TRISO fuel fabrication capability Approach • Focus is on developing and testing UCO TRISO fuel –Develop fuel fabrication and QC measurement methods, first at lab scale and then at industrial scale –Perform irradiation testingover a range of conditions (burnup, temperature, fast neutron fluence) –Perform post-irradiation examination and safety testingto demonstrate and understand performance during irradiation and during accident conditions –Develop fuel performance modelsto better predict fuel behavior –Perform fission product transport experiments to improve understanding and refine models of fission product transport

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Investigation of Coincidence Counting for Improving Minimal Detectable Activity of 110m Ag in Single Particle Gamma Analysis

Post-irradiation examination (PIE) of fuel particles from the fourth Advanced Gas Reactor Fuel Development and Qualification (AGR) Program irradiation (AGR-5/6/7) is being performed at Oak Ridge National Laboratory (ORNL). Tristructural isotropic (TRISO)-coated particles and associated compacts for the AGR-5/6/7 experiment fabricated by BWX Technologies Nuclear Operations Group were formed into a graphite matrix compact and irradiated at the Advanced Test Reactor at Idaho National Laboratory. At ORNL, particles are deconsolidated from the graphite matrix compact and individually scanned for emitted gamma rays with the Irradiated Microsphere Gamma Analyzer (IMGA). The IMGA system comprises a single high purity germanium (HPGe) detector, an automated particle handling vacuum system, and an ORTEC DSPEC-50 digital spectrometer for gamma ray analysis. IMGA quantifies gamma ray-emitting fission product inventories of individual TRISO particles, and these inventories can be compared with the measured average inventories per particle and radionuclide inventories predicted by AGR-5/6/7 physics calculations to determine if a particle experienced radionuclide release. Details on IMGA data collection methods can be found in the literature. The TRISO particle’s SiC layer provides structural support, as well as a barrier for fission product release during irradiation or subsequent safety testing. A weakened or compromised SiC layer can be identified by the release of radionuclides, such as 137 Cs, which is detected by IMGA. However, select radionuclides, such as 90 Sr, 110 mAg, and 154 Eu have been shown to migrate through an intact SiC layer. Measurement of the radionuclide 110 mAg is significant as its release has been shown to be particularly sensitive to in-reactor conditions (e.g., temperature) with broad variable particle to-particle release behaviors observed within a single compact. As such, 110 mAg activity is often used for particle selection for comprehensive PIE as bounding 110 mAg retention particles are hypothesized to represent limits in particle behaviors within a compact. As TRISO particle fuel PIE activities continue over time, IMGA measurements of the 110 mAg inventory are eventually hindered because of its relatively short half-life (~250 days). As the fuel ages from its end of irradiation (EOI) date, the measurement uncertainty and minimum detectable activity (MDA) of 110m Ag increase because the detector background continuum begins to dominate. For particles from the second AGR irradiation experiment (AGR-2), the 110m Ag MDA was above 20% of the calculated average particle inventory after approximately five half-lives, and 110m Ag activity was no longer measurable with IMGA after approximately seven half-lives. Therefore, coincidence counting approaches have been explored to determine feasibility of leveraging new approaches to overcome limitations associated with increasing MDA over time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Participation in and Assessment of the Third DNCSH Public Workshop

Advanced reactor development and deployment is introducing fuel forms, materials, and configurations that differ from those used in traditional light-water reactors (LWRs), including tristructural-isotropic (TRISO)-based fuels, moderators, high-temperature materials, and compact core designs. These attributes can fall outside the range of applicability of existing criticality safety validation bases and may require additional work in code validation and benchmark applicability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

High-temperature oxidation behavior of the SiC layer of TRISO particles in low-pressure oxygen

Surrogate tristructural-isotropic (TRISO)-coated fuel particles were oxidized in 0.2 kPa O 2 at 1200–1600°C to examine the behavior of the SiC layer and understand the mechanisms. The thickness and microstructure of the resultant SiO 2 layers were analyzed using scanning electron microscopy, focused ion beam, and transmission electron microscopy. The majority of the surface comprised smooth, amorphous SiO 2 with a constant thickness indicative of passive oxidation. The apparent activation energy for oxide growth was 188 ± 8 kJ/mol and consistent across all temperatures in 0.2 kPa O 2 . The relationship between activation energy and oxidation mechanism is discussed. Raised nodules of porous, crystalline SiO 2 were dispersed across the surface, suggesting that active oxidation and redeposition occurred in those locations. These nodules were correlated with clusters of nanocrystalline SiC grains, which may facilitate active oxidation. So these findings suggest that microstructural inhomogeneities such as irregular grain size influence the oxidation response of the SiC layer of TRISO particles and may influence their accident tolerance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Tristructural isotropic (TRISO)-coated fuel particles are designed for use in high-temperature gas-cooled nuclear reactors, featuring a structural SiC layer that may be exposed to oxygen-rich environments over 1000 °C. Surrogate TRISO particles were tested in 0.2–20 kPa O 2 atmospheres to observe the differences in oxidation behavior. Oxide growth mechanisms remained consistent from 1200–1600 °C for each P O$_2$ , with activation energies of 228 ± 7 kJ/mol for 20 kPa O 2 and 188 ± 8 kJ/mol for 0.2 kPa O 2 . At 1600 °C, kinetic analysis revealed a change in oxide growth mechanisms between 0.2 and 6 kPa O2. In 0.2 kPa O 2 , oxidation produced raised oxide nodules on pockets with nanocrystalline SiC. Oxidation mechanisms were determined using Atom probe tomography. Active SiC oxidation occurred in C-rich grain boundaries with low P O$_2$ , leading to SiO 2 buildup in porous nodules. Here, this phenomenon was not observed at any temperature in 20 kPa O 2 environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Method for measurement of TRISO kernel and layer volumes by X-ray computed tomography

Layer dimensions are key parameters for as-fabricated tristructural-isotropic (TRISO) particle fuel as well as for post-irradiation examination of particle performance. Layer thicknesses are typically measured by optical microscopy of the particle cross section near mid-plane, while layer volumes are estimated with serial sectioning and microscopy. This method for measuring layer volumes is limited due to the resolution limit imposed by slice thickness and the effect of the polishing process on delicate irradiated particle microstructures. In this study, image processing software has been developed to segregate the three-dimensional (3D) TRISO particle images provided by x-ray computed tomography (XCT) into high-resolution volumetric data for the kernel, all particle layers, and any internal voids introduced during irradiation. These data can be used to analyze key in-reactor behaviors of TRISO particles such as kernel swelling or buffer shrinkage, which are important inputs for fuel performance modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Investigation of the structure and chemical nature of Pd fission product agglomerations in irradiated TRISO particle SiC

Tristructural-isotropic (TRISO)-coated fuel particles are used in high-temperature gas-cooled nuclear reactors. Although the polycrystalline 3C-SiC layer acts as the main barrier to fission product release, post-irradiation examinations have shown that certain metallic fission products are found outside of the TRISO coated particle, with no observable micro-cracks or other mechanical failure feature. In this study, an atomic resolution transmission electron microscopy investigation of a SiC layer of a neutron irradiated (19.38% fissions per initial metal atom average burnup; 1072°C time-averaged temperature) TRISO-coated particle from the first advanced gas reactor experiment at Idaho National Laboratory containing mainly Pd and U was conducted. The fission product Pd was found to be present at dislocation cores associated with twins, stacking faults, and their intersections. The Pd present at the dislocation cores were bonded to Si, suggesting that the diffusion of Pd along the SiC dislocation cores involves the formation of Pd-Si bonds. Evidence of single atom grain boundary migration was also found. Fission product agglomerations at grain boundaries most likely consist of a random solid solution containing Si-FP, Si-C, and C-C type bonding arrangements. No difference in the Si- or C-bonding environment is found for the unoccupied grain boundary adjacent to a fission product agglomeration compared to the bulk SiC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

AGR 5/6/7 Data Qualification Report for ATR Cycles 162B through 168A

This report provides the qualification status of experimental data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation. AGR-5/6/7 was conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high temperature gas-cooled reactors. The objectives of the AGR-5/6/7 experiments are to: (i) irradiate reference-design fuel particles to support fuel qualification, (ii) establish operating margins for the fuel beyond normal operating conditions, and (iii) provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination (PIE) and safety testing. The test train contains five separate capsules that were independently controlled and monitored. Each capsule contains multiple 12.51-mm-long compacts filled with low enriched uranium carbide/oxide (UCO) TRISO fuel particles. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) is to verify successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data on fission-gas release from failed particles during irradiation. The AGR-5/6/7 capsules were irradiated in the ATR northeast flux trap location. The experiment began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of 360.9 effective full power days. The AGR 5/6/7 experiment was able to remain in the reactor core during all three Powered Axial Locator Mechanism (PALM) cycles (163A, 165A, and 167A) without overheating its fuel compacts. This report includes irradiation monitoring data from nine ATR Cycles: 162B, 163A, 164A, 164B, 165A, 166A, 166B, 167A, and 168A, as stored in the Nuclear Data Management and Analysis System (NDMAS). During irradiation, data records consisted of instantaneous measurements recorded every minute and provided by text files automatically every 2 hours. The AGR 5/6/7 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates [capsule inlet, outlet, and downstream at detector], pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates and release to birth rate ratios [R/Bs]) for each of the five capsules. A total of 94,989,908 TC temperature and sweep gas data records were received and processed by NDMAS for AGR 5/6/7 irradiation. Of these records, 41,593,387 (or 43.7% of the total) met data collection and accuracy requirements and are labeled as Qualified. A total of 57,746,693 TC temperature readings were captured from 54 installed TCs. Among them, 10,034,676 TC temperature records (only 17.4%) were Qualified and 47,701,371 TC temperatures (or 82.6%) are Failed due to 48 TC failures (63.5%) and due to missing values (19.1%). To assess performance of the operational TCs, analysis of daily correlations between TCs found no evidence of virtual junction failure for any TCs. Analyses on control charts of TC temperature differences revealed trending in TC readings for TC2, 4, 5, and 13 in Capsule 3, but there is no conclusive indication of TC drift failure that caused those trends. Therefore, TC control charts are not used to disqualify TC data, but only for users’ consideration. For sweep gas flow rates, a total of 31,519,747 gas flow records (84.4%) are Qualified for use for AGR-5/6/7 experiment; 5,723,468 gas flow records (15.4%) are Failed due mostly to missing values; and 74,641 high sweep gas flow rates (0.2 %) are Trend. A large number of Failed missing TC temperature and gas flow values were caused by an error in the data output script that outputted a ‘NULL’ value when values were unchanged. This problem was fixed during the outage of Cycle 166B, which led to a substantially decreased number of missing values during the last three cycles. Nonetheless, a large amount of non-missing data remained because of the high data acquisition frequency (1-minute) and still provided sufficient data to effectively monitor the experiment as designed. For FPMS data, NDMAS received and processed fission product release and R/B data for nine ATR cycles, when ATR core reached full power during AGR 5/6/7 irradiation. These data consist of 110,388 release rate records and 110,388 R/B records for the twelve radionuclides (Kr 85m, Kr 87, Kr 88, Kr 89, Kr 90, Xe 131m, Xe 133, Xe 135, Xe 135m, Xe 137, Xe 138, and Xe 139) for each of the five capsules. Equivalent numbers of uncertainty records associated the release rates and R/B values were provided. To date, qualification status of the FPMS data stored in the NDMAS dat

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

HERMES-400 Microcalorimetry Measurements of Nuclear Fuel at Idaho National Laboratory

Microcalorimetry updates from HERMES-400 located at Idaho National Laboratory will be presented. In particular, the current outlook as well as future upgrades will be discussed. Data will be presented for Byron and TRISO particle fuel measurements highlighting improved energy resolution compared to traditional gamma-ray spectroscopy.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Microstructure of irradiated AGR TRISO particle buffer layers as measured by X-ray computed tomography

Shrinkage of the initially low-density buffer layer in tristructural isotropic (TRISO) coated fuel particles during irradiation is a well-known phenomenon with potential implications for fission product and actinide transport, as well as potential pyrocarbon fracture that in some cases has been observed to impact particle performance. During post-irradiation examination, the buffer layer's structure is commonly determined using 2D microscopy of a particle cross section or a series of particle cross sections at staggered depths. Although these methods provide a general idea of the irradiated buffer microstructure, they do not provide a full picture of the TRISO coating microstructure. By contrast, x-ray computed tomography (XCT) provides full 3D imaging of the TRISO particle. Particles from various compacts irradiated for the Advanced Gas Reactor Fuel Development and Qualification Program were imaged using XCT. Some of these particles were selected because of abnormal fission product inventories (e.g., low 137 Cs), whereas others were randomly selected from the center of the fission product inventory distribution. Qualitative and quantitative analysis techniques were applied to the randomly selected particles as representatives of typical TRISO behavior to study the post-irradiation structure of the buffer layer. These results showed that, while separation of the buffer and inner pyrolytic carbon layers was a common behavior during the radiation-induced shrinkage of the buffer, a portion of the original buffer/inner pyrolytic carbon interface remained intact throughout irradiation in nearly all cases. Furthermore, these results also indicated clear trends in the degree of buffer densification with irradiation temperature and fluence.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Multi-scale fission product release model with comparison to AGR data

TRistructural ISOtropic (TRISO) particle fuel is central to several advanced, high-temperature reactor designs. Each particle consists of a fuel kernel encapsulated by three layers of carbon and ceramics that prevent the release of fission products and ensure physical integrity. Despite outstanding retention properties, fission product release has been observed from intact particles. To better understand and quantify fission product release from TRISO particles, a multiscale, mechanistic model of fission product transport is being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Previous work focused on silver (Ag) transport and improved Ag release predictions. The work described in this report builds on this experience to better understand cesium (Cs) transport in silicon carbide (SiC), the main barrier to the release of fission products. Atomistic simulations provide bulk and grain boundary (GB) Cs diffusivities in SiC, which are used by phase field simulations in the mesoscale code Marmot to determine the temperature, microstructure, and irradiation-dependent Cs diffusivity at the mesoscale in SiC. This approach attributes the different temperature regimes experimentally observed for Cs diffusivities in SiC to a transition from bulk-dominated diffusivity at high temperatures to a GB-dominated regime at low temperatures, providing new insight. The multiscale, mechanistic effective diffusivity is then implemented in the fuel performance code BISON and further validated by comparing Cs release predictions from Advanced Gas Reactor (AGR)-1 and AGR-2 post-irradiation measurements. The new model improves BISON’s predictability. This document also reports improvements made on Ag transport modeling by accounting for different GB types having different diffusivities. Moreover, this report details preliminary efforts to model palladium (Pd) attack of the SiC at the mesoscale using a phase field approach. Pd attack and its impact on accelerated Ag transport remains a misunderstood phenomenon, and we use the model to demonstrate that the formation of lamellae that has been observed in experiments can be explained by the reaction of Pd with SiC to form alternating layers of graphite and Pd 2 Si. This effort aims to improve our understanding of the reaction and eventually provide a model for BISON to account for Pd penetration and its effects on fission product release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗