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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

TRISO Fuel Development and Qualification in the US

Presentation of TRISO Fuel Development and Qualification in the US. Including Status of HTGR and TRISO Technology Circa 2000, Advanced Gas Reactor focus on licensing and domestic TRISO fuel fabrication capabilities. Irradiation experiments discussed include AGR-1, AGR-2, AGR-5/6/7, and AGR-3/4. Results, of tests and Kernel and coating behaviors during irradiation, Fission product release from UCO Fuel Compacts, study of failed particles, safety and accident performance summaries, and Parfume modeling and capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview

DOE AGR fuel development and qualification program overview to include TRISO fuel development and program elements, AGR timeline, Irradiation test results, performance evaluation results, major accomplishments and milestones over 12 months, and university-led TRISO related research projects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE fuel work from the final test irradiation (Advanced Gas Reactor [AGR]-5/6/7) is expected to commence at INL in early 2021, but the PIE preparations work began in FY2016. The work scope in this statement of work includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities; technical input to the moisture/air-ingress furnace design, fabrication, and equipment qualification; and technical support for development of equipment and techniques for planned PIE evolutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under potential accident conditions. PIE fuel work from the final test irradiation (Advanced Gas Reactor [AGR]-5/6/7) is expected to commence at INL in early 2021, but the PIE preparations work began in FY2016. The work scope in this statement of work includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities; technical input to the moisture/air-ingress furnace design, fabrication, and equipment qualification; and technical support for development of equipment and techniques for planned PIE evolutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-3/4 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), began at INL in April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, preparations for PIE testing and analysis, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

INL ART AGR-5/6/7 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the last in the series of test irradiations, Advanced Gas Reactor (AGR)-5/6/7, will begin at Idaho National Laboratory (INL) in approximately September 2020, but because of the complexity of the experiment preparations for the PIE are going to begin in FY2016. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing project management and technical support to PIE-related activities, providing technical input to the moisture furnace design, providing technical support for development of equipment and techniques for planned PIE evolutions, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

DOE Advanced Gas Reactor Fuel Development and Qualification Program

Breakdown of the DOE Advanced Gas Reactor Fuel Development and Qualification Program. Including discussion topics on TRISO technology Status Circa 2000, New Production Reactor (NPR) Fuel Experience, Fuel Qualification, US DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program, Initial AGR Program Reference HTGR Design, Fuel Qualification Approach, Fuel Performance Modeling, Fuel Fabrication, Selected AGR-1, AGR-2, and AGR-5/6/7 Fuel Property Means, AGR Program TRISO Fuel Key Performance Data, Irradiation Performance: Fission Gas R/B, Irradiation Testing Results, Kernel and Coating Behavior During Irradiation, Locating and Studying Failed Particles Greatly Improves Understanding of Fuel Performance, Fission Product Release from UCO Fuel Compacts: AGR-1 and AGR-2 Examples, HTGR Accident Safety Testing of TRISO Fuel, Evaluating Behavior During D-LOFC Accidents, Safety Test Results for US UCO Fuel, Particle Failure Evaluation, Fuel Performance Summary, Ongoing Work and Outstanding Data Needs, Core Oxidation, Industry Engagement, and Coated-Particle-Fueled Reactor Concepts and Fuel Designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Role of microstructure on CO corrosion of SiC layer in UO₂-TRISO fuel

The Advanced Gas Reactor Fuel Qualification and Development (AGR) program has focused on qualification of UCO kernel tristructural-isotropic (TRISO) particle fuel relative to UO₂ kernel TRISO particle fuel. However, a UO₂ kernel variant was included in the second AGR irradiation experiment (AGR-2) for comparison and to connect to historic fuel irradiation data. The development of a multiscale, post-irradiation examination (PIE) analysis approach through the AGR Program has allowed for a comprehensive understanding of individual particle failure. This approach has been applied to gain an understanding of SiC layer failure in UO₂ kernel TRISO fuel from AGR-2 after safety testing at 1600–1700 °C. Particle failure by intergranular CO corrosion, facilitated by a compromised inner pyrolytic carbon layer, has been confirmed through the combined application of modern x-ray computed tomography and electron microscopy techniques. In addition, a relationship between grain boundary character and CO corrosion has been identified. This finding provides an opportunity to develop mitigating strategies to improve the resilience of the SiC layer to CO corrosion in UO₂ TRISO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview

Slides discussing the DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview to include focus on LEU UCO TRISO fuel in cylindrical compacts, with data for fuel qualification in support of reactor licensing and a path to establish a domestic commercial TRISO fuel fabrication capability. Timelines from 2004 through 2029. Major program highlights include: Developed fuel performance models based on past observed behavior, Study to understand differences between US and German (“good”) fuel, Developed and improved fuel fabrication and characterization methods, Fabricated lab-scale fuel (AGR-1) with very high quality; excellent in-pile and post-irradiation high-temperature fuel performance, Scale-up coating fabrication (AGR-2) with excellent in-pile performance, AGR-3/4 fuel fabrication and irradiation experiment to assess FP transport behavior, Scale-up matrix fabrication (AGR-5/6/7); & fuel performance evaluation in progress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Thermal Model Heat Rate Predictions of the AGR-5/6/7 Experiment

Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train. AGR-5/6/7 will test low-enriched uranium oxycarbide (UCO – a heterogeneous mixture of uranium oxide and uranium carbide) TRISO fuel. The AGR-5/6 portion of the experiment will provide data to support qualification of the selected reference fuel design, while the AGR-7 portion will serve as a margin test, irradiating the fuel beyond normal operating conditions. During irradiation, the temperature in each capsule is controlled by varying the helium / neon gas mixture flowing through the capsule until the desired thermocouple temperature is reached. Further adjustments can be made by changing the lobe power and the neutron filter. Three neutron filters are available to use as the experiment burns up the fissionable fuel. A thermal finite element model has been created for the five capsules comprising the AGR-5/6/7 experiment. Heat rates from a physics analysis are imported into the model along with fast neutron fluence. Graphite shrinkage due to the fast neutron fluence is incorporated into the model. Gas gaps change as a function of fast neutron fluence. Radiation heat transfer is a major contributor in this model. This is a large model with more than 1 million finite element brick elements. More than 150 different parts are modeled in the finite element model and communicate with each other from a heat transfer sense. More than 50 thermocouples are used in the experiment and are compared to actual measurements. The experiment is composed of five separate stainless steel capsules all welded together. This paper presents a method used to predict the gas mixture in the event that all of the thermocouples fail in a particular capsule. Curve fitting of heat rates from previous cycles and projecting them into the future is the basis of the method.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Results of the AGR-5/6/7 UCO TRISO fuel irradiation test in the Advanced Test Reactor

AGR-5/6/7 was the last in a series of irradiation experiments sponsored by the U.S. Department of Energy in support of the development and qualification of TRISO coated particle fuel for use in a high-temperature gas-cooled reactor. This experiment was conducted to verify the performance of the reference-design TRISO-coated low-enriched UCO fuel for modular high-temperature gas-cooled reactor normal operating conditions and to explore fuel performance at temperatures substantially beyond those typical of normal operation. A total of 194 UCO fuel compacts in five capsules were irradiated in the Advanced Test Reactor for 360.9 effective full-power days, achieving final burnup ranging from 5.66% to 15.26% fissions per initial heavy metal atom and fast neutron fluence ranging from 1.62 × 10 25 to 5.55 × 10 25 n/m 2 (E > 0.18 MeV). Calculated time-averaged fuel temperatures ranged from 467 °C to 1432 °C, with a peak fuel temperature of 1536 °C. During the first five irradiation cycles (∼180 effective full-power days), 85m Kr fission gas release-rate-to-birth-rate ratios were 10 −7 –10 −6 , indicating no particle failures. Near the end of the sixth cycle, a substantial number of in-pile particle failures occurred in Capsule 1. The fission gas release from this capsule impacted the readings from the other four capsules and led to unreliable fission gas release measurements for all capsules during the last four cycles. A preliminary post-irradiation examination of Capsule 1 fuel and internal components revealed the in-pile particle failures resulted from operational issues with the capsule, not subpar performance of the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Cylindricity Sensitivity Thermal Model of the AGR-5/6/7 Experiment in the Advanced Test Reactor

The AGR-5/6/7 experiment is currently being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory and is approximately 70% complete. Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train, AGR-5/6/7 is testing low-enriched uranium oxycarbide TRISO fuel. The AGR-5/6/7 test train has 5 capsules with thermocouples and independent gas control mixtures. Unique to this paper is a sensitivity study concerning the cylindricity of the graphite holders containing the fuel compacts and their eccentricity in relation to the stainless-steel capsule walls. Each capsule has small nubs on the outside used for centering the graphite holder inside the stainless-steel capsule with a small gas gap used to control temperature. Due to machining tolerances of these nubs and vibration wearing the nubs down when the experiment is running in the reactor, the possibility exists that the holder may move around radially. Each capsule is equipped with several thermocouples placed at various radii and depths within each graphite holder. This paper will show the sensitivity of offsetting the graphite holder for various radii in 45-degree increments around the circle with the objective of minimizing the difference between the measured thermocouples and the modeled thermocouple temperatures. Separate gas mixtures of helium/neon are introduced into this gas gap between the holder and capsule wall and changed as necessary to maintain the desired thermocouple temperatures to keep the fuel compacts at constant temperature as the nuclear reactor conditions change. The goal of the sensitivity study is to find a radius and an angle to offset the holder from perfectly centered for each of the five capsules separately. The complex thermal model includes fission heating, gamma heating, radiation heat transfer, and heat transfer via conduction and radiation across the control gaps. Subroutines linked to the thermal model offer an easy method to offset the graphite holder from the capsule walls without remeshing the entire model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan details the strategy, mission, scope, and goals—both near-term and long-term—along with the structure and organization of nuclear fuels and materials research, development, and demonstration (RD&D) activities within the Fuel Cycle Technologies (FCT) program. The FCT program, tasked by the U.S. Department of Energy (DOE), employs a science-based approach to advance fuel technologies. This approach integrates theory, experiments, and multi-scale modeling and simulation (M&S) to develop a predictive understanding of fuel fabrication processes and fuel/cladding performance under irradiation, moving beyond traditional empirical methods. The long-term goals of the AFC are guided by the AFC Strategic Plan and align with the DOE Office of Nuclear Energy (NE) Roadmap [1], which outlines a multi-decade vision for demonstrating and qualifying advanced fuel forms to support diverse fuel cycle options. Near-term goals focus on enhancing accident tolerant fuels (ATF) for Light Water Reactors (LWR), a significant challenge that demands balancing immediate objectives with ongoing progress toward advanced reactor missions. Accelerating the traditional fuel qualification process to meet ATF objectives is another critical challenge. A detailed set of 5-year goals, summarized below, has been developed in line with the overarching science-based fuel development approach: • Advanced LWR Fuel Technologies: By 2027, support the development of advanced LWR fuel technologies with improved performance and enhanced accident tolerance. This includes high burnup (HBu), low enriched uranium (LEU)+, coated cladding, and doped fuel, aimed at complementing industry-led significant LWR uprates and plant refurbishments. • Tristructural Isotropic (TRISO) Fuel: Achieve qualification by 2028 and develop improved designs for emerging markets. • Metal Fuel: Achieve qualification by 2028 and develop improved designs for emerging markets. • Molten Salt Fuel: By 2027, deploy a robust program that enables fuel salt qualification technologies needed to support fuel salt research and development (R&D), focusing on emergent needs to derisk fuel salt production and utilization in advanced reactors. • Long-Term ATF: Develop fuel technologies that enable significant power uprates (~50%) in refurbished or new LWRs while optimizing fissile material utilization and waste disposal. The 5-year milestones in the AFC Execution Plan are contingent on an assumed budget. This Execution Plan will be updated annually to reflect actual funding profiles as budget guidance becomes available, ensuring milestones are adjusted accordingly. In summary, the AFC Execution Plan presents a comprehensive strategy to advance nuclear fuel technologies through a science-based approach, addressing both near-term and long-term goals while adapting to funding realities.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-2 Compact 6-4-1 Post-Irradiation Examination Results

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2017). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗