X-ray computed tomography of deconsolidated TRISO Particles from the AGR-5/6/7 irradiation experiment capsule 1 Compact
Not Available
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
SEARCH · Engineering Papers
Results for “AGR-5/6/7”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Thermal Model of the AGR-5/6/7 Experiment with Offset Gas Gaps
Not Available
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
AGR-5/6/7 PIE Overview 2024
This is a status update and description of major post-irradiation examination activities under the DOE's AGR TRISO fuel program. This presentation will be given at the annual program review hosted by INL the week of July 15.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
AGR-5/6/7 CAPSULE 1 THERMAL MODEL WITH OFFSET GAS GAPS
Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO) fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a 3 D finite-element thermal model that was subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty caused by the graphite holder nub-to-capsule clearance. This thermal model was revised to examine the most probable graphite offset position for Capsule 1 for four different days during the irradiation. In the analysis, the offset distance and azimuthal direction at both the top and bottom of the holder were varied. The best-fit offset was estimated based on the minimum root mean square error (RMSE) of the residuals (measured minus calculated thermocouple [TC] temperatures) for the operational TCs in Capsule 1. Near the end of the irradiation, when the graphite had shrunk and the gas gaps had grown, a peak temperature was found with maximum offset along the TC lines, with nickel introducing a potential mechanism for fuel particle failures.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Initial Results from Post Irradiation Examination of AGR-5/6/7 TRISO Fuel
Explore the source record for details and available documents.
Cureton, Will [ORNL] (ORCID:0000000319764150)↗
AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps
Fuel compact temperatures are a crucial factor in assessing irradiation performance of the Tristructural isotropic fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a three-dimensional finite element thermal model, which is subject to simulation uncertainty. The most dominant factor in uncertainty of calculated fuel temperature is the gas gap uncertainty due to the nub-to-shell clearance because of fabrication error. A revision to the thermal model was made to examine the most probable offset position for four different dates of interest. The offset position options varied in magnitude and azimuthal direction of offset of the holder top and bottom position. The best-fit offset of the holder is estimated based on the minimum root mean square error of residuals for all operational thermocouples (TCs). From these results, the following conclusions were made: 1) During earlier cycles (162A – 164A) when numerous TCs were still operational, the best-fit offset distance varied in range [0.002 – 0.0035 in.] for both the top and bottom holder while offset azimuthal direction varied widely, especially for the holder bottom. 2) Holder offset led to slightly lower Capsule 1 average temperature, but wider temperature variation (lower minimum and higher peak fuel temperatures). 3) The maximum offset of 0.006 in. to the northwest direction for both the top and bottom holder yielded a much higher temperature than the original peak fuel temperature, which increased from 1422 to 1557 ? (or 135 ? increase). This could be one of the reasons for massive particle failure near the end of Cycle 165A. 4) Even though, the highest temperature at the TC-1-7 tip reached slightly higher than 1000?, the temperature along the TC wire reached as high as 1335? assuming a holder offset of 0.006 in. to the northwest. Consequently, at this temperature nickel from this TC wire could cause particle failures in Capsule 1.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
AGR-1, AGR-2, AGR-3/4, and AGR-5/6/7 DimensionalChange Analysis
A series of fuel irradiation experiments have been planned in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to support the licensing and operation of the Advanced Reactor Technologies high temperature gas-cooled reactor. The advanced gas reactor (AGR) experiments are comprised of multiple independent capsules containing multiple cylindrical fuel compacts, placed inside of a graphite cylinder in ATR. The purpose of the AGR experiments is to provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for accident testing, and support the development of fuel performance and fission product transport models. The advanced graphite creep (AGC) experiments provide irradiation creep data for design and licensing. To date, six irradiation campaigns have been completed: AGR-1 (December, 2006 – November, 2009); AGR-2 (June, 2010 – October, 2013); AGR-3/4 (December, 2011 – April, 2014); AGC-1 (September, 2009 – January, 2011); AGC-2 (April, 2011 – May, 2012); and AGC-3 (November, 2012 – April, 2014).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
Future TRISO fuel irradiations in the Advanced Test Reactor
AGR-5/6/7 was thought to be the last experiment in the AGR series and the last experiment directly sponsored by DOE. Over the last five years three developments have caused a reevaluation of that assumption. First is that interest in TRISO fueled reactors has spiked dramatically. Virtually all of the new reactor designs are for SMRs or micro-reactors . Some of these reactors use new fuel designs with higher concentrations of U235. This results in changes to the particle design and even fuel pellet design. Second, a design flaw in AGR-5/6/7 Capsule 1 produced a large number of particle failures which resulted in the loss of data from Capsule 1. An additional irradiation could fill in this lost data set. Thirdly, many reactor developers have decided to use AGR spec fuel but different irradiation goals have been identified.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
SILICON CARBIDE THERMOMETRY USING RAMAN SPECTROSCOPY ON IRRADIATED TRISO PARTICLES
Silicon carbide (SiC) passive thermometry has emerged as a promising post-irradiation examination (PIE) technique for estimating irradiation temperature near the end of irradiation. While dilatometry techniques have been traditionally used to analyze prismatic samples after irradiation, Raman spectroscopy has recently been shown to provide comparable results by analyzing Raman-active phonon modes. In this work, Raman spectroscopy has been applied to the SiC layer of cross-sectioned irradiation tristructural isotropic (TRISO) particles from the AGR-5/6/7 experiment to evaluate the feasibility of particle-scale passive thermometry. Two-dimensional Raman mapping was used to measure the position of the SiC longitudinal optical (LO) phonon, which was then converted to an apparent irradiation temperature using a previously established empirical correlation. Particles from AGR-5/6/7 Compacts 2-2-1 and 5-1-3 were selected as their calculated time-averaged, volume-averaged (TAVA) temperatures (828°C and 706°C, respectively) fall within the range of the sensitivity of the experimental approach. The use of SiC thermometry was anticipated to confirm or highlight potential deviations from calculated end-of-life TAVA temperature across compacts. Four particles from Compact 2-2-1 and two particles from Compact 5-1-3 were selected based on 110mAg inventory (either some measurable activity or below the minimum detection limit) which is commonly used as an indicator of in-pile temperature variation of particles within the same compact. Across every particle selected it was determined that the average LO peak position was located around 968 cm-1 to 969 cm-1. Using the previously determined empirical correlation, this corresponds to an irradiation temperature around 950°C to 966°C, which does not align with the reported TAVA temperature values. This discrepancy in apparent irradiation temperatures likely reflects a combination of uncertainties in the calculated particle temperatures and differences in irradiation history between the present specimens and those used to establish the empirical Raman calibration such as neutron flux (damage rate) and SiC microstructure (as fabricated and irradiated).
Vawdrey, Josh [ORNL]↗
Deconsolidation and Leach Burn Leach of Seven As Irradiated AGR 5/6/7 TRISO Fuel Compacts from Capsules 2, 3, 4, and 5
Seven as-irradiated Advanced Gas Reactor (AGR) 5/6/7 compacts underwent destructive post-irradiation examination via deconsolidation-leach-burn leach at Idaho National Laboratory (INL). The selection of the compacts extended the upper and lower limits of time-average volume-average (TAVA) temperature for compacts that had gone through deconsolidation-leach-burn leach so far. The measured inventories of fission products and actinides in the compact matrix and outer pyrolytic carbon were reported. Results indicated unexpectedly higher rates of fuel kernel leaching compared to compacts from AGR-1 and AGR-2. These failure rates were attributed to damage during post-irradiation sample handling, rather than irradiation itself. AGR-5/6/7 compacts have little or no matrix coverage for some particles at the top and bottom ends of cylindrical fuel compacts, making them more fragile. Sixty particles were randomly sampled from each compact, and the gamma results were reported. Three SiC shells from were identified, one of which showed signs of chemical attack in the high-irradiation temperature compact.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Strategy to safely enable X-ray computed tomography examination of highly radioactive tristructural isotropic nuclear fuel
Nondestructive post-irradiation examination of nuclear fuels and materials is useful in collecting data to inform commercial licensing of new nuclear concepts. This work details the planning, shielding design, and workflow of an X-ray computed tomography examination of an irradiated AGR-5/6/7 fuel compact with a dose rate of 1318 R/hr on contact from both β and ɤ-ray radiation with 120 R/hr of the dose rate coming exclusively from ɤ-rays. Post-irradiation examination of highly radioactive samples can help reduce the timeframe from conceptualization of novel nuclear concepts to their commercial implementation as not only is less time spent waiting for experiments to decay away to acceptable levels for examinations, but data from these experiments can more quickly inform model efforts and subsequent experiments. While this work represents the hottest radiological sample that has been openly transported and examined at Idaho National Laboratory’s Irradiated Materials Characterization Laboratory to date, it is anticipated that this work and its lessons learned will facilitate future examinations of similar or even more radioactive specimens.
42 - ENGINEERING↗
BISON analyses of TRISO fuel performance, its dependence on time-at-temperature, and possible implications for fuel design and qualification
The Advanced Gas Reactor Fuel Development and Qualification (AGR) program has established a substantial technical foundation to support private entry into the U.S. high-temperature gas-cooled reactor market. However, emerging tristructural isotropic (TRISO)-fueled reactor applications include small modular reactors and microreactors with longer fuel residence times, which may expose fuels to higher time-at-temperature (TAT) values than were explored by the AGR program. Increased TAT could affect diffusive and thermomechanical behaviors such as Pd penetration, fission gas release, creep, and fission product transport. In this work, we applied multiscale best-estimate BISON fuel performance modeling to assess these effects within a representative design space based on the AGR-5/6/7 experiment and analyzed trends in predicted particle and compact fuel performance metrics with possible implications for near-term fuel design and qualification. BISON unambiguously predicted that TRISO fuel performance is sensitive to TAT. Increasing TAT was not predicted to increase the magnitude of failure-inducing tangential stresses in particle coating layers. Predictions obtained using a mechanistic model for Pd penetration indicated that penetration depth does not depend strongly on TAT. While these observations suggest that AGR testing provides a conservative upper bound for the steady-state operation of TRISO particles at lower powers and higher residence times, BISON also predicted that the release of poorly retained Ag would increase with TAT. Because these analyses applied models to extrapolate beyond the available experimental data, the authors recommend performing targeted experiments to confirm these predictions. Nevertheless, these predictions may provide reactor developers with enough confidence to make near-term design decisions associated with the potential fuel performance trade-offs of increasing TAT.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Analysis of Silver Release from Furnace-Tested TRISO Particles
High temperature testing of intact TRISO particles previously irradiated in the AGR-5/6/7 experiment was performed in the Furnace for Irradiated TRISO Testing (FITT) to directly confirm diffusive release of silver and europium from intact TRISO particles. Testing was conducted from 1,100–1,600°C for exposure times up to 100 h to directly confirm silver through layer release below safety testing temperatures. The FITT analysis showed highest levels of silver release at 1,300–1,400°C which confirms athermal release behaviors previous observed in step-wise and transient safety tests. Additionally, release was non-uniform with some particles releasing a majority of their inventory while others did not appear to release silver under identical testing conditions, which was consistent with historic observations. An assessment of the effective silver diffusion coefficient in the SiC layer was conducted and indicated maximum values in the 1,300–1,400°C range. The magnitude of the calculated diffusion coefficients also exceeded currently accepted diffusion coefficients. The release behavior of europium from intact particles was also analyzed in FITT for at 1,450–1,550°C for 500 h to 750 h. Direct confirmation of europium release below safety testing temperatures was confirmed absent contributions from matrix release. The analysis indicated europium release follows a general Arrhenius behavior and suggests general uniform release and indicates irradiation conditions influence observed release response. Calculated diffusion kinetics agree well with historic experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview
AGR-3/4 post-irradiation examination and data analysis AGR-5/6/7 PIE and safety testing Supplemental fuel microanalysis and method development Fuel oxidation testing Air/moisture Ingress Experiment (AMIX) system development (deployed in FY23) Single particle testing in FITT Data management and analysis Fuel performance modeling
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 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 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↗