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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.

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At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Fabrication of UN microsphere kernels for Fully Ceramic Microencapsulated fuel

Fully Ceramic Microencapsulated (FCM) fuel is of interest to the U.S. Department of Energy's Fuel Cycle Research and Development program as an Accident Tolerant Fuel (ATF). The FCM design uses a TRISO (Tri-Structural Isotropic) particle with a ZrO{sub 2} surrogate kernel, embedded in a dense SiC matrix. TRISO technology, originally developed for High Temperature Gas Reactors, has an established high level of performance retaining fission products and the SiC matrix of an FCM compact shows good steam oxidation behavior. This work reports on a fabrication route for producing dense UN microspheres used as the kernel of the TRISO particles in FCM fuel. The UN microspheres are of high purity and density with enough porosity to accommodate volatile fission products. Increasing the density is nevertheless of interest potentially allowing for an additional adjustable property for optimizing the fuel kernel and TRISO particle characteristics. Current work is focused on investigating the effect of a feedstock microstructure on final product density. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Irradiated AGR 5/6/7 Compact 5-6-2 Examination Plan

This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 5-6-2 taken from the Advanced Gas Reactor (AGR) experiment, AGR 5/6/7. This work will be performed in accordance with the general objectives outlined in the AGR 5/6/7 PIE Plan and guidance in the ORNL PIE Statement of Work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial examination of fuel compacts and triso particles from the US AGR-2 irradiation test - 18443

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test. These compacts were selected for examination due to indications that they may have contained particles that released cesium due to a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzer (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using x-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and x-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Electron probe microanalysis of irradiated and 1600 deg. C safety-tested AGR-1 Triso fuel particles with low and high retained {sup 110M}Ag - 18570

Advanced Gas Reactor (AGR)-1 fuel Compact 4-3-3 achieved 18.63% FIMA and was exposed to a safety test at 1600 deg. C. Two particles, AGR1-433-003 and AGR1-433-007, with measured-to-calculated {sup 110m}Ag inventories of <22% and 100%, respectively, were selected for electron microprobe analysis to determine whether the distribution or abundance of fission products differed proximally and distally from the deformed kernel in AGR1-433-003 and how this compared to fission product distribution in AGR1-433-007. On the deformed side of AGR1-433-003, xenon, cesium, iodine, europium, strontium, and tellurium concentrations in the kernel buffer interface near the protruded kernel were up to six times higher than on the opposite, non-deformed side. At the silicon carbide (SiC)-inner pyrolytic carbon (IPyC) interface proximal to the deformed kernel, palladium and silver concentrations were 1.2 and 0.04 wt% respectively. On the SiC-IPyC interface distal from the kernel deformation, those elements measured 0.4 and 0.01 wt%, respectively. Palladium and silver concentrations at the SiC-IPyC interface of AGR1-433-007 were 2.05 and 0.05 wt.%, respectively. Rare earth element concentrations at the SiC-IPyC interface of AGR1-433-007 were ten times higher than at the SiC-IPyC interfaces measured in AGR1-433-003. Palladium permeated the SiC layer of AGR1-433-007 and the SiC layer of the non-deformed side of AGR1-433-003. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Initial results from safety testing of US AGR-2 irradiation test fuel - 18574

Two cylindrical compacts containing tri-structural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO{sub 2}-kernel TRISO particles have undergone 1600 deg. C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ({sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, {sup 155}Eu, {sup 90}Sr, and {sup 85}Kr) were monitored while heating the fuel specimens to 1600 deg. C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600 deg. C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO{sub 2} compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO{sub 2} compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically similar to the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600 deg. C. However, additional silver release was observed later in the safety testing due to the UO{sub 2} TRISO with failed SiC. Failure of the SiC layer in the UO{sub 2} fuel may have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Key results from irradiation and post-irradiation examination of AGR-1 UCO Triso fuel - 18576

The AGR-1 irradiation experiment was performed as the first test of tri-structural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 x 10{sup 5} coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136 deg. C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 deg. C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment - 18579

The PARFUME (Particle Fuel Model) code was used to predict silver release from tri-structural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO{sub 2}) kernels. Post-irradiation examination (PIE) measurements were performed to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO{sub 2} compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation - 18594

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Pressure Loading with In-Situ Permeability Measurement of SiC Ceramic Matrix Composite Tube

Silicon carbide fiber in a silicon carbide matrix composite (SiC-SiC) has been actively researched as a potential structural material to be used in advanced high temperature gas reactors (HTGR) and as an accident tolerant cladding material for light water reactors (LWR). This ceramic composite material demonstrates desirable properties for use in nuclear reactor core that include retention of strength from ambient to LWR and HTGR operating and accident temperatures and stability under neutron radiation. However, the ceramic composite material must be designed to meet all the necessary, application driven requirements before it could be implemented in a reactor design. As one of the requirements for a fuel cladding material, the composite has to maintain high level of impermeability as to retain fission gases throughout the fuel cycle. General Atomics (GA) had previously reported results on permeability levels obtained by CVI produced SiC-SiC composites. It has been determined that for LWR size cladding tube that is 51 mm long the permeability allocation equivalent to currently used claddings is 7.3 E-10 atm*cc/sec. With high level of attention to possible sources of leaks in a test setup, it has been demonstrated that a CVI SiC-SiC composites produced by GA can achieve a leak rate of 1 E-12 atm*cc/sec, and therefore it meets the permeability allocation for a fuel cladding. Further study showed that a particular SiC-SiC composite tube can remain impermeable after being pressure loaded well above LWR normal operating conditions. In the study done by Deck et al., hoop stress was applied to a composite tube using pressure load on the inside surface in a load-unload stepwise sequence with increasing maximum applied stress at each step and measuring permeability between steps. The tested composite comprised of 70% thickness of SiC-SiC composite which provides fracture toughness and bears a majority of the applied stress and a 30% thickness of monolithic SiC layer on the outside surface to provide an impermeable barrier. The composite remained impermeable up to internal pressure of 56 MPa which is more than 2.5 times greater than a typical pressure at the end of life in a fuel rod.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Alloy 709 Advanced Austenitic Stainless Steel

Provide material solutions to enabling design, construction, licensing and operation of advanced reactors Including Fast Reactors, Gas-cooled Reactors and Molten Salt Reactors (solid or liquid fuel) Could be of modular or micro designs, and from hundreds of MWe to kWe Provide technical bases needed for design and licensing of advanced reactor components Develop & validate improved high temperature design methodology Provide qualification data (to NQA-1 or equivalent) on structural materials Utilize consensus standards when appropriate (e.g., ASME, ASTM, etc.) Target resolution of issues needed for near to mid-term deployment of advanced reactors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Post-irradiation Heating Tests of As-Irradiated AGR-3/4 TRISO Fuel Compacts

Four post-irradiation heating tests of fuel compacts from the U.S. Advanced Gas Reactor (AGR)-3/4 irradiation experiment were completed. In addition to tristructural isotropic (TRISO)-coated driver fuel, each compact contained designed-to-fail (DTF) particles with fuel kernels coated only in pyrocarbon so as to simulate exposed kernels. Tests at 1600/1700°C, 1400°C, and 1200°C were performed to measure fission product release as a function of time and temperature. Silver release was highest in the 1200°C test, supporting the observation that silver release rates are highest in the 1100–1300°C range. Compared to tests of AGR-1 compacts with no exposed kernels, the Cs-134 and Kr-85 releases were noticeably higher in AGR-3/4. The exposed kernels’ contributions to Eu and Sr release are inconclusive, due to the difficulty in distinguishing among the combined effects of higher irradiation temperatures in these particular AGR-3/4 compacts, the presence of the DTF particles, and the Fuel Accident Condition Simulator (FACS) test temperatures. These data can be used to make inferences about fission product retention in exposed kernels as a function of time and temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Generating An Advanced Cross-section Library For HTGR Pebble Bed Depletion Calculations Using Reduced-Order Model Generation Techniques

For code development, Advanced Reactor Technologies - Gas Cooled Reactors Program (ART-GCR) rely on a collaboration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, but the cross sections generation and the methodology definition is part of this program area goals. Based on previous studies in FY23, the size of microscopic cross section libraries increases rapidly with the number of tabulations, requiring significant amount of memory and drastically slowing down the Griffin calculations when evaluating cross sections via the multivariate linear interpolation approach. Rising to these challenges, this work investigates constructing Reduced-order Models (ROMs) for the multi-group microscopic cross sections to accelerate the cross section evaluation in Griffin. A database of multigroup cross sections is first collected considering all possible parameters that a designer could change for optimization. Down-selection of the ROM techniques afterward shows Deep Neural Network (DNN) as the best candidate when jointly consider memory efficiency, predictive accuracy, computational cost, scalability, flexibility and ease of implementation of the algorithms in comparison to the multidimensional interpolation. This work develops a specific interface that enables the cross section predictions using pre-trained DNN models into Griffin leveraging the existing ROM capabilities. DNNs have been trained for all isotopes for use in Griffin. Preliminary Griffin testing shows that DNNs exhibit exceptional predictive accuracy and the use of DNNs provides orders of magnitude improvement in memory efficiency compared to conventional interpolation techniques. With such ROM techniques, it holds great promise to further increase the fidelity of the Pebble Bed Reactor (PBR) simulation by increasing the number of tabulations/state variables during cross section evaluation, while maintaining the computational cost affordable in Griffin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Ceramography and Thermal Modelling of Irradiated TRISO Particles using BISON

This study presents a comprehensive analysis of irradiated TRISO fuel particles from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program conducted at the Advanced Test Reactor (ATR) as part of my internship. Failure Analysis using high-resolution microscopy was performed to characterize mechanisms in TRISO particle layers, including the kernel, buffer, inner pyrolytic carbon (IPyC), silicon carbide (SiC), and outer pyrolytic carbon (OPyC) coatings. Failure modes were systematically classified. Thermal conductivity properties of each TRISO layer were evaluated through experimental measurements and computational models for use in BISON finite element code. Experimental thermal conductivity values ranged from 0.5 W/m·K for the buffer layer to 168 W/m·K for SiC, while BISON simulations provided averaged values of 0.57 W/m·K (buffer), 4.0 W/m·K (PyC layers), and 15.95 W/m·K (SiC). Temperature-dependent correlations from PARFUME were implemented for kernel thermal properties.

BISON↗
DOE OSTI2025

A Machine Learning Approach Toward Improving QA/QC of Coated Particle Fuels

Coated particle fuels, such as the tristructural isotropic (TRISO) fuel particle, have been studied and developed for decades for high-temperature gas reactor (HTGR) applications because of their efficiency and relative stability under off-normal conditions. Critical to this development is a strong understanding of the relationship between fuel fabrication, properties and performance, as is how the former can be adjusted and improved to optimize the latter. Accordingly, wide-scale implementation of coated nuclear particle fuels requires thorough and robust quality assurance/quality control (QA/QC) methods for fabrication, characterization, and deployment

Conry, Bryan [ORNL] (ORCID:0000000236089604)↗
DOE OSTI2025

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↗
DOE OSTI2025

Fission Product Quantification and Burnup Analysis via Gamma Spectrometry of the AGR-5/6/7 TRISO Fuel Experiment

Five capsules containing varying configurations of TRISO fuel compacts were irradiated in the most recent Advanced Gas Reactor (AGR) TRISO test campaign, AGR-5/6/7. As part of the post-irradiation examination (PIE) of this work all the fuel compacts and capsule graphite holders were measured with the Precision Gamma Scanner (PGS) to determine isotopic activities and locations using gamma spectrometry. Measuring the compacts allowed for the comparison of physics model depletion calculations with the measured compact activity to validate models. These measurements also are used to find compacts of interest with low fission product retention that may be tested in subsequent PIE. Measurements of the holders can confirm fission product distribution and migration. Of the isotopes measured, Cs-137 and Cs-134 were used to estimate location of failed particles as well as determine a measured burnup estimation. Results from Capsule 1 compacts and the holder indicate many fuel particle failures occurred. There are high amounts of cesium in the holder near Stack 9 of Capsule 1 that corresponds to low fractions?measured-to-calculated (M/C) ratio?of cesium remaining in those compacts. All other capsules had cesium retention ratios near 1. The isotope 110mAg had less consistent results. Capsule 2 compacts exhibited a M/C of 0.20, much lower than anticipated for the temperatures these compacts underwent. Capsule 3 experienced much higher temperatures, >1300C and had an average M/C of 0.75. These are the opposite results than what was predicted by past AGR experiments for the relationship between temperature and isotope retention. Capsule 4 and 5 showed good fission product retention and agreement with the physics model predictions. The burnup results from the PGS measurements showed strong agreement between the physics models and the measured values. Using a ratio of Cs-134/Cs-137 to calculate the burnup match closely with the physics models. Additionally, the Cs-137 burnup results were less accurate but extremely close to the model results.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Air-Cooled RCCS CFD Modeling Validation 2025

Slide deck for the Advanced Reactor Technologies Gas Cooled Reactor (ART-GCR) program review. The slides show the progress on the air-cooled reactor cavity cooling system (RCCS) CFD validation work. The validation of the air-cooled RCCS is performed using the experimental facility at the University of Wisconsin-Madison. The slide deck provides a progress update on this year's achievements. The natural convection tests under uniform power are modeled and compared with the experimental results. The results show good agreement with the experimental results. A sensitivity study on the RANS turbulence models is performed. Additionally, the contribution of radiative and convective heat transfer within the heated cavity is calculated and compared to forced convection setups.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗