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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

AGR-2 TRISO Layer Thickness Imaging Archive

As a part of fuel quality control characterization, optical microscopy images of particle cross sections near midplane were acquired at Oak Ridge National Laboratory (ORNL). These particles were produced by the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-2 irradiation campaign. These images may be of use for the development of image processing algorithms with the benchmark values measured at ORNL. This report provides those benchmark values, along with the raw images and data generated by the ORNL particle layer thickness analysis process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

AGR-2 TRISO Shadow Imaging Archive

As a part of fuel quality control characterization, backlit shadow images of particles produced by the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-2 irradiation campaign were acquired at Oak Ridge National Laboratory (ORNL). These images may be of use for the development of image processing algorithms with the benchmark values measured at ORNL. This report provides those benchmark values, along with the raw images and data generated by the ORNL shadow imaging process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Estimation of Fission Product Transport Parameters for Cesium in the AGR-3/4 TRISO Fuel Experiment

A one-dimensional (1D) finite-element model of fission product transport in the AGR-3/4 experiment has been developed using the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and implemented in the fuel performance code, BISON. The model resolves capsule-specific geometries, materials, and temperature histories and simulates radial migration of fission products from the fuel compact through the inner ring, outer ring, and into the sink ring. Model parameters governing diffusion and sorption were estimated for key fission products – cesium (Cs), and europium (Eu) – by simultaneously fitting modeled isotopic concentration profiles and total ring inventories to a post-irradiation experimental measurement. These data include gamma scanning, liquid scintillation for Sr-90, radial deconsolidation leach-burn-leach analysis, tomographic reconstructions, and destructive physical sampling. A mortar-based interfacial sorption framework was implemented to enforce physically consistent mass transfer and flux conservation across gas gaps. Two classes of parameter sets were derived: a least-squares best-fit, and a safety-oriented conservative-fit, what applies strong penalties for underprediction of sink inventories. Across all twelve capsules, the model successfully reproduces the dominant radial transport trends for Cs, Sr, with decreasing concentrations from the compact outward through successive rings. Cs behavior is captured most consistently, while strontium predictions reveal systematic trade-offs between compact accuracy and conservative sink-ring bounding. The results demonstrate that sink ring weighted calibration provides conservative, safety relevant bounds on low temperature fission product transport, but at the cost of underpredicting compact inventories for Sr isotopes. These discrepancies highlight the need for additional physics, including fast-slow diffusion model, incorporating trapping mechanism in the transport behavior. Overall, this work establishes a robust, capsule-specific modeling framework for AGR-3/4 fission product transport and provides a defensible basis for parameter selection in source-term and fuel performance analyses for high temperature gas-cooled reactors.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

AGR Program TRISO Fuel Topical Report Dialog

AGR Program Overview and Current Activities Fission Product Transport Data Short-Lived Fission Product Release Fuel Performance Margin Data Oxidation Testing Data

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Select Proliferation Studies on TRISO-fueled, Heat-Pipe-cooled Microreactors

Nuclear microreactors carry the potential to open up new markets for the nuclear industry, as their expected cost competitiveness in non-traditional market segments (e.g., mines, military bases, extraterrestrial surfaces, and remote areas), and their inherent safety features make them deployable when other power sources are unavailable or difficult to exploit. For countries that have not traditionally participated in nuclear power, microreactors represent a clean energy solution [1]. However, their use, especially in non-weapons states, may entail challenges in terms of maintaining international nuclear safeguards [2]. Furthermore, the deployment locations where microreactors may prove most cost competitive would be difficult to access by state and International Atomic Energy Agency (IAEA) inspectors [3]. In addition to the isolated nature of potential deployment sites, the low-power characteristic of microreactors suggests that numerous microreactors would need to be deployed to meet energy demands. That, coupled with the unique physics of many current microreactor designs, opens up a new area of research with respect to nonproliferation and safeguards concerns [2]. Whereas traditional facilities are inspected as isolated cases when looking for signs of diversion or misuse; microreactors may need to be assessed in the context of the entire fleet to which they belong. International safeguards necessitate timely detection of any significant quantities (SQs) of material that are being diverted (e.g., 1 SQ of special nuclear material diverted over the course of a 1-year period) [4]. For low-enriched uranium, the IAEA defines 1 SQ as corresponding to 75 kg of 235U. The purpose of the present paper is to explore the detectability threshold for material diversion in microreactors by relying on critical control drum angles, excess reactivity, and the reactor lifetime as the selected operational parameters. For this assessment, a heat-pipe-cooled microreactor was regarded as the base design. While the conclusions reached in this paper are not readily extendable to the design of actual microreactors, the analysis herein enables conclusions to be drawn regarding the level of accuracy needed for reference calculations in order to detect diversion scenarios by utilizing the selected operational parameters (i.e., mainly control drum angles, critical insertion angle, and the reactor lifetime).

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗