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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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Characterization of Legacy Fuels: Progress to Fuel Cladding Chemical Interaction Presentation for Metal fuel workshop 14-16 may 2024
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.
AGR Program activities for FY24 and future, including DOE Organizational changes impacting AGR Program and future coated particle fuel development.
AGR-5/6/7 goals and test train configuration, status of AGR-5/6/7 mass balance work, preliminary results of Ag-110m, Cs-134, and Eu-154.
AGR-3/4 goals and design, experiment method and sample selections and challenges, results of as-irradiated RDLBL and as-irradiated vs. FACS-tested, and conlusion.
This study focuses on the pre-transient characterization of U-10Zr test and sibling fuel pins for the THOR-M-LOF test series. Using neutron radiography, element contact profilometry (ECP), precise gamma scan (PGS), and gas assay, sampling, and recharge (GASR) analysis, it was confirmed that the fuel pins were intact and suitable for testing. Key fuel behaviors quantified include axial elongation, diametral strain, fluff structure geometry, axial isotope distribution, and fission gas release. Any deviations from historically expected behaviors were investigated and attributed to factors other than the irradiation behavior of the fuel pin. These pre-transient measurements establish a baseline for future post-transient analysis, which will be used to inform fuel performance models and safety criteria for sodium-cooled fast reactors (SFRs). The results will enhance understanding of transient fuel behavior and expand limited data on LOF scenarios.
Micro-tensile characterization of AGR-2 TRISO Particle Layers and Interlayer Regions, including experimental motivation, results, irradiated fuel kernel APT, Buffer, SiC omparison, conclusion, and future work.
Thermophysical Property Characterization of Irradiated TRISO Compacts to include motivation, experiment background, TCM Experimental Method, TCM Results, Summary, and future work.
This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.
Thermophysical Property Characterization of Irradiated TRISO Compacts to include motivation, experiment background, TCM Experimental Method, TCM Results, Summary, and future work.
Presentation on TRISO Fuel development and qualification status to be given at the Advanced Reactor Technologies (ART) annual program review.
Graphite overview: Material science, supply chain facts, nuclear material science, licensing thoughts
This is the final project poster for external release.