10 Search Results

The silicon carbide (SiC) coating in a tristructural isotropic (TRISO) particle acts as a barrier to fission product release during reactor operation and accident scenarios. Oxidation and subsequent failure of the SiC layer during a rare air ingress event is a proposed mechanism for fission product release in a high-temperature gas-cooled reactor (HTGR). Although previous oxidation studies have analyzed unirradiated TRISO particle response, this study compared the oxidation behavior of irradiated and unirradiated TRISO particles from the second Advanced Gas Reactor Fuel Development and Qualification Program irradiation experiment (AGR-2). Particles with exposed SiC were subjected to six varying oxidizing testsmore » in the Furnace for Irradiated TRISO Testing (FITT), examined for failure fraction with the Irradiated Microsphere Gamma Analyzer (IMGA) and characterized with focused ion beam and scanning/transmission electron microscopy techniques to analyze the oxide layer. Uncorrelated unirradiated particle failures throughout the series of exposures suggests that external factors inherent to the experiment increased particle failure sensitivity. However, irradiated particle observations indicated an increased failure response at 400 h 1400 °C in both 2% and 21% O₂ atmospheres above failure associated with external factors. Oxide thickness measurements after 400 h at 1400 °C revealed a greater oxidation rate than predicted by parabolic growth, which was attributed to the increased complexity of the oxide structure at longer exposure times. Altering the atmosphere from 21% to 2% O₂ reduced the average oxide thickness by approximately 12%–14% in both irradiated and unirradiated particles at 400 h 1400 °C. Altogether, the minor variations observed between irradiated and unirradiated particles in this study led to the conclusion that unirradiated TRISO particles can be used to approximate irradiated TRISO oxidation kinetics.« less

The Transformational Challenge Reactor (TCR) fuel form was designed to contain large, densely packed uranium carbonitride (UCN) tristructural-isotropic (TRISO) fuel particles within a 3D printed SiC matrix, increasing the uranium density compared to conventional TRISO fuel forms and offering full geometric freedom for core design. Here, this work summarizes initial low-burnup, high-power irradiation testing of TCR fuel materials, including loose UCN TRISO particles and integral fuel compacts with ~55% TRISO particles by volume, to evaluate fission gas retention. Fission gasses were fully retained in all loose particle tests and in integral compacts irradiated at low (<250 °C) surface temperatures. Initialmore » testing at higher (~700–750 °C) fuel surface temperatures showed fission gas release (FGR) and complete fracture of three compacts, but no FGR was observed in later high temperature tests (~300–750 °C) of both fueled compacts and loose TRISO particles. Calculated thermal stresses in the failed compacts were far less than the measured strength of the SiC matrix and the stresses in some failed compacts were less than those in compacts that did not show FGR. Thermal stress-induced matrix cracks also would not cause complete fracture because the tensile stresses transition to compression in the higher temperature regions. Therefore, fuel failure was likely not caused by thermal stresses and may have been related to leakage currents from the electrical heaters and erratic fuel surface temperatures that were only observed in the test for which failure was observed. In any case, the matrix cracks propagated through the coatings of TRISO particles located in the high-density matrix regions on the peripheries of the compacts, resulting in measurable fission gas release. The discussion focuses on the importance of understanding matrix density distributions and the particle-matrix interface properties to prevent matrix cracks from causing TRISO particle failures.« less

In this work, the morphology in the interface region between the inner pyrolytic carbon layer (IPyC) and silicon carbide (SiC) layers in tristructural-isotropic (TRISO) particle fuel from the AGR-2 irradiation experiment were studied using focused ion beam-scanning electron microscopy tomography. This work quantitatively described the interface and corresponding relevant metrics to understand how the microstructural features at the IPyC/SiC interface may influence actinide and fission product interactions with the SiC layer. Particles were selected with varied ^110mAg retention rates, and their volumes were reconstructed and analyzed for distributions of pores, fission product/actinide features, SiC, and IPyC. It was found thatmore » porosity accommodates fission products in the interface and SiC layers. The largest fission product/actinide precipitates were found in the interface region. This was also where the largest number fraction of fission products/actinides was found, consistent with SEM showing fission product/actinide pileup along selected areas of the interface region.« less

Here, hot-isostatic pressing has been applied to Zirconium plate fabricated using the ultrasonic additive manufacturing (UAM) technique to enhance interfacial bond quality. Specimens heated to 800 C for 1 h at 100 MPa pressure showed grain growth across many prior foil-to-foil interfaces, thereby increasing foil adhesion. In addition to the material softening induced by a loss of Hall-Petch strengthening, premature failure of specimens loaded parallel to the build direction was observed. Premature specimen failure was attributed to the local delamination at prior foil-to-foil boundaries where grain growth was pinned by Ti impurities introduced during the UAM process via the interactionmore » between a Ti-alloy buffer foil and the welded Zr-foils underneath. In addition, the presence of Ti along select foil interfaces resulted in the nucleation and growth of secondary (Zr,Ti)(Fe,Cr) laves phases during higher-temperature thermomechanical processing. Using a combination of micro-scale X-ray computed tomography, fractography, and in-situ digital image correlation, the effect of defects along prior foil-to-foil boundaries was revealed as delamination-assisted plasticity accelerated specimen failure in preferred tensile orientations. These findings underscore the importance of impurity control when optimizing weld quality of higher-strength material systems using ultrasonic welding.« less

Tristructural-isotropic (TRISO) fuel is being considered for multiple reactor designs due to its exceptional performance. This work studies pyrocarbon (PyC) and silicon carbide (SiC) substrates that are near-representative of the inner pyrocarbon (IPyC) and SiC in TRISO fuel. Focused ion beam (FIB)-scanning electron microscopy (SEM) tomography was used to understand the influence of different processing routes on the PyC/SiC interface structure. Additionally, this work outlines a methodology for characterizing the IPyC/SiC interface when considering fission product accommodation and retention in TRISO fuel design. The high-fidelity images collected with FIB-SEM tomography can be used to achieve a more in-depth interpretation ofmore » microstructural parameters and pore-shape analysis that can inform simulations and processing.« less

Ceramic-metallic nuclear fuels are a candidate fuel for nuclear thermal propulsion systems due to their high heat transport properties, which are necessary in very high-temperature environments. The conventional fabrication of uranium nitride–molybdenum fuel has been thoroughly studied in the past, but modern manufacturing techniques have presented a unique opportunity for further development within this field. This work demonstrates the use of advanced manufacturing techniques to produce nuclear fuel pellets composed of uranium nitride microspheres encased in a molybdenum matrix. Binder jetting is used to print molybdenum disks that are filled with uranium nitride microspheres and afterward sintered using spark plasmamore » sintering. Two fuel pellets were fabricated to demonstrate the methodology and to provide a baseline analysis of the effects of temperature and pressure processing conditions. Characterization of the sintered fuel pellets includes detailed microstructural analysis and thermal conductivity measurements.« less

Three distinct composite architectures of silicon carbide (SiC) and pyrocarbon (PyC) were incorporated into the SiC coating layer of tristructural-isotropic (TRISO) nuclear fuel particles. The composite architectures are meant to increase the resistance of SiC coating layer to cracking and fission product attack during operation and accident scenarios. All composite layers were produced using the existing fluidized bed chemical vapor deposition apparatus that is used for production of TRISO fuel particles without modifications. Detailed characterization of the composite microstructure was carried out via optical and electron microscopy. Nano-indentation examination confirms that mechanical properties of the SiC phase was not affectedmore » in the composite architectures, however, the resistance to crack propagation in this coating layer was greatly increased in all cases when compared to the reference monolithic coating layer. The stress required to debond the SiC-inner PyC interface in the reference TRISO particles was determined to be ~1 GPa using micropillar compression technique. The high strength may explain the ease of crack propagation from the inner PyC to SiC in the reference design. In the composite architectures, the means of crack deflection were effectively incorporated at this interface. Lastly, finite element analysis of stress evolution in the fuel particles during normal operation with the reference and composite SiC coating layer architectures did not show any significant differences between the variants.« less

Fast reactor fuel cladding candidate materials require proficiency in extreme environments consisting of high temperatures and irradiation doses in excess of 150 displacements per atom (dpa). Nanostructured oxide dispersion strengthened (ODS) alloys have been developed extensively for this purpose due to their notable high temperature strength, creep resistance, and irradiation resistance. However, their properties can deteriorate if interstitial impurities such as C and N are not well controlled during the fabrication process. A new Fe-12Cr nanostructured ODS alloy OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding) with solute additions of Mo, Ti, and Nb has been developed to provide themore » desired properties mentioned above while simultaneously sequestering impurities within the matrix. After extrusion at 850°C, the as-extruded microstructure consists of an average 490 nm grain size and a high number density (6.8 × 10²³ m^-3) of 2.2 nm diameter (Y,Ti,O) nanoclusters distributed homogeneously in the matrix. Atom probe tomography investigations suggest non-stochiometric compositions for the smallest nanoclusters. In addition, a second population of nanometer scale (Nb,Ti) rich carbonitrides is also present in the microstructure that captures the potentially detrimental C and N impurity atoms present in the matrix. Atom probe tomography results indicate elemental segregation of Cr, Mo, and Nb to grain boundaries in the as-extruded material, consistent with previous investigations of solid solution strengthening by solute additions. In conclusion, the ability of OFRAC to sequester impurities introduced from the powder metallurgical approach to nanostructured ferritic alloy development, compounded with its beneficial mechanical properties, makes this alloy a competitive candidate for fast reactor applications.« less

The kernel of fully ceramic microencapsulated (FCM) fuel requires a material with high fissile density. For that reason, among others, UN is the appropriate chemical state for low enriched U fuel. The UN kernels are spheres ~800 μm in diameter made using a sol-gel process. The effect of additives on the chemistry and density of the UN microspheres are investigated in this work. Gadolinium nitrate hexahydrate, Gd₂O₃, B, SiO₂ and SiC were incorporated into the sol-gel broth in varying concentrations. It was found that Gd can serve as both a sintering aid and burnable poison when added to the sol-gelmore » broth as a Gd(NO₃)₃6H₂O. However, even with the increased theoretical density of the UN microspheres, the U density was still too low for the FCM design that replaces UO₂ pellets in a commercial light water reactor. Silicon carbide and B were also successfully added but produced a lower density final product. Other sol-gel processing variables were investigated. In conclusion, the pour density of the sol-gel feedstock was found to influence the final converted UN kernel density.« less

The structure and speciation of fission products within the SiC barrier layer of tristructural-isotropic (TRISO) fuel particles irradiated to 19.6% fissions per initial metal atom (FIMA) burnup in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) was investigated. As-irradiated fuel particles, as well as those subjected to simulated accident scenarios, were examined. The TRISO particles were characterized using synchrotron X-ray absorption fine-structure spectroscopy (XAFS) at the Materials Research Collaborative Access Team (MRCAT) beamline at the Advanced Photon Source. The TRISO particles were produced at Oak Ridge National Laboratory under the Advanced Gas Reactor Fuel Development and Qualification Programmore » and sent to the ATR for irradiation. XAFS measurements on the palladium and silver K-edges were collected using the MRCAT undulator beamline. Analysis of the Pd edge indicated the formation of palladium silicides of the form Pd_xSi (2 ≤ x ≤ 3). In contrast, Ag was found to be metallic within the SiC shell safety tested to 1700 °C. To the best of our knowledge, this is the first result demonstrating metallic bonding of silver from fissioned samples. Knowledge of these reaction pathways will allow for better simulations of radionuclide transport in the various coating layers of TRISO fuels for next generation nuclear reactors. In conclusion, they may also suggest different ways to modify TRISO particles to improve their fuel performance and to mitigate potential fission product release under both normal operation and accident conditions.« less