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Raftery, Alicia M.

Publications and source records attributed to Raftery, Alicia M..

Properties and microstructure evolution of silicon nitride and zirconium nitride following Ni ion irradiation

We report that silicon nitride and zirconium nitride have been proposed as potential materials for multiple nuclear applications (inert matrix fuels, accident tolerant fuels, space nuclear power, fusion reactor design), yet knowledge on their behavior under irradiation remains limited. Ion irradiations were performed using 15 MeV Ni 5+ ions on Si 3 N 4 and ZrN samples, with midrange doses (around 3 µm) from 1 to 50 dpa and temperatures from 300 to 700°C. Volumetric lattice swelling was determined by grazing incidence X-ray diffraction, defect production and evolution were tracked using Transmission Electron Microscopy, and nanoindentation was performed to quantify the ceramics’ mechanical properties evolution. The results from these irradiation studies on nitride ceramics help fill the current gap present in the literature. Behavior consistent with past work on irradiated Si 3 N 4 was observed with respect to mechanical properties and defect formation up to 15 dpa and 500°C. Failure of the grain boundary sintering aid in Si 3 N 4 was observed above these conditions. Different behavior was observed in both nitrides at 50 dpa and 700°C, where lattice swelling increased past potential saturation values. Unreported cavity formation was witnessed in both materials under all irradiation conditions, with stable number density and slight size increase above 15 dpa. The mechanism for the cavity formation remains to be determined.

36 MATERIALS SCIENCE↗

Fabrication of UN-Mo CERMET Nuclear Fuel Using Advanced Manufacturing Techniques

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 plasma 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Separate effects irradiation testing of miniature fuel specimens

We report that qualification of new nuclear fuels is necessary for their deployment and requires a thorough understanding of fuel behavior under irradiation. Traditionally, nuclear fuels have been qualified by performing exhaustive integral tests under a limited range of prototypic conditions designed for their specific reactor application. While some integral fuel testing is essential, basic data on behavior and property evolution under irradiation can be obtained from separate effects tests. These irradiations could offer reduced cost, reduced complexity, and in the case of accelerated testing, reduced time to achieve a given burnup. Furthermore, it may be desirable to design test irradiations capable of deconvoluting the myriad effects of burnup, temperature gradients, and other factors inherent to integral irradiation tests. Oak Ridge National Laboratory has developed an experimental capability to perform separate effects irradiation testing of miniature fuel specimens in the High Flux Isotope Reactor (HFIR): the “MiniFuel” irradiation vehicle. The small size (<4 mm 3 ) of the fuel specimens simplifies the design, analysis, and post-irradiation examination. By reducing the fuel mass, the total heat generated inside the experiment vehicle can be dominated by gamma heating in the structural materials instead of fission heating in the fuel. This essentially decouples the fuel temperature from the fission rate, allowing for highly accelerated testing (3X-18X the burnup rate of a typical light water reactor for 235 U enrichments varying from 0.22 wt% to 8 wt%) and an extremely flexible experiment design that can accommodate a wide range of fuel temperatures (~100 °C to >1200 °C), compositions, enrichments, and even geometries without requiring detailed analyses for each fuel variant. In conclusion, this paper summarizes the experiment design concept, evaluates potential applications for specific fuel forms, and briefly describes the first set of experiments on uranium nitride kernels that have been assembled and are currently being irradiated in the HFIR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Lithography-based Ceramics Manufacturing Technologies

In this work, surrogate powders were prepared with lithography-based ceramic printing. Lanthanum oxide, a powder from Oak Ridge National Laboratory (ORNL), was prepared by milling, characterizing, and printing of coupon samples. The printing resulted in net-shaped green parts. Alumina, a more developed powder, was also printed and characterized at ORNL, and these parts were surrogate geometries larger than coupons that were sintered to net shape within 0.3% of the desired geometry.

36 MATERIALS SCIENCE↗

Production and Characterization Package of Reference and Advanced Feedstock

This milestone report summarizes the synthesis routes used to produce ceramic nuclear fuel feedstock at Oak Ridge National Laboratory (ORNL). The solution-gelation process is used to fabricate microspheres, which are then converted to ceramic fuel materials using various processing conditions. Resulting fuel materials include uranium dioxide (UO 2 ), uranium monocarbide (UC), and uranium mononitride (UN). The microsphere size distribution, density, and composition of the fuels can be altered using the conversion routes defined in this report. The resulting materials can be used as the feedstock in a number of additive manufacturing (AM) processes. The present materials were prepared using depleted or natural uranium to facilitate exploratory development activities. However, the laboratories used for this work are authorized for enriched uranium and are actively utilized for this purpose to fabricate fuel test articles for irradiation. If desired, the techniques demonstrated in this document can be readily used for enriched material production to support the Transformational Challenge Reactor (TCR) program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In situ X-ray characterization of uranium dioxide during flash sintering

The in situ response of stoichiometric and non-stoichiometric uranium dioxide during flash sintering is examined using high energy X-ray diffraction. Our results show that the onset of flash is driven by an increase in temperature and controlled by the applied field with no evidence of an accumulation of defects. The incubation time, that is the time after the application of the field before the flash occurs, is found to be material specific with hyper-stoichiometric samples requiring lower fields to flash. For low current/voltage fields we quantify very little change in the atomic and microstructure of the different uranium dioxide samples post-flash. Microstructural changes are identified for high fields and currents, where joule heating and sample temperatures are high, resulting in the complete transformation of the U 4 O 9 phase. Our results highlight the usefulness of high energy X-ray characterization in understanding the subtle structural changes that occur during the flash sintering process.

36 MATERIALS SCIENCE↗

Development of a characterization methodology for post-irradiation examination of miniature fuel specimens

The miniature fuel (MiniFuel) irradiation experimental design at Oak Ridge National Laboratory (ORNL) enables benchmark testing and separate effects testing of novel fuel types in the High Flux Isotope Reactor (HFIR) in support of the Advanced Fuels Campaign (AFC). The first round of MiniFuel sub-capsules will be available for post-irradiation examination at the end of 2018. This report establishes the procedure for disassembly, fuel retrieval, post-irradiation analysis of the fission product behavior (fission gas release and fuel swelling), and microstructural characterization of the irradiated MiniFuel specimens. All methods were tested and validated out-of-cell and are ready to be implemented in-cell for use on the first set of irradiated MiniFuel specimens. The path forward regarding further characterization techniques is presented, which will involve analyzing other specimen geometries using the methods presented in this report and expanding the post-irradiation analysis of MiniFuel specimens to include additional properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In-cell Re-fabrication and Loss-of-coolant Accident (LOCA) Testing of High Burnup Commercial Spent Fuel

A high burnup commercial spent fuel segment was re-fabricated into a pressurized test train for fuel safety testing in the Irradiated Fuels and Examination Laboratory (IFEL) hot cell facility at Oak Ridge National Laboratory. Descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident (LOCA) test train are provided. An in-cell integral LOCA test was performed in the Severe Accident Test Station (SATS) by subjecting the fuel segment to a steam environment, internal pressure, temperatures up to 1200 ºC, and water quench.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Completion of the Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

This document outlines the irradiation of silicon carbide cladding tube specimens in the High Flux Isotope Reactor at Oak Ridge National Laboratory. The cladding tube specimens consisted of monolithic, composite, and coated SiC specimens in order to test the effect of these various materials on the overall cladding performance during irradiation. A total of 18 specimens were irradiated for one cycle, with 9 specimens irradiated at low heat flux conditions and 9 specimens at high heat flux conditions. The specimens were inserted in cycle 475 in September 2017 and reached an average irradiation dose of approximately 2.6 dpa.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Tensile Testing of Irradiated Grade 92 Ferritic-Martensitic Steels at the IMET Hot Cell Facility

This report summarizes the completed post-irradiation tensile testing of five ferritic-martensitic (FM) steel SS-J2 specimens at the Irradiated Materials Examination and Testing (IMET) hot cell facility at Oak Ridge National Laboratory (ORNL). A brief description of the motivation for the study, the properties and irradiation conditions of the samples, and the tensile testing conditions are provided. The results will be used to compare the mechanical properties of the irradiated specimens to un-irradiated conditions to determine how irradiation affects the steel alloy material behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Onset conditions for flash sintering of UO 2

In this paper, flash sintering was demonstrated on stoichiometric and non-stoichiometric uranium dioxide pellets at temperatures ranging from room temperature (26°C) up to 600°C. The onset conditions for flash sintering were determined for three stoichiometries (UO 2.00 , UO 2.08 , and UO 2.16 ) and analyzed against an established thermal runaway model. The presence of excess oxygen was found to enhance the flash sintering onset behavior of uranium dioxide, lowering the field required to flash and shortening the time required for a flash to occur. Finally, the results from this study highlight the effect of stoichiometry on the flash sintering behavior of uranium dioxide and will serve as the foundation for future studies on this material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗