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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Dislocation Loops in Proton Irradiated Uranium-Nitrogen-Oxygen System

Here in this study, we investigated the dislocation loop types formed in the proton-irradiated uranium-nitrogen-oxygen (U-N-O) system, which involves uranium mononitride (UN), uranium sesquinitride (a-U2N3), and uranium dioxide (UO2) phases. The dislocation loop formation is examined using specimens irradiated at 400°C and 710°C. Based on the detailed transmission-based electron microscopy characterization with i) the morphology-based on-zone and ii) the invisibility-criterion based two-beam condition imaging techniques, only a single type of dislocation loop in each phase is found: a/2?110?, a/2?111?, or a/3?111? dislocation loops in UN, a-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in UN phases indicate a critical loop size of ~ 6 nm, above which perfect loops are energetically favorable. This could explain the absence of faulted loops in the experimental observation of the irradiated UN phase at two temperatures. This work will enhance the understanding of irradiation induced microstructural evolution for uranium nitrides as advanced nuclear fuels for the next-generation nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Raman spectroscopy of uranium nitride kernels

Uranium nitride is an advanced fuel candidate for a wide variety of advanced nuclear reactors. This work summarizes the first characterization of UN kernels by Raman spectroscopy. First-principles density functional theory calculations were performed to predict the Raman spectra of uranium sesquinitride (U 2 N 3 ), uranium dinitride (UN 2 ), uranium mononitride (UN), uranium monocarbide (UC), as well as U-N-C (UN 1-x C x ) and a U-N-C-O mixture. Further, a core–shell structure was identified by scanning electron microscopy and Raman spectroscopy imaging. A signal at ~500 cm -1 was identified on the periphery of the core-shell structure, possibly corresponding to U 2 N 3 and/or UN 2 . This signal broadens and shifts to 470 cm -1 because of the formation of UNC, UNCO or U 2 N 3+x structures. The culmination of this work demonstrates the feasibility of using Raman spectroscopy to identify variations in composition and phases in UN kernels.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Complete Survey of Fuel Candidates for Microreactor Purposes

This report summarizes various fuel types that may be applicable to several microreactor concepts, which are defined as (1) very high temperature reactors (VHTR), (2) sodium fast reactors (SFR), (3) system for nuclear auxiliary power (SNAP) reactors, (4) gas fast reactors (GFR), and (5) molten salt reactors (MSR). The fuel systems that were assessed include: uranium mononitride (UN), uranium monocarbide (UC), uranium dioxide (UO 2 ), uranium oxycarbide (UCO) Tristructural Isotropic (TRISO), UN TRISO, mixed oxide (MOX), metallic fuels, and metal hydrides. While UCO TRISO has undergone significant testing through the advanced gas reactor (AGR) program, the very high cost necessitates consideration of other fuel types. UN and UC were identified as fuels that should receive further investigation due to their thermophysical and mechanical properties. Minimal irradiation performance data shows that these fuels are good candidate fuels for microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Fabrication of uranium nitride

A method of producing uranium nitride (e.g., uranium mononitride) and/or uranium-containing intermediates that can be used to further produce uranium nitride includes reacting a reaction mixture comprising uranium carbide with a gas comprising hydrogen and nitrogen, cooling the reaction mixture to a temperature suitable to produce a phase comprising U2N3, and heating the reaction mixture to a temperature suitable to convert the phase comprising U2N3 to a phase comprising UN. The method produces highly purified UN, is insensitive to excess carbon reactants, and can be readily integrated into existing uranium production facilities as part of a continuous process.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Resilience of uranium mononitride/zirconium carbide composites and uranium-zirconium carbonitride in hot hydrogen for nuclear thermal propulsion

Nuclear thermal rockets require fuels capable of withstanding flowing hydrogen propellant up to 3200 K. Presently, there has not been a fuel type that reliably operates at these conditions. A promising candidate anticipated to endure this demanding environment is a ceramic-ceramic composite comprising of uranium mononitride and zirconium carbide. Further, this investigation assesses the behavior and resilience of variations of this composite and its resultant homogenized form (uranium-zirconium carbonitride) under two hot hydrogen conditions (2273 K and 3000 K). The findings revealed that composites that homogenize into UZrCN exhibit superior structural integrity in hydrogen compared to heterogeneous counterparts. Consequently, this study underscores the potential of homogenized uranium-zirconium carbonitride for enhanced performance in nuclear thermal propulsion applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Assessment of uranium nitride interatomic potentials

Uranium mononitride (UN) is a promising nuclear fuel due to its high fissile density, high thermal conductivity, and suitability for reprocessing. In this study, two uranium nitride interatomic potentials are assessed: Tseplyaev and Starikov's angular-dependent potential and Kocevski et al.'s embedded atom model potential. Predictions of the thermophysical and elastic properties of UN, UN 2 , and α- and β-U 2 N 3 computed using both potentials are assessed and compared to available experimental data. Notably, the Tseplyaev potential performs better with the energetic aspects of UN, e.g., specific heat capacity and point defect formation energies, whereas the Kocevski potential performs better with the structural aspects of UN, e.g., thermal expansion as well as with the elastic properties. The reasons why the Kocevski potential underestimates the UN specific heat are explained by examining the UN phonon properties modeled using both potentials. The Kocevski potential shows better identification of the mechanical stability ranges of UN, UN 2 , and α- and β-U 2 N 3 , reasonably predicting the melting point of UN and predicting stable structures for UN 2 and α- and β-U 2 N 3 . On the other hand, the Tseplyaev potential predicts a premature phase change of both UN and UN 2 and cannot stabilize α- nor β-U 2 N 3 . However, the Kocevski potential cannot predict a stable α-U phase and is thus not suitable for the calculation of formation energies for non-stoichiometric point defects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mechanistic nuclear fuel performance modeling of uranium nitride

Uranium mononitride (UN) is a nuclear fuel candidate for advanced reactor designs and an alternative being considered for light water reactors due to its higher thermal conductivity and uranium density than UO 2 . As with any nuclear fuel, swelling and fission gas release are important factors for safety, while also being some of the hardest phenomena to predict with a high degree of confidence. Getting a grasp on the gas swelling behavior and release is crucial to lower the barrier for UN utilization. An accelerated swelling rate at high temperatures observed experimentally, sometimes referred to as “breakaway swelling,” further complicates the prediction of fuel performance of UN. A mechanistic model has been developed using a multiscale approach to describe the intragranular and intergranular fission gas behavior. Lower-length-scale calculations have been employed to inform models of the gas and self-diffusion behavior, resolution rate, and bubble shape. Leveraging previous work on high burnup UO 2 , two populations of intragranular bubbles are considered; small bulk bubbles and larger bubbles located along dislocations. The dislocation bubbles were found to be crucial to the overall swelling behavior, and the breakaway swelling transition was associated with the transition in the gas atom diffusion mechanism from an irradiation-induced athermal diffusion regime at lower temperatures to an intrinsic thermal equilibrium regime at higher temperatures, accelerating the growth of the dislocation bubbles. Similarly, the threshold for fission gas release was associated with the grain boundary vacancy diffusivity surpassing the gas atom diffusivity at sufficiently high temperatures, allowing the over-pressurized grain boundary bubble to grow in size and interconnect. Using thermo-mechanical models with the fission gas model, two integral fuel pin assessment cases were simulated. Finally, this work demonstrates the ability of a multiscale approach to accelerate the understanding of advanced fuel forms when experimental data is limited.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

UZrCN Synthesis via Arc Melting - A Novel Synthesis Study

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO2) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

36 MATERIALS SCIENCE↗
DOE OSTI2024

UZrCN Formation via Arc Melting – A Novel Synthesis Study

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO2) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

36 MATERIALS SCIENCE↗
DOE OSTI2024

UZrCN Synthesis via Arc Melting

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO 2 ) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO 2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Uranium nitride (UN) pellets with controllable microstructure and phase – fabrication by spark plasma sintering and their thermal-mechanical and oxidation properties

Dense uranium mononitride (UN) pellets with controlled microstructures and tailored grain size from large-grained to a few microns are synthesized by spark plasma sintering (SPS) combined with high energy ball milling. The impacts of the sintering conditions on fuel microstructure, grain size, physical density, and phase behavior are systematically investigated, and the thermal-mechanical properties and oxidation behavior of the SPS densified UN pellets are characterized. Higher sintering temperatures and longer ball milling durations and thus finer starting UN powders promote sintering and densification, and dense UN pellets above 95% theoretical density can be achieved by SPS at 1873 K for 10 min. UN phase purity is maintained in the SPS-densified pellets sintered at a lower temperature and short duration. A phase heterogeneity with secondary UO 2 or uranium sesquinitride (U 2 N 3 ) occurs for the UN pellets sintered at higher temperatures using finer UN powders. The hardness and fracture toughness of the SPS-densified UN pellets increase with smaller grain sizes and higher densities to 7.9 GPa and 3.5 MPa m 1/2 , respectively. Furthermore, both small (1–2 μm) and large grain-sized (30–50 µm) UN pellets exhibit good thermal conductivity. Dynamic oxidation testing by a thermogravimetric analyzer in air shows that the onset temperature for oxidation varies with microstructure and phase heterogeneity of the SPS densified UN pellets. Particularly, the smaller-grained (micron-sized) UN pellets containing uranium oxides and U 2 N 3 display lower weight gain and significantly-reduced oxidation kinetics, and full oxidation completes at a temperature above 1173 K when tested with a ramp rate of 10 K/min.

36 MATERIALS SCIENCE↗
DOE OSTI2022

LMFR design and optimization methodology

This paper presents and applies a methodology for designing and optimizing Liquid Metal-cooled Fast Reactors (LMFRs). The methodology, called the Design and Optimization Methodology (DOM), achieves its goals by the effective exploration of a search space. The paper presents two case studies to demonstrate DOM. Both studies were based on the WEC long-life core LFR design. In the first study, DOM was used to explore the search space in the vicinity of the WEC (Westinghouse Electric Company LLC) long-life core LFR design trying to find a more optimized design. The objective of this optimization attempt was to reduce the fuel cycle cost. The WEC long-life core LFR design uses uranium mononitride (UN) fuel with natural nitrogen enrichment. In the second study, DOM was used to design a competitive UN fueled Lead-cooled Fast Reactor (LFR) which uses {sup 15}N enriched UN fuel. Based on the results from the second study, an estimation of the {sup 15}N enrichment cost required to obtain an economic core was performed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Hot Hydrogen Testing of W-Coated UN Kernels in a Mo30W Matrix

Ceramic uranium mononitride (UN) is being considered as a reactor fuel for nuclear thermal propulsion. To avoid or reduce the dissociation of UN at the high temperatures needed, embedding it in a metallic matrix (cermet) has been proposed. To assess the viability of this concept, hot hydrogen testing of tungsten-coated UN kernels embedded in a Mo-30 wt% W (Mo30W) alloy matrix has been performed at temperatures from 1800°C to 2300°C. Both the isolated kernels and kernels consolidated by spark plasma sintering in the Mo30W matrix were tested. In addition to direct observations and mass loss measurements, the samples were analyzed by X-ray diffraction (XRD) and scanning electron microscopy (SEM)/energy dispersive X-ray spectroscopy (EDS) after each run. The decomposition of UN started at 1800°C despite the coating and matrix, and increased at 2000°C. Uranium seeped through the tungsten grain boundaries of the coating at all temperatures. The consolidated sample expanded irregularly at 2000°C through the formation of voids, and SEM/EDS analysis showed uranium-containing veins in the matrix consisting of U 2 Mo according to the XRD data. The observed pore generation at 2000°C was explained by the formation of water vapor from residual oxides and diffused hydrogen. At 2200°C and above, both the kernels and the consolidated samples melted through the formation of uranium or low–melting point uranium-molybdenum alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Formation of uranium nitride nanoparticles via mechanical alloying of uranium-molybdenum alloy fuels in gaseous nitrogen

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and fuel loading capability. However, U-Mo systems are susceptible to irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interactions. To address these shortcomings, this research investigated the formation of uranium mononitride (UN) nanoparticles within a 90 wt% U/10 wt% Mo (U-10Mo) matrix to act as a prospective defect sink for fission products at nanometric hetero-interfaces. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen atmosphere. Variations of the milling process investigated included media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. Characterization of the fuel microstructure was completed using light element analysis, X-ray diffraction, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. UN nanoparticles measuring 1–5 nm in radius were observed in the U-Mo matrix as early as 1 h into the mechanical alloying process. Milling time in excess of 10 h was found to lead to deleterious effects induced by the stainless-steel milling media.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

High volume packing fraction TRISO-based fuel in light water reactors

We report that for a decade, fully ceramic microencapsulated (FCM) fuel, containing tri-structural isotropic (TRISO) fuel particles in a silicon carbide (SiC) matrix, has been investigated as an accident-tolerant fuel for light water reactors (LWRs). Other examples exist of TRISO-based concepts for LWR fuels with different matrix materials. Previous studies assumed TRISO particle volume packing of approximately 0.44 in SiC (or another) matrix, the highest realistic packing fractions possible with conventional manufacturing. Recent advances in advanced manufacturing have yielded the development and demonstration of a fuel form that consists of conventionally manufactured TRISO particles in a 3D-printed SiC matrix with significantly higher possible TRISO packing fractions (0.5–0.7). This increased uranium loading enhances the viability of using TRISO-based particle fuel forms in LWRs. The viability of high-packing-fraction TRISO-based particle fuel forms in LWRs is assessed from the perspective of fuel cycle length, achievable fuel burnup, reactivity coefficients, and fuel cycle performance. Higher-packing-fraction TRISO-based fuel enables either longer cycle lengths (by ~25% at a packing fraction of 0.55 relative to 0.44) at a constant enrichment or decreased enrichments (by ~25% at a packing fraction of 0.55 relative to 0.44) at a constant cycle length. Studies of different fuel kernel types (uranium nitride, uranium oxycarbide, and uranium carbide) yield similar results, although the cycle length of uranium oxycarbide is shorter than for uranium nitride or uranium carbide (due to the lower density of uranium oxide). This work also characterized the production of 14 C resulting from neutron absorption in 14 N during operation for uranium mononitride fuel kernels; the ratio of 14 C/N was 1–2 at. % at discharge. For the fuel cycle evaluation, the activity of spent nuclear fuel and high-level waste at 100 and 100,000 years was lower for high-packing-fraction fuels than for conventional LWR fuel. Environmental impact metrics were similar overall, but higher on the front end of the fuel cycle and lower on the back end of the fuel cycle. Reactivity coefficients of higher-packing-fraction TRISO-based fuel were reasonable compared with those of conventional fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Formation and Consolidation of UN Nanostructures in U-Mo Fuel via Mechanical Alloying and Spark Plasma Sintering

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and high fuel loading capability. However, U-Mo systems suffer from irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interaction. To address these shortcomings, this thesis investigated the formation of uranium mononitride (UN) nanoparticles within a U-10Mo matrix. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen cover gas (99.9995%) atmosphere. Variations were made in stainless steel milling media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. UN nanoparticulates were successfully formed within the U-Mo matrix and was characterized utilizing light element analysis, X-ray diffractometry, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. Presence of the UN nanoparticles was found as early as 1-hour into the mechanical alloying process, and significant iron contamination found after 10-hours of milling. The U-Mo/UN powders were compacted using spark plasma sintering techniques in order to stabilize the nanostructure. U-Mo/UN compacts containing small amounts of neodymium and helium were analyzed to observe the defect migration tendencies of fission products within the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

NASA CERCER Development TDM Presentation

In NASA's Space Nuclear Propulsion program there is a development effort of for a composite fuel containing zirconium carbide (ZrC) and uranium mononitride (UN), this fuel form is referred to as a CERCER (ceramic-ceramic). These series of slides were put together to showcase the primary achievements in our research over the past year to our program sponsors. This is a yearly meeting to evaluate progress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗