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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Phase Relationships in the Carbon–Titanium–Uranium System for Ultra-High Temperature Nuclear Fuels

Carbides of uranium have attracted interest as fuels for nuclear thermal propulsion (NTP) to drive deep-space exploration owing to their attractive thermal and neutronic properties. Optimization of NTP technology, however, requires ultra-high temperature reactor environments to maximize the ratio of thrust to propellant to achieve peak rocket engine efficiency. Incorporation of transition metals into uranium carbides offers a pathway to increase the melting point of carbide fuels to address the operational challenges posed by NTP. Here, a thermodynamic model has been developed to examine the phase relationships in the C–Ti–U system at NTP conditions. Calculated phase relationships at ultra-high temperatures predict a stable (U, Ti)C solid solution. Additionally, experimental work on C–Ti–U synthesis via arc melting has been carried out, providing data on phase constitution and compositions that can be used to further refine the thermodynamic model that has been developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

UO 2 + 5 vol% ZrB 2 nano composite nuclear fuels with full boron retention and enhanced oxidation resistance

The boron isotope ( 10 B) can be used as a neutron absorber in UO 2 to control the reactivity of nuclear fuel pellets, however, the boron source can react with oxygen source in UO 2 to form B 2 O 3 that vaporize readily at temperatures above 1200 °C. Unfortunately, the sintering of UO 2 fuel requires hours holding at high temperature (>1700 °C), resulting in an inevitable B loss during sintering and unpredictable B concentration in final product. It is challenging to incorporate boron through a conventional sintering method. In this work, we demonstrated that spark plasma sintering (SPS), a field assisted sintering technology, can effectively densify UO 2 + 5 vol% ZrB 2 composite fuel pellets by rapid consolidation at 1600 °C for a short duration of 5 min under an applied pressure of 40 MPa. Further, thermogravimetric analysis (TGA) measurements confirm that ZrB 2 is fully retained inside the composite fuel pellets. Inside the composite fuel pellets, nano sized ZrB 2 particles are uniformly distributed along the grain boundaries of the UO 2 matrix. The ZrB 2 particle transforms to a glassy B 2 O 3 phase covering the sample surface and grain boundaries of UO 2 matrix after a simple post-sintering annealing at 1000 °C in flowing Argon gas for 4 h. The formed glassy B 2 O 3 slows down the diffusion of oxygen ions and postpones the onset temperature for oxidation of UO 2 from 400 °C to 550 °C. This study demonstrates the capability of SPS, an advanced fuel manufacturing technique, to achieve a full retention of ZrB 2 in UO 2 oxide fuel and increase oxidation resistance through a simple post-sintering annealing. The reported work holds great engineering potential for development of advanced oxide fuel for nuclear application.

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DOE OSTI2022

Near-infrared spectra and molar absorption coefficients of trivalent lanthanides dissolved in molten LiCl–KCl eutectic

Determining the concentration of the dissolved lanthanide species in LiCl–KCl eutectic salt is important to the development of pyrochemical reprocessing of used nuclear fuel. In this process, lanthanide fission products are found dissolved in the electrorefiner electrolyte in their trivalent oxidation state. The presence of dissolved trivalent lanthanides increases the liquidus temperature of the electrolyte mixture and can lead to the formation of insoluble oxide or oxychloride phases and must therefore be continuously monitored and controlled during the operation. Absorbance spectroscopy is a promising method for continuous measurement of the concentration of lanthanides and other elements dissolved in the electrolyte. The absorption of light by elements is linearly proportional to the concentration of the element for relatively dilute solutions according to the Beer-Lambert law. Although measurement of the absorption of ultraviolet and visible range light by lanthanides in LiCl–KCl eutectic molten salt have been explored previously, near infrared (NIR) absorption spectroscopy has received far less attention. It may, however, provide a better analytical signal when insoluble phases are present due to less Raleigh scattering compared to shorter wavelength radiation. Additionally, it may allow for concentration determination for certain elements using NIR absorption features where UV and visible range features are overlapping with features from other species. In this study, we report the UV–Vis–NIR spectra of the trivalent lanthanide chlorides of neodymium, samarium, and dysprosium in LiCl–KCl eutectic. Molar absorption coefficients are reported for analytically useful absorption maxima, with a focus on the molar absorption coefficients for NIR absorption maxima which have not been reported previously. Additionally, we observe a NIR-range absorption band of Nd3+ which was previously predicted but never experimentally observed. Here, we compare the calculated crystal field levels to the newly observed absorbance band and find them to be in good agreement with previous predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Recovering high-purity uranyl nitrate from simulated used nuclear fuel dissolver solutions by crystallization: rejecting technetium

The separation of U from Tc and other problematic fission product elements like Mo and Ru, along with Sr, Zr, Cs, and Nd, has been achieved via the crystallization of uranyl nitrate hexahydrate (UNH). Rejection of technetium as pertechnetate anion ( 99 TcO 4 – ) is an especially important feature of this system, as it otherwise tends to follow U (VI) in extractive separations. It also raises the salient question regarding why this oxoanion cannot replace nitrate within the crystalline lattice of UNH. Results showed high-yield (>90 %), high-purity (>99 %) recovery of U as UNH from solutions containing 99 TcO 4 – by simple reduction of temperature from 60°C to 20°C. There was no observable interaction of 99 TcO 4 – with UO 2 2+ . The addition of other cations like, Sr 2+ , Zr 4+ , Cs + , and Nd 3+ , also did not form secondary, contaminant solid phases, leaving the > 99 % of the fission product elements in the mother liquor, while the U was recovered at > 90 %. Similarly, Mo and Ru, when added to the mixture, were shown to behave as the other fission-product elements, remaining in the mother liquor during crystallization. As a result, DFT calculations showed that, despite the higher binding strength of TcO 4 – , HMoO 4 – , and BiO 3 – with the UO 2 2+ cation compared to NO 3 – , the hydrogen-bonding network of the two coordinated ions and four waters of hydration in the UNH crystal structure is the driving force for the high specificity of this separation.

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DOE OSTI2020

Microstructure, Thermal, and Mechanical Properties Relationships in U and UZr Alloys (Final Report)

Uranium-zirconium (U-Zr) alloys are candidate fuel systems for transmutation based reactors that can be used to burn long-lived minor actinides and fission products in fast spectrum reactors. Metallic fuels have also been gaining more recent attention for applications as accident tolerant fuels, as well as for use in small modular reactors. This research focused on a “science-based” approach to capture the connections between U and U-Zr alloys’ three-dimensional (3-D) microstructure, thermal properties, and mechanical properties through closely coordinated experiments and modeling efforts from the unirradiated to the irradiated fuels. Advanced characterization and modeling techniques were used to understand irradiation-induced microstructural evolution and its direct impact on the thermal and mechanical properties of U and U-Zr fuel. Closely coordinated experiments and modeling were performed to provide crucial data that does not currently exist. Overall, this research spanned multiple length and time scales within the models and experiments. The scope of the research encompassed the understanding of the irradiation effects in U and various U-Zr alloys with particular attention paid to three task areas: (1) microstructural evolution, (2) in-situ/ex-situ thermal and mechanical properties, and (3) multiscale modeling connections to microstructure, thermal, and mechanical properties. This research resulted in (1) the 3-D characterization of neutron irradiated U-Zr fuel in multiple phase regions to better understand fission gas swelling and constituent redistribution, (2) development of a microstructural model linking thermal and mechanical properties via in situ Raman and nanoindentation, (3) and mesoscale phase field modeling was coupled with the AEH method in the MOOSE framework was used to calculate the effective thermal conductivities of U-Zr fuels consisting of α-U and δ-UZr 2 heterogeneous microstructures.

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DOE OSTI2022

Spent Nuclear Fuel and Reprocessing Waste Inventory: Spent Fuel and Waste Disposition

This report provides information on the inventory of spent nuclear fuel (SNF) in the United States located at Nuclear Power Reactor (NPR) and Independent Spent Fuel Storage Installation (ISFSI) sites, as well as SNF and reprocessing waste located at U.S. Department of Energy (DOE) sites and other research and development (R&D) centers as of the end of calendar year 2021. Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future nuclear power generation scenarios, based on information available and assumptions made at the time the scenarios were developed in the spring of 2022. The report also includes select information on the characteristics associated with the wastes examined (e.g., type, packaging, heat generation rate, decay curves).

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DOE OSTI2022

Development and Evaluation of Neutron Thermalization Integral Benchmarks for Advanced Reactor Applications (Final Report)

In this project, integral benchmarks that aim to examine thermal neutron scattering data for graphite (ideal and nuclear), light water, and molten salt FLiBe will be analyzed. In particular, the project aims to: 1) Develop the slowing-down-time experiment that was performed at the OREAL facility of ORNL as a graphite benchmarks; 2) Identify and examine a slowing-down-time experiment in light water as a potential benchmark; 3) Reexamine the molten salt MSRE-MSR-EXP-001 benchmark to explicitly capture the impact of TSL variations.

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DOE OSTI2024

FY2024 Mid-Year Report: Verification of Spent Fuel Inside Dry Storage Casks by Cask Top Fast Neutron Mapping

This project is developing a prototype scanner array verification system for detection of missing fuel assemblies in spent-fuel storage casks. The prototype consists of six fast-neutron scintillator detectors mounted to a linear actuator frame that is placed on the top of a spent fuel cask to scan across all fuel assembly positions. The scanner array was assembled and tested at LLNL in prior years. A field test schedule has been requested at the Idaho National Laboratory (INL) Cask Farm site for FY2024. Following the Field Test, we will present results and discuss the technology with the IAEA. The IAEA may have special requirements for portability, shipping, lifting, and installation. We will incorporate additional improvements to the system based on lessons learned from the Field Test and feedback from the IAEA. If successful, the technology can be transferred to the IAEA or other stakeholders for assessment.

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DOE OSTI2023

Impact of Light Water Covariance on Integral Benchmarks [Slides]

New light water covariance has been generated. Impact of covariance quantified using two metrics on systems deemed sensitive to thermal scattering of light water. Analysis shows systems might not be sensitive enough to show significant differences. What differences were shown are most notable when comparing against the ENDF/B-VIII.0 covariance.

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DOE OSTI2024

ORNL Nuclear Data Evaluation Contribution to NCSP (ND-2) [Slides]

This presentation is titled ORNL Nuclear Data Evaluation Contribution to NCSP (ND-2). This lecture includes resonance evaluations, SAMMY fitting results, and resonance evaluation features. The presentation finishes with evaluations in alkaline earth-metals and halogens.

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DOE OSTI2024

Nuclear Fuel Cycle and Supply Chain (NFCSC) Technical Monthly September FY-24

AFC hosted the Light Water Reactor (LWR) Fuels Research Workshop and Electric Power Research Institute (EPRI) Collaborative Research on Advanced Fuel Technologies (CRAFT) Meeting September 9-13, 2024. The Department of Energy's (DOE) Advanced Fuels Campaign (AFC) Accident Tolerant Fuel (ATF) program is pivotal in advancing clean energy through collaborative innovation in fuel technology for light water reactors (LWRs). For over a decade, AFC has engaged a wide array of stakeholders—researchers, industry participants, and regulatory bodies—both within the U.S. and internationally to develop ATF technology. This initiative aims to enhance the safety and economic performance of both Pressurized Water Reactors (PWRs) and LWRs. To foster community relationships and disseminate research, AFC initiated annual LWR Fuel workshops in 2024, supplementing ongoing industry-led CRAFT workshops.

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DOE OSTI2020

Regulatory Development & Support for Advanced Reactor Technology Program

During 2012, the Department of Energy (DOE) instituted an Advanced Reactor Concepts Technical Review Panel process to evaluate viable reactor concepts from industry and identify research and development needs. Early in the process, the Technical Review Panel members and advanced reactor designers noted the need to develop a regulatory framework for non-light water advanced reactors to reduce risk and uncertainty to the industry. Concurrently, and in response to Congressional direction, the Nuclear Regulatory Commission (NRC) provided a report to Congress on advanced reactors. In this report, NRC noted that several reactor vendors also identified the need for regulatory guidance for advanced non-light water reactor designs.

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DOE OSTI2023

Waste Attributes of SMRs Scheduled for Near Term Deployment

This is a short presentation as part of a "roundtable" panel on Back-end Fuel Cycle Implications of Advanced Reactor Fuel Cycles. It covers an evaluation of waste generation rates projected for small modular reactors expected to be constructed this decade which are compared to a reference gigawatt-scale light water reactor like those in the current commercial fleet. For ease of comparison, results are normalized per unit of electricity generated.

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