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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 829 records · Page 46
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Challenges and Solutions for Fast Neutron Irradiation of Bulk Material Specimens

Reactor developers continue to recognize opportunities for further enhancing fast spectrum reactor designs with advanced core materials, but all the material test reactors currently available to the United States are thermal spectrum designs. Fortunately, the Advanced Test Reactor and High Flux Isotope Reactor are versatile high flux facilities where spectral modification strategies can be used to reduce undesirable thermal neutron capture transmutation damage and augment fast flux delivered to specimens. New opportunities to leverage high flux regions and specially designed fast flux boosting experiment configurations can be used to achieve meaningful fast fluences on large specimens in ATR. New optimization potentials can be employed to achieve even higher fluences, albeit for smaller specimens, using thermal neutron filters in HFIR test positions. These capabilities, while not true fast reactors, can provide highly relevant environments for researchers needing to study the effects of fast neutron damage in bulk material specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Quantifying Uncertainties in Heat Capacity Measurements of Molten Salts Determined Using Differential Scanning Calorimetry

Uncertainty in specific heat capacity values of a molten salt determined by using differential scanning calorimetry (DSC) was assessed based on the precision of replicate measurements of heat flows used in the calculation and effects of corrections that are commonly made to heat flow measurements. The ratio method of determining heat capacity was applied using the results of replicate heat flow measurements made with two empty cells, a sapphire reference material, and three samples of a doped NaCl-UCl 3 salt mixture. Replicate measurements with empty cells were used to quantify the effects of system instabilities and sensitivities on the measured heat flows of sapphire and salt. The combined effects of uncertainties in individual heat flow measurements made with blank cells using this system were quantified to be 2.6 μV based on isothermal holds before and after the scan, with cell placement adding the greatest uncertainty. This value was used as the tolerance for accepting background-corrected heat flows measured with sapphire and salt to calculate the specific heat capacity. The acceptable heat flows measured for sapphire and salt over the temperature range of 540 to 725 °C resulted in calculated specific heat capacity values ranging from 0.53 to 0.91 J g −1 K −1 with an overall average value of 0.70 J g −1 K −1 and an uncertainty of 0.22 J g −1 K −1 at the 99 % confidence level. The combined uncertainty in the specific heat capacity masked detection of any effect of temperature or salt composition that occurred.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Grain size dependence of thermally induced oxidation in zirconium carbide

Here complementary analytical approaches were employed to probe the effect of grain size on thermally induced oxidation of zirconium carbide (ZrC) utilizing thermogravimetric analysis, differential scanning calorimetry, and Raman spectroscopy, as well as synchrotron-based and laboratory-based X-ray diffraction (XRD) experiments. The oxidation mechanism and phase behavior of nanocrystalline ZrC (grain size ~ 20 nm) were compared with that of the more documented microcrystalline ZrC (grain size ~ 1 µm). Synchrotron XRD at the Advanced Photon Source with a hydrothermal diamond anvil cell (HDAC) used as a sample chamber revealed that the onset of oxidation is at ~ 380 °C for microcrystalline ZrC which is in agreement with previous work. In contrast, the critical oxidation temperature was ~ 330 °C for nanocrystalline ZrC. Additional high-temperature synchrotron XRD experiments at the National Synchrotron Light Source II using a lamp furnace in combination with Raman analysis showed that tetragonal ZrO 2 forms as an initial oxidation product and transforms at higher temperatures to the monoclinic phase. Thermogravimetric analysis (TGA) coupled with differential scanning calorimetry (DSC) confirmed the X-ray results of a lower critical oxidation temperature for the nanocrystalline sample. The phase transformations in the oxide phase with associated critical temperatures were also evident in the thermodynamic data as exothermic heat events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Chlorination of uranium metal in molten NaCl-CaCl 2 via bubbling HCl

Molten chloride salt fast reactors (MCFRs) will require UCl 3 dissolved in molten salt mixtures as fuel for nuclear fission. For infusing the salt with UCl 3 , bubbling HCl into NaCl-CaCl 2 in contact with U metal was investigated. The reaction was run up to 9 h and yielded U concentration up to 0.652 wt.%. Open circuit potential between a W electrode and Ag/AgCl reference electrode yielded a potential consistent with uranium existing as U(III) in the salt. Furthermore, this demonstrates that HCl can be a very effective chlorinating agent to infuse MCFR fuel with UCl 3 starting from U metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Gamma-ray spectra analyses of molten salts in spent nuclear fuels pyroprocessing facilities for mass measurement

To evaluate radioactive tracer dilution and gamma spectroscopy as a safeguards technique for monitoring the mass of liquid salt in molten salt systems, the gamma spectroscopy data of salt samples from the pyroprocessing facilities in Idaho National Laboratory were acquired and analyzed. The primary focus is the uncertainty analysis of the gamma radioactivity of 154 Eu isotope in the molten salts. The facilities include an electrorefiner for pyroprocessing spent oxide fuels (SOF-ER) and Mark-IV ER for processing the metallic fuels from Experimental Breeder Reactor-II (EBR-II). The relative gamma radioactivity uncertainty was consistently at 3% for 137 Cs for both Mark-IV ER and SOF-ER salts. However, for 154 Eu, it was 8% for SOF-ER salt and 7% for Mark-IV ER salt. Here, the main reason for the higher uncertainty for 154 Eu is believed to be the lower counting statistics in gamma spectroscopy. To improve the gamma radioactivity uncertainty of isotope 154 Eu, a longer data acquisition time (12 h instead of 4 h) and a higher count rate gamma detector for gamma spectroscopy were tested and it was found that the uncertainty for 154 Eu was significantly improved from 7% or 8–3%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Image fusion for neutron tomography of nuclear fuel

Image fusion, the process of combining different images together, can be useful to create a more complete picture. In this work, image fusion is applied to neutron tomography of nuclear fuel with the goal of enhancing the information obtained about the fuel. Different reconstruction methods, such as Feldkamp, Davis and Kress filtered back projection and Simultaneous Reconstruction Technique, were combined to enhance image quality. This methodology was shown to reduce noise and ring artifacts without sacrificing sharp edges, allowing for a more accurate representation of sample geometry. Technique enhancements and future applications for the neutron imaging community are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Gamma-ray spectrometry analysis methods for radioisotope dissolution and mixing for nuclear forensics applications

We report the production of surrogate nuclear explosive debris involves the dissolution of irradiated fissile material, elemental isolation, and recombination based on required fission product ratios. Gamma-ray spectrometry is used following each step to assess the radioactive fission product inventory. A high-purity germanium detector is used to measure each sample, and the detector efficiencies are determined prior to sample measurements. Through radiochemical techniques, elements are isolated and recombined into samples with specific fission product ratios. As a result, a matrix-based approach can account for imperfect chemical separations to calculate the proper mixture volumes to obtain intended volatile-to-refractory fission product ratios, or R-values.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Qualitative assessment of uranium ore concentrates and related materials using scanning electron microscopy

Several studies have evaluated the morphology of uranium compounds produced under controlled conditions at the laboratory scale, but it is unclear whether the morphological characteristics of these materials persist in commercially produced uranium ore concentrates (UOCs). To assess the morphology in “real-world” UOCs, we qualitatively evaluated the morphological profile of secondary electron images from over 100 commercial UOCs using a previously published lexicon. We observe differences between samples with differing chemical composition and samples with similar chemical composition and differing provenance. Further, this work contextualizes morphology for commercially produced UOCs and will provide a basis for future machine learning efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development, feasibility, and uncertainty of radioactive 22 Na tracer dilution and gamma spectroscopy for mass determination of molten salt for pyroprocessing spent nuclear fuels

Accurate knowledge of the total mass of molten salts both for pyroprocessing spent nuclear fuels and for molten salt reactors is necessary for safeguards purposes. However, it is challenging to know or measure the total mass of molten salt due to the complicated shapes as well as the compositional—and thus density—changes that occur over time during operation. Here, we investigated radioactive tracer dilution (RTD) as a potential safeguards technique for application to the total mass measurement of LiCl–KCl–UCl 3 salt used for uranium electrorefining, with a focus on the feasibility, uncertainty, and fission product effect of RTD for salt mass measurement. To this end, experimental-scale RTD tests (600–700 g of total mass) were initially carried out by adding 22 Na tracer salt into LiCl–KCl salt containing 0, 5, and 28wt% radioactive salts from an electrorefiner (ER) for refining uranium. Upon the completion of the experimental-scale RTD tests, we performed a scale-up RTD test by directly adding 22 Na tracer salts to the ER, which uses about 12 kg of LiCl–KCl–UCl 3 salt during normal operation. This paper reports the main features of RTD for total mass measurement of molten salt and summarizes the results of the experimental-scale and scale-up RTD tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Time dependent chlorination of CeO 2 , La 2 O 3 and Nd 2 O 3 by ZrCl 4 dissolved in eutectic LiCl–KCl

Partial chlorination of CeO 2 , La 2 O 3 , and Nd 2 O 3 was achieved using ZrCl 4 dissolved in molten LiCl-KCl at 450°C over a period of 4-5 hr. ZrCl 4 was pre-volatilized into an argon stream to maintain a constant vapor pressure for equilibration with the salt. This approach was effective at maintaining the ZrCl 4 concentration at the solubility limit of 1.3 mol% in the molten salt throughout the duration of two of the chlorination experiments and within about 60% of the solubility limit in the third experiment. Based on the analysis of salt samples, fractional conversion after four hours of reaction for CeO 2 , and five hours for La 2 O 3 , and Nd 2 O 3 were calculated to be 0.97, 0.28, and 0.65 respectively. Assuming a shrinking core diffusion-limited model, the diffusivity of ZrCl 4 through the hypothesized ZrO 2 layer was calculated for the chlorination of CeO 2 , La 2 O 3 , and Nd 2 O 3 were 1.76 x 10 -10 cm 2 /sec, 7.85 x 10 -12 cm 2 /sec, and 7.42 x 10 -12 cm 2 /sec, respectively. Given the extremely low values calculated for diffusivity, it was concluded that the process is chemical reaction rate limited and that a complete reaction of each rare earth oxide could be achieved. Furthermore, a chemically controlled shrinking core model was shown to also fit the experimental data closely.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Noble gas constraints on spent fuel irradiation histories

Fission gas isotopic compositions are sensitive to a variety of reactor operating parameters that include the neutron flux, neutron fluence at different neutron energies, and operating temperature. Measurements of fission gas isotopic compositions thus have potential to constrain reactor simulations for nuclear forensics and safeguards applications. In this paper, we present Kr and Xe isotope measurements from a suite of samples obtained from locations that span the axial length a fuel pin with a well-characterized irradiation history and compare these data to spatially resolved reactor simulations. Here, we observed positive correlations between fluence sensitive isotopic ratios and burnup and between a flux sensitive ratio and power, although some discrepancies are observed between the measured data and model predictions. These differences may be due to simplifications in the model and/or inaccuracies in the cross sections. A much broader measurement to model comparison is required to better understand the discrepancies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Experimental determination of the electrochemical properties of bismuth chloride in eutectic LiCl–KCl and LiCl–KCl–CaCl 2 molten salts

Bismuth has been investigated as a potential liquid electrode for molten salt electrorefining, but the electrochemical behavior of Bi 3+ /Bi redox couple has received scant attention in LiCl–KCl and LiCl–KCl–CaCl 2 eutectics. Electroanalytical techniques were used to determine properties of Bi 3+ /Bi. Diffusivities of 6.71 and 8.93 × 10 –6 cm 2 s –1 and standard apparent potentials of 0.1166 and 0.1187 V versus Ag/AgCl (4.75 mol%) were recommended for BiCl 3 in LiCl–KCl and LiCl–KCl–CaCl 2 eutectics, respectively, at 680 ± 3.1 K. The ICP-MS analysis of BiCl 3 content limited the accuracy of the results. Furthermore, the reliability of the results improved with a more thorough drying procedure for CaCl 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Morphology and particle size (MaPS) exercise: testing the applications of image analysis and morphology descriptions for nuclear forensics

Image analysis techniques have been applied and shown to be a valuable tool in nuclear forensics analysis. The interlaboratory exercise reported here has tested quantitative and qualitative approaches for characterizing nuclear materials. Particle size, surface features and morphology descriptions were compared by four laboratories on a common image set generated by Scanning Electron Microscopy and Digital Light Microscopy. Quantitative analysis of the image sets through the Morphological Analysis for MAterials software highlighted the strength of image analysis, but also that the application of the software alone can introduce significant bias in the analysis. Qualitative morphology descriptions following the process outlined by Tamasi et al. (J Radioanal Nuclear Chem 307:1611–1619, 2015) were compared with a discussion on the robustness and reproducibility of the results. Finally, future work should continue to focus on proficiency and standardization of image analysis through continued exercises within the extended nuclear forensics community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Updated uranium reference material 231 Pa/ 235 U consensus model ages for nuclear forensics

Radiochronometry provides the model age of nuclear materials, which is useful for understanding the production history of materials found outside of regulatory control. Certified reference materials (CRMs) are important for radiochronometry to increase confidence in measurement quality; however, the absence of 231 Pa/ 235 U CRMs necessitates that nuclear forensic laboratories measure the 231 Pa/ 235 U model ages of preexisting uranium reference materials. Here, in this work, new consensus 231 Pa/ 235 U model ages are reported from three nuclear forensic laboratories for three reference materials using current analytical methods. These updated consensus values can be used for quality control of 231 Pa/ 235 U model age measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Single particle trace element analysis of uranium oxide standards with laser-induced breakdown spectroscopy

Femtosecond Laser-Induced Breakdown Spectroscopy (LIBS) was used to analyze over 1000 individual particles of uranium oxide trace element standard CRM124. Intra-granular analysis showed elemental concentrations were consistent across different surface locations on each uranium grain. However, inter-granular comparison revealed large heterogeneity between different grains of the same sample. By comparing the distribution of elemental concentrations across the particle populations, we were able to correctly identify a manufacturing process involving wet chemistry, rather than a dry mixing of two powders to give an overall chemical average. In conclusion, this demonstrates the utility of LIBS as a large-scale single particle analysis tool for nuclear forensics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Uncertainty improvement of 22 Na based radioactive tracer dilution for determining total mass of pyroprocessing molten salt systems by 154 Eu removal

To determine the total salt mass of the molten salt systems for pyroprocessing spent nuclear fuels, a 22 Na based radioactive tracer dilution was studied in Idaho National Laboratory in recent years. This 22 Na based RTD technique was deemed feasible, but due to the gamma energy peak of 22 Na coinciding with one of the energy peaks of 154 Eu radioisotope in the molten salt, the uncertainty of the 22 Na radioactivity in the 22 Na-spiked salt samples was quite high. To improve the uncertainty of the 22 Na based RTD technique, we proposed to chemically remove the 154 Eu of the salt samples by DGA resin for gamma spectroscopy. The effectiveness of removing 154 Eu on uncertainty improvement was evaluated. Furthermore, it was found that (1) the 154 Eu fission product effect on the uncertainty and detection limit can be effectively eliminated by chemically removing the 154 Eu during the salt sample preparation and (2) the uncertainty of 22 Na radioactivity in the salt samples for electrorefining was significantly improved from 13% to 2%, showing the potential of practical engineering application of 22 Na based RTD as a safeguards technique for molten salt systems for pyroprocessing spent nuclear fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2022

Mechanical Property Evaluation of CuNb Composites Manufactured with High-Pressure Torsion

Copper is under consideration as the optimum material for both high heat flux applications and high pulsed magnets. One challenge is that copper has low strength which is problematic to deployment in these applications. One solution is to alloy copper with BCC elements to improve its mechanical properties. However, the limited solubility of the BCC elements in copper requires high deformation processes to be used in order to manufacture these 3D composites. Here, high energy ball milling combined with high pressure torsion was used to manufacture 3D Cu-Nb composites. After the consolidation the mechanical properties of the composites were measured using micro- and nanohardness testing at room and elevated temperature. The results indicated that after 10 turns during the high-pressure torsion consolidation, the mechanical properties of the composites were completely saturated displaying uniform properties across the manufactured disk. Performing the high pressure torsion at temperature further improved consolidation of the disk. The high temperature nanoindentation also indicated a change in the deformation mechanism between 200 and 500 ºC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗