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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 1,585 records · Page 88
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Leveraging interpolation models and error bounds for verifiable scientific machine learning

Effective verification and validation techniques for modern scientific machine learning workflows are challenging to devise. Statistical methods are abundant and easily deployed, but often rely on speculative assumptions about the data and methods involved. Error bounds for classical interpolation techniques can provide mathematically rigorous estimates of accuracy, but often are difficult or impractical to determine computationally. Here, in this work, we present a best-of-both-worlds approach to verifiable scientific machine learning by demonstrating that (1) multiple standard interpolation techniques have informative error bounds that can be computed or estimated efficiently; (2) comparative performance among distinct interpolants can aid in validation goals; (3) deploying interpolation methods on latent spaces generated by deep learning techniques enables some interpretability for black-box models. We present a detailed case study of our approach for predicting lift-drag ratios from airfoil images. Code developed for this work is available in a public Github repository.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

Solubility of U(VI) in chloride solutions. III. The stable oxides/hydroxides in MgCl 2 systems: Pitzer activity model for the system UO 2 2+ –Na + –K + –Mg 2+ –H + –OH – –Cl – –H 2 O(l)

Here we have developed new chemical, thermodynamic and activity models for the system UO 2 2+ –Na + –K + –Mg 2+ –H + –OH – –Cl – –H 2 O(l) within the Pitzer approach. The new thermodynamic model is based on previously reported data treated within the SIT approach for NaCl and KCl systems, as well as on new experimental data determined in this work for the MgCl 2 system. The solubility of uranium(VI) was studied in 0.01–5.15 mol·kg w –1 MgCl 2 solutions at pH m = 4.1–9.7 (with pH m = –log [H + ]). Experiments were performed under Ar atmosphere at T = (22 ± 2) °C. In all investigated systems, the solubility of U(VI) is controlled by metaschoepite, UO 3 ·2H 2 O(cr). In contrast to previously investigated NaCl and KCl systems, no ternary Mg–U(VI)–OH(s) solid phases formed in alkaline MgCl 2 solutions within the timeframe of this study (≤ 200 days). A very significant increase in the solubility (up to 3 log 10 -units) is observed in acidic to near-neutral pH m conditions when increasing MgCl 2 concentration from 0.01 to 5.15 mol·kg w –1 , which reflects the strong ion interaction processes taking place in concentrated MgCl 2 brines. The solubility of UO 3 ·2H 2 O(cr) in the investigated NaCl, KCl and MgCl 2 solutions is well described with the solubility and hydrolysis constants recommended by Altmaier et al., (2017) and NEA–TDB, and a Pitzer activity model derived in the present work. The latter model considers experimental data reported in the present study and available in the literature for NaCl, KCl and MgCl 2 systems (solubility, potentiometric and spectroscopic data), in combination with well-stablished estimation methods and correlations with SIT coefficients. Finally, chemical, thermodynamic and Pitzer activity models provided in this work for the system UO 2 2+ –Na + –K + –Mg 2+ –H + –OH – –Cl – –H 2 O(l) accurately describe all evaluated datasets, and represent an adequate tool for the calculation of U(VI) solubility and aqueous speciation in a variety of geochemical conditions including concentrated brine systems of relevance in salt-based repositories for nuclear waste disposal.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

A web-based decision support system for collaborative mitigation of multiple water-related hazards using serious gaming

The paper presents a web-based application developed as a medium for executing a serious game to enhance multi-jurisdictional collaborative planning and decision making for mitigation of multiple hazards related to water (i.e., floods, soil erosions, water quality). The gaming application is integrated into the Iowa Watershed Decision Support System (IoWaDSS) designed to serve as a comprehensive multi-sectoral data repository along with the simulation results of process modeling to educate and inform communities on the status of their watersheds during extreme conditions. The interactive serious gaming application was devised to encourage public engagement, facilitate communication and positive relationship between watershed communities, and make the decision process more attractive and transparent for the stakeholders. The web environment offers easily-accessible, user-friendly interfaces that match the skills and knowledge level of a non-technical user. Developed with adaptive design and flexible architecture, the gaming application is generalizable and extendable to support hazard mitigation in other watersheds.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Laboratory investigation of the role of desorption kinetics on americium transport associated with bentonite colloids

Understanding the parameters that control colloid-mediated transport of radionuclides is important for the safe disposal of used nuclear fuel. We report an experimental and reactive transport modeling examination of americium transport in a groundwater–bentonite–fracture fill material system. A series of batch sorption and column transport experiments were conducted to determine the role of desorption kinetics from bentonite colloids in the transport of americium through fracture materials. We used fracture fill material from a shear zone in altered granodiorite collected from the Grimsel Test Site (GTS) in Switzerland and colloidal suspensions generated from FEBEX bentonite, a potential repository backfill material. The colloidal suspension (100 mg L –1 ) was prepared in synthetic groundwater that matched the natural water chemistry at GTS and was spiked with 5.5 × 10 –10 M 241 Am. Batch characterizations indicated that 97% of the americium in the stock suspension was adsorbed to the colloids. Breakthrough experiments conducted by injecting the americium colloidal suspension through three identical columns in series, each with mean residence times of 6 h, show that more than 95% of the bentonite colloids were transported through each of the columns, with modeled colloid filtration rates (k f ) of 0.01–0.02 h –1 . Am recoveries in each column were 55–60%, and Am desorption rate constants from the colloids, determined from 1-D transport modeling, were 0.96, 0.98, and 0.91 h –1 in the three columns, respectively. The consistency in Am recoveries and desorption rate constants in each column indicates that the Am was not associated with binding sites of widely-varying strengths on the colloids, as one binding site with fast kinetics represented the system accurately for all three sequential columns. As a result, our data suggest that colloid-mediated transport of Am in a bentonite-fracture fill material system is unlikely to result in transport over long distance scales because of the ability of the fracture materials to rapidly strip Am from the bentonite colloids and the apparent lack of a strong binding site that would keep a fraction of the Am strongly-associated with the colloids.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

First STAX detector installation at the National Institute for Radioelements (IRE)

The Source Term Analysis of Xenon (STAX) project has been installing stack detectors at medical isotope production facilities to measure radioxenon emissions to investigate the effect of radioxenon releases on nuclear explosion monitoring. This paper outlines the installation of the first STAX detection system at the National Institute for Radioelements (IRE) in Fleurus, Belgium which has been operating for over three years and transferring collected data to the STAX repository. Information about the equipment installed, the data flow established, and calculations for determination of radioxenon releases from the facility are presented. Further. data quality was investigated to confirm values reported by STAX automated data processing and in a comparison of collected STAX data with data collected by IRE for regulatory reporting.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Multistage bioassociation of uranium onto an extremely halophilic archaeon revealed by a unique combination of spectroscopic and microscopic techniques

The interactions of two extremely halophilic archaea with uranium were investigated in this paper at high ionic strength as a function of time, pH and uranium concentration. Halobacterium noricense DSM-15987 and Halobacterium sp. putatively noricense, isolated from the Waste Isolation Pilot Plant repository, were used for these investigations. The kinetics of U(VI) bioassociation with both strains showed an atypical multistage behavior, meaning that after an initial phase of U(VI) sorption, an unexpected interim period of U(VI) release was observed, followed by a slow reassociation of uranium with the cells. By applying in situ attenuated total reflection Fourier-transform infrared spectroscopy, the involvement of phosphoryl and carboxylate groups in U(VI) complexation during the first biosorption phase was shown. Differences in cell morphology and uranium localization become visible at different stages of the bioassociation process, as shown with scanning electron microscopy in combination with energy dispersive X-ray spectroscopy. Finally, our results demonstrate for the first time that association of uranium with the extremely halophilic archaeon is a multistage process, beginning with sorption and followed by another process, probably biomineralization.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

rcsb-api : Python Toolkit for Streamlining Access to RCSB Protein Data Bank APIs

The Protein Data Bank (PDB) was founded in 1971 as the first open-access digital data resource in biology to serve as the single global archive for three-dimensional (3D) macromolecular structure data. Current PDB holdings exceed 230,000 experimentally determined structures of proteins, nucleic acids, viruses, and macromolecular machines. The RCSB Protein Data Bank RCSB.org research-focused web portal facilitates search, analyses, and visualization of every PDB structure along with more than one million Computed Structure Models from AlphaFold DB and the ModelArchive. It is powered by a set of publicly available Application Programming Interfaces (APIs) that both support RCSB.org users and provide programmatic access to PDB data. Given the breadth and levels of granularity encompassed in this rich data collection, efficiently accessing the information programmatically may be challenging for new users. RCSB PDB has developed a Python software package, rcsb-api , that facilitates easy and efficient use of RCSB PDB APIs within a Python environment. This software tool is designed to streamline access to the extensive corpus of data housed within the PDB, enabling researchers to search, retrieve, and analyze 3D biostructure data seamlessly. Its use will accelerate research in structural biology, molecular biology and biochemistry, drug discovery, and bioinformatics by providing more efficient tools for data integration and analysis. The new toolkit is available on GitHub (github.com/rcsb/py-rcsb-api) and published to the public Python package repository (PyPI) to foster wider usage and support basic and applied research in fundamental biology, biomedicine, and the energy sciences.

FAIR principles↗
DOE OSTI2012

Dissolution Behavior of Plutonium Containing Zirconia-Magnesia Ceramics

This study explores the dissolution properties of zirconia-magnesia ceramics containing plutonium as the basis of an inert atrix nuclear fuel. The magnesium oxide phase remains pure MgO, while the zirconia incorporates a small amount of magnesium oxide along with all of the plutonium oxide and erbium oxide. The performance of the material under reactor and repository environments was examined. Reactor conditions are examined using a pressure vessel to expose the material to 300 degrees C water. To assess the performance of the material as a waste form it was submerged in 90 degrees C water for 1000 h. In both aqueous dissolution studies there was minimal release of less than 0.8 wt.% of plutonium from the material. To examine the potential for recycling, the dissolution behavior of the fuel matrix was examined in acidic solutions: pure nitric acid and a nitric acid-hydrofluoric acid-peroxide solution. Both acidic media exhibit potential for dissolving plutonium from the zirconia matrix. The experiments performed in this study are meant to lay a foundation for the chemical performance of zirconia-magnesia inert matrix fuel containing fissile material and burnable poison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4,136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

converter slag↗
DOE OSTI2019

Effect of solution chemistry on the iodine release from iodoapatite in aqueous environments

To ensure the safe disposal of nuclear waste, understanding the release process of radionuclides retained in the nuclear waste forms is of vital importance. Iodoapatite Pb 9.85 (VO 4 ) 6 I 1.7 , a potential waste form for iodine-129, was selected as a model system for ceramic waste forms in this study to understand the effect of aqueous species on iodine release. Semi-dynamic leaching tests were conducted on bulk samples in cap-sealed Teflon vessels with 0.1 mol/L NaCl, Na 2 CO 3 , Na 3 PO 4 , and Na 2 SO 4 solutions under 90 °C, fixed sample surface area to solution volume ratio of 5/m, and periodic replacement of leaching solutions. The reacted solutions were then analyzed by Inductively Coupled Plasma-Mass Spectrometry and Inductively Coupled Plasma-Optical Emission Spectrometry; the leached surfaces were characterized by X-ray diffraction, scanning electron microscopy, and infrared spectroscopy. The finding shows that, compared to deionized water, the ion-rich solutions enhanced the iodine release as a result of the increased ionic strength, reduced activity coefficient of dissolved species, and increased solution pH. Surface reactions can lead to the formations of secondary phases by ion-exchange and precipitation. These findings indicate that an ion-rich environment in the geological repository can be detrimental to the disposal safety of the nuclear waste form.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Testing fast reactor fuels in a thermal reactor: Comparison of transmutation metallic fuel alloys behavior by scanning electron microscopy

To optimize nuclear waste repository performance, the destruction of minor actinide elements, particularly Np and Am, in a neutron fast spectrum reactor is possible by incorporating these elements into nuclear fuel. Evaluating the performance of minor actinide containing fuel is of paramount importance to enabling this technology. However, such a task is challenging without an available domestic fast spectrum test reactor. A comparison of fuel performance tested in an available domestic thermal reactor at the Idaho National Laboratory, the Advanced Test Reactor, and in a fast spectrum reactor in France (Phénix) is presented here in this study. This study evaluates the capability of using a cadmium shrouded test position to mimic the power profile along the fuel radius present in fast spectrum reactors so that thermally driven phenomenon (e.g., constituent redistribution) can be evaluated in a thermal reactor and determined to be prototypical of a fast reactor. Thus, optical microscopy and scanning electron microscopy has been performed on irradiated 35U-29Pu-4Am-2Np-30Zr fuel samples (where the number preceding the element is the weight percent concentration) from the two mentioned reactors that present similar irradiation temperatures and power conditions. The results indicate that fuel performance phenomena are reproducible in the two irradiation conditions. The redistribution of Zr occurred in the same manner for the two samples. Similar partitioning of U-Pu-Zr phases was observed, and the behavior of Am was similar in the analyzed specimens. Finally, the overall microstructure evolution seems not to be affected by minor actinides addition compared to expected behavior of conventional U-19Pu-10Zr ternary metal fuels for both specimens. Slight differences in fuel cladding chemical interaction were, however, observed. This difference is likely driven by difference in cladding composition rather than irradiation conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). Here, t o investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Simulation of dilatancy-controlled gas migration processes in saturated bentonite using a coupled multiphase flow and elastoplastic H 2 M model

Dilatancy-controlled gas flow in preferential pathways plays a key role in the safety analysis of radioactive waste repositories. This is particularly the case for bentonite, an often-preferred barrier material. Gas flow in preferential pathways is characterized by localization and spontaneous behavior, which is challenging to simulate in numerical models due to strong hydro-mechanical coupling. To analyze a laboratory experiment in the framework of the DECOVALEX-2023 project, this study introduced a new approach of combining continuous modelling methods with spatial material properties derived from material heterogeneities and experimental observations. The proposed model utilized hydro-mechanical spatial distributions, namely Young's modulus and gas entry pressure, and elastoplasticity combined with a linear swelling model. A conceptual strain-dependent permeability approach simulated dilatancy-controlled gas flow based on hydro-mechanical coupling. To test the effectiveness of the presented approach, a gas injection test in a compacted, saturated bentonite sample was simulated using the open-source code OpenGeoSys 5.8 and compared with experimental observations. The presented methodology is capable of simulating localized gas flow in preferential pathways. The spatial distributions of Young's modulus and gas entry pressure affect the swelling pressure, relative permeability and, in combination with the strain-dependent permeability model, also the intrinsic permeability.

42 ENGINEERING↗
DOE OSTI2016

Development in corrosion resistance by microstructural refinement in Zr-16 SS 304 alloy using suction casting technique

Highlights: • Grain refinement was made in Zr–16 wt.% SS alloy while prepared by suction casting process. • Distribution of Laves phase, e.g., Zr{sub 2}(Fe, Cr) was raised in suction cast (SC) Zr–16 wt.% SS. • Corrosion resistance was improved in SC alloy compared to that of arc-melt-cast alloy. • Grain refinement in SC alloy assisted for an increase in its corrosion resistance. - Abstract: Zirconium (Zr)-stainless steel (SS) hybrid alloys are being considered as baseline alloys for developing metallic-waste-form (MWF) with the motivation of disposing of Zr and SS base nuclear metallic wastes. Zr–16 wt.% SS, a MWF alloy optimized from previous studies, exhibit significant grain refinement and changes in phase assemblages (soft phase: Zr{sub 2}(Fe, Cr)/α-Zr vs. hard phase: Zr{sub 3}(Fe, Ni)) when prepared by suction casting (SC) technique in comparison to arc-cast-melt (AMC) route. Variation in Cr-distribution among different phases are found to be low in suction cast alloy, which along with grain refinement restricted Cr-depletion at the Zr{sub 2}(Fe, Cr)/Zr interfaces, prone to localized attack. Hence, SC alloy, compared to AMC alloy, showed lower current density, higher potential at the breakdown of passivity and higher corrosion potential during polarization experiments (carried out under possible geological repository environments, viz., pH 8, 5 and 1) indicating its superior corrosion resistance.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A publicly available PyTorch-$\mathrm{ABAQUS}$ $\mathrm{UMAT}$ deep-learning framework for level-set plasticity

Here this paper introduces a publicly available PyTorch-ABAQUS deep-learning framework of a family of plasticity models where the yield surface is implicitly represented by a scalar-valued function. In particular, our focus is to introduce a practical framework that can be deployed for engineering analysis that employs a user-defined material subroutine (UMAT/VUMAT) for ABAQUS, which is written in FORTRAN. To accomplish this task while leveraging the back-propagation learning algorithm to speed up the neural-network training, we introduce an interface code where the weights and biases of the trained neural networks obtained via the PyTorch library can be automatically converted into a generic FORTRAN code that can be a part of the UMAT/VUMAT algorithm. To enable third-party validation, we purposely make all the data sets, source code used to train the neural-network-based constitutive models, and the trained models available in a public repository. Furthermore, the practicality of the workflow is then further tested on a dataset for anisotropic yield function to showcase the extensibility of the proposed framework. A number of representative numerical experiments are used to examine the accuracy, robustness and reproducibility of the results generated by the neural network models.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Immobilization and image-guidance methods for radiation therapy of limb extremity soft tissue sarcomas: Results of a multi-institutional survey

Radiation therapy for limb-extremity soft tissue sarcoma (STS) requires accurate, reproducible dose delivery. However, patient positioning is challenging and there is a lack of existing guidelines to assist institutional standardization. Therefore, we conducted a multi-institutional international survey of STS immobilization, image guidance methods, and treatment protocols to investigate current practice. Seventy-three UK radiotherapy centers and 15 hospitals in 7 other countries completed a questionnaire on STS immobilization and image-guidance procedures. Specifically, the survey collated information on the current usage of immobilization equipment, including custom devices, patient setup tolerances, the use of written protocols, the modality and frequency of image guidance, the method of treatment, allocated treatment times, and the application of surgical clips. Multiple combinations of immobilization devices were reported. In the UK, 12%, 40%, 30%, 12%, and 5% use 1, 2, 3, 4, and 5 types of device for lower limb STS. Vacuum bag plus either foot or ankle support was most common (66%). Of 15 international centers, 27%, 60%, 7%, 0%, 7% use 1, 2, 3, 4, 5 devices, with vacuum bags (73%) and thermoplastic (47%) predominant, similar to UK values of 77% and 52%. For image guidance, in the UK, 37% use kV planar, 34% use MV planar, and 16% use cone-beam CT for the first 3 fractions and then weekly. Internationally, daily imaging was more prevalent with 33% using kV planar, 7% MV planar, and 40% cone-beam CT daily. Custom devices plus combinations of devices, along with 5- and 10-mm set-up tolerances, were most commonly reported. Less than half of centers have written treatment protocols. Conventional treatment is most common in the UK, with only 42% using conformal techniques. Treatment is allocated between 10 and 30 minutes. Twenty-six percent of UK centers and 53% of international centers use surgical clips. Across treatment centers, there is no consistent approach to STS immobilization, image-guidance methods, or treatment protocols assessed by this survey. A wide variety of immobilization devices and configurations are utilized, and the frequency and modality of imaging are similarly diverse. In the absence of guidelines, the creation of an online repository of example immobilization techniques could enable centers to compare a diversity of cases. The availability of a forum for viewing and discussing a range of cases could potentially lead to improved patient setup and reduce the time taken to devise an individual immobilization strategy.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2024

Nuclear waste attributes of near-term deployable small modular reactors

The nuclear waste attributes of near-term deployable SMRs were assessed using established nuclear waste metrics, which are the DU mass, SNF mass, volume, activity, decay heat, radiotoxicity, and decommissioning LLW volumes. Metrics normalized per unit electricity generation were compared to a reference large PWR. Three SMRs, VOYGR, Natrium, and Xe-100, were selected because they represent a range of reactor and fuel technologies and are active designs deployable by the decade’s end. The SMR nuclear waste attributes show both some similarities to the PWR and some significant differences caused by reactor-specific design features. The DU mass is equivalent to or slightly higher than the PWR. Back-end waste attributes for SNF disposition vary, but the differences have a limited impact on long-term repository isolation. SMR designs can vary significantly in SNF volume (and thus heat generation density). However, these differences are amenable to design optimization for handling, storage, transportation, and disposal technologies. Nuclear waste attributes from decommissioning vary depending on design and decommissioning technology choices. Given the analysis results in this study and assuming appropriate waste management system and operational optimization, there appear to be no major challenges to managing SMR nuclear wastes compared to the reference PWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗