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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Isotope-Specific Analysis of Neutron-Irradiated Lithium Aluminate Ceramics by Nuclear Magnetic Resonance Spectroscopy

The effects of neutron irradiation on lithium aluminate targets and the permeation of tritium into surrounding components are primary concerns in the development of materials for use in tritium production. The high specific activity and volatility of the tritium in activated samples greatly increase the challenges of experimental analysis. This paper reports the determination of the chemical forms and quantities of H-3 and Li-6 in irradiated and non-irradiated LiAlO2 by nuclear magnetic resonance (NMR) spectroscopy, and demonstrates that measurements can be performed in a non-destructive, safe manner to obtain isotope-selective, quantitative information on highly hazardous materials.

Tritium, solid state nuclear magnetic resonance sp↗
DOE OSTI2020

The Impact of Ethylenediaminetetraacetic Acid (EDTA) on the Sorption of Nd(III), Th{sup (IV)}, and U(VI) onto Dolomite in WIPP-Relevant Brines, GWB and ERDA-6 - 20356

Storage of legacy, transuranic waste at the Waste Isolation Pilot Plant (WIPP) requires accurate risk models for performance assessments. These models must predict the long-term fate and transport of contaminants in the WIPP and surrounding environment. Laboratory support is necessary for the development of an accurate model through better understanding of contaminant behavior in the WIPP. The WIPP is a deep geologic repository that is characterized by high ionic strength due to evaporite deposits in the Salado formation (porewaters up to 7.4 M). High concentrations of sodium chloride (NaCl), calcium chloride (CaCl{sub 2}), and magnesium chloride (MgCl{sub 2}) are particularly of concern due to their potential impact on solubility and ultimately transport of contaminants of concern. Additionally, ligands capable of forming strong complexes with metals are found in significant concentrations in the WIPP environment due to their disposal from waste processing and formation during degradation of repository components (e.g. cellulose degradation by calcium hydroxide in cement). Consequently, there is a need to investigate the fate and transport of contaminants in the WIPP environment with a focus on the effects of high ionic strength and strongly coordinating ligands. Contaminants of high concern include the actinides (U(VI), Pu(IV), and Am(III)). Laboratory support of the WIPP modeling efforts was provided by this research. A series of batch experiments were utilized to study the impact of ionic strength, contaminant concentration, and the presence of ethylenediaminetetraacetic acid (EDTA) on the sorption of Nd(III), Th(IV), and U(VI) onto dolomite, a carbonate mineral found within the Culebra formation. The Culebra formation is the most transmissive layer above the WIPP, making transport of contaminants the most likely route in previous risk assessments. Two brines specifically designed to imitate conditions within the WIPP environment were used in these experiments. The U.S. Energy Research and Development Administration Well 6 (ERDA-6, 95%) brine simulates a low Mg environment, while the generic weep brine (GWB, 95%) simulates the WIPP environment with a high Mg concentration. Mg is added to the repository in the form of MgO as an engineered barrier and is naturally present in some formations. It is expected to scavenge carbonate from solutions decreasing formation of the highly soluble and competitive uranium-carbonate complexes. EDTA is used during manipulations of nuclear materials due to its ability to form strong, stable complexes with the actinides and may be found in high concentrations (up to 0.08 mM) within the WIPP as waste canisters degrade. Further, it may provide additional avenues for unwanted transport of contaminants. Nd(III) and Th(IV) were used as stable analogues of Am(III) and Pu(IV), respectively, to represent the most common oxidation states of actinides in the WIPP environment in addition to U(VI). The potential for colloid formation was also investigated with low and high concentrations of contaminants (10 and 1000 μg/L) using three size-dependent particle separation steps. The results of this research provide a better understanding of potential behavior of actinides in the presence of EDTA and within WIPP-relevant brines. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Quantum effects on the dynamics and properties of soft materials

The quantum effects of nuclear and electronic motion play an important role in the structure, dynamics, and function of soft materials, yet they are difficult to capture with conventional classical simulations or static electronic–structure methods. In this work several complementary approaches for treating quantum effects in polymeric and soft–matter systems are demonstrated, with a focus being on the hydrogen-bonded networks, ion and charge transport, and photoactive chromophores. The proton transfer, tunneling, and isotope effects are captured within the reduced-dimensionality models by implementing grid-based nuclear quantum dynamics in terms of the discrete variable and Fourier bases. The nuclear quantum dynamics is extended to larger systems by employing the quantum trajectories and quantum–thermal bath schemes combined with on-the-fly electronic structure, enabling the description of high-dimensional polymeric environments at feasible cost. The dynamics in the electronic degrees of freedom, simulating the optical response in large chromophores such as chlorophylls, is performed using the real-time time-dependent density functional theory implemented in the real-space multigrid (RMG) code. These approaches are demonstrated on case studies of the proton and hydroxide transport in hydrated polymer membranes, charge transfer in conjugated polymers, and the optical spectra of chlorophyll chromophores relevant to polymerized chlorophyll materials and chlorophyll–polymer hybrids. The reviewed methods and applications highlight practical routes of including quantum effects in simulations of soft functional materials.

Garashchuk, Sophya [Univ. of South Carolina, Colum↗
DOE OSTI2021

Oak Ridge National Laboratory Official TM Cover: Development and Qualification of Tungsten Reference Materials for LA-ICP-MS

Reference materials have been developed to provide areal density measurements of tungsten that is deposited on graphite when using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS). Several reference materials of varying areal densities have been produced using a magnetron sputtering, thin film deposition technique. With multiple reference materials that have varying areal densities, a calibration curve has been developed for translating the detector response of an ICP-MS into a quantitative measurement. These results provide the opportunity to make calibrated comparisons across graphite samples that were exposed to tungsten transport experiments for nuclear fusion research. In addition, these methods may be applied to provide individual areal density measurements for the five unique isotopes of tungsten. It is found that areal density measurements via LA-ICP-MS agree with characterization for the same samples when using the more traditional Rutherford backscattering ion beam technique. Qualification of these reference materials was completed using ion beam laboratories at the University of Tennessee, Knoxville and Massachusetts Institute of Technology, and Sandia National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104

The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Multi-scale fission product release model with comparison to AGR data

TRistructural ISOtropic (TRISO) particle fuel is central to several advanced, high-temperature reactor designs. Each particle consists of a fuel kernel encapsulated by three layers of carbon and ceramics that prevent the release of fission products and ensure physical integrity. Despite outstanding retention properties, fission product release has been observed from intact particles. To better understand and quantify fission product release from TRISO particles, a multiscale, mechanistic model of fission product transport is being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Previous work focused on silver (Ag) transport and improved Ag release predictions. The work described in this report builds on this experience to better understand cesium (Cs) transport in silicon carbide (SiC), the main barrier to the release of fission products. Atomistic simulations provide bulk and grain boundary (GB) Cs diffusivities in SiC, which are used by phase field simulations in the mesoscale code Marmot to determine the temperature, microstructure, and irradiation-dependent Cs diffusivity at the mesoscale in SiC. This approach attributes the different temperature regimes experimentally observed for Cs diffusivities in SiC to a transition from bulk-dominated diffusivity at high temperatures to a GB-dominated regime at low temperatures, providing new insight. The multiscale, mechanistic effective diffusivity is then implemented in the fuel performance code BISON and further validated by comparing Cs release predictions from Advanced Gas Reactor (AGR)-1 and AGR-2 post-irradiation measurements. The new model improves BISON’s predictability. This document also reports improvements made on Ag transport modeling by accounting for different GB types having different diffusivities. Moreover, this report details preliminary efforts to model palladium (Pd) attack of the SiC at the mesoscale using a phase field approach. Pd attack and its impact on accelerated Ag transport remains a misunderstood phenomenon, and we use the model to demonstrate that the formation of lamellae that has been observed in experiments can be explained by the reaction of Pd with SiC to form alternating layers of graphite and Pd 2 Si. This effort aims to improve our understanding of the reaction and eventually provide a model for BISON to account for Pd penetration and its effects on fission product release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A Dataset of 3D Structural and Simulated Transport Properties of Complex Porous Media

Physical processes that occur within porous materials have wide-ranging applications including - but not limited to - carbon sequestration, battery technology, membranes, oil and gas, geothermal energy, nuclear waste disposal, water resource management. The equations that describe these physical processes have been studied extensively; however, approximating them numerically requires immense computational resources due to the complex behavior that arises from the geometrically-intricate solid boundary conditions in porous materials. Here, we introduce a new dataset of unprecedented scale and breadth, DRP-372: a catalog of 3D geometries, simulation results, and structural properties of samples hosted on the Digital Rocks Portal. The dataset includes 1736 flow and electrical simulation results on 217 samples, which required more than 500 core years of computation. This data can be used for many purposes, such as constructing empirical models, validating new simulation codes, and developing machine learning algorithms that closely match the extensive purely-physical simulation. This article offers a detailed description of the contents of the dataset including the data collection, simulation schemes, and data validation.

3D images↗
DOE OSTI2022

Engineering defect energy landscape of CoCrFeNi high-entropy alloys by the introduction of additional dopants

The concept of high-entropy alloys (HEAs) focusing on tuning the overall chemical complexity represents a novel alloy design strategy. In contrast, alloying of a metallic matrix with minor doping elements with limited and localized tunability has been a common practice to improve material performance. Combining the idea of globally engineering defect energy landscape in HEAs and the localized doping strategy in dilute alloys, in this work, we explore doping effects of minor elements in a HEA matrix to further enhance the overall and localized chemical tunability, aiming to improve its irradiation resistance. Specifically, we study the influence of minor Al, Cu, Ti, and Pd substitutional doping elements on defect energetics in a CoCrFeNi model HEA based on density-functional theory (DFT) calculations. The DFT results indicate that the formation and migration energies of vacancies can be strongly influenced when a dopant is introduced at the first nearest neighbor shells around a vacancy. On the other hand, interstitial energetics are only slightly affected. Among the four elements, Ti and Pd generally decrease vacancy formation energies and increase vacancy migration energies more significantly than Al and Cu. The doping effects become more pronounced when the concentration of the substitutional dopants increases. Based on the energy distributions obtained from DFT, we build a kinetic Monte Carlo (kMC) model to assess the impact of dopants on vacancy-mediated diffusivity in the doped HEAs. Our results suggest that Ti and Pd can lower the tracer diffusivity in the considered HEAs and act as trapping sites, whereas Cu may enhance the atomic transport. This work indicates that substitutional doping in HEAs is an effective strategy in metallurgy to further tune the defect and transport properties of complex alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Particle Swarm Optimization Algorithm for Critical Experiment Design

Nuclear criticality experiments are used to validate nuclear cross section data used by simulation software. This is typically achieved by designing a critical system with a high sensitivity to a certain material’s cross section. Once the experiment has been carried out, a high fidelity model of the system is developed into a benchmark. When this benchmark model is simulated by a transport code, some of the difference between the experimental and computational effective neutron multiplication factor can be attributed to inaccurate nuclear data. Nuclear data evaluators then can make adjustments accordingly to improve cross section data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Thermal expansion of plasma-exposed tungsten

We report results from a systematic analysis of thermal expansion of plasma-exposed tungsten based on molecular-dynamics simulations using models of tungsten with distributions of helium (He) bubbles in the tungsten matrix. We distinguish between two approaches of filling the bubbles with He, where the amount of He in the bubble can or cannot vary with temperature. In the former case, the thermal expansion coefficient decreases monotonically with the porosity and He content of the tungsten matrix, while in the latter case, the thermal expansivity increases monotonically with increasing porosity and He content. The latter condition, where the He content in the bubble is determined at the implantation temperature and remains constant with varying temperature in the tungsten matrix, is consistent with He species transport in tungsten used as a plasma-facing component (PFC) in nuclear fusion reactors and implies the development of biaxial compressive thermal strains in the PFC material that contribute to accelerating the growth of a nanostructure on PFC tungsten surfaces. Furthermore our analysis advances the fundamental understanding of thermal expansion in PFC tungsten and contributes to the development of a thermophysical property database for properly incorporating effects of realistic heat loads into modeling the dynamical response of PFC tungsten under fusion reactor operating conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Derivation of at-birth neutron energy distribution of americium-beryllium source based on ISO8529 spectral data

In many practical applications, the accuracy of neutron transport calculations depends on an estimate of the energy distribution provided in the source definition. While the energy distribution of neutrons emitted by spontaneous fission sources is well-defined, the spectra of neutrons produced in (α,n) reactions in compound mixtures, such as americium-beryllium (AmBe), depend on the microstructural properties of the source material, which may vary. Although the energy of neutrons produced in nuclear reactions can be calculated from first principles, such computations require detailed knowledge of the key characteristics of the source material, which are often unknown. Here, this study focuses on deriving the at-birth neutron energy distribution for an AmBe source by applying reverse transport calculations to the externally measured high-precision spectrum recommended in the latest revision of the international standard. The derived at-birth spectrum was validated through indirect energy-sensitive neutron emission rate measurements of AmBe sources calibrated at national primary metrology institutions. The resulting at-birth neutron energy distribution serves as essential input data for radiation transport modeling, providing a validated spectrum for broader experimental and computational applications.

AmBe↗
DOE OSTI2020

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗
DOE OSTI2024

Development and Demonstration of a Risk Assessment Approach for Approval of a Transportation Package of a Transportable Nuclear Power Plant for Domestic Highway Shipment

The U.S. Department of Defense (DoD) Strategic Capabilities Office (SCO) has tasked PNNL to address the regulatory challenges associated with confirming the safe transport of Transportable Nuclear Power Plants (TNPPs) containing irradiated nuclear fuel. A previous report—Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages (PNNL-31867)—determined that the expected radioactive inventory in the irradiated fuel of a TNPP would likely require shipment in an NRC-approved Type B package (or spent nuclear fuel cask) but that a TNPP “package” is unlikely to meet the entire suite of NRC requirements set forth in Part 71 of Title 10 of the Code of Federal Regulations (CFR) for a Type B package. It was therefore concluded that shipment of this initial TNPP transportation package, as well as possibly others, under existing regulations would likely require NRC approval using the 10 CFR 71.12 (“Specific exemptions”) process that relies on risk-informed decision making supported by quantitative risk assessment (i.e., Probabilistic Risk Assessment).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Plasma Flow Reactor Investigation into Nucleation of Uranium Oxide on Silica Substrates to Better Understand Fallout Formation

Understanding particulate formation in nuclear debris is critical for predicting fallout transport after a nuclear event. Improved characterization of fallout formation and transport could lead to better guidance for emergency response in a post-detonation scenario. By analyzing how U-oxides nucleate onto different forms of SiO 2 (crystalline and amorphous), we can gain insight into how entrained environmental materials may incorporate into fallout. In this experiment a Plasma Flow Reactor (PFR), was used to replicate the high temperature and extreme flow conditions in a nuclear fireball. Uranyl nitrate was injected into the plasma, and PFR-generated Uranium oxides were allowed to condense onto amorphous (nanoparticles of varying sizes) or crystalline (quartz) SiO 2 substrates. SiO 2 substrates were characterized before and after U-oxide deposition using Scanning Electron Microscopy (SEM) based techniques (i.e. EDS) in order to characterize how U-oxides may nucleate onto these substrates. After a collection time of 4 minutes at an RF coil distance of 25cm, both the amorphous SiO 2 nanoparticles and the crystalline structures demonstrated dendritic nucleation of U-oxide species as identified by SEM/EDS.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Tracer Gas Model Development and Verification in PFLOTRAN

Tracer gases, whether they are chemical or isotopic in nature, are useful tools in examining the flow and transport of gaseous or volatile species in the underground. One application is using detection of short-lived argon and xenon radionuclides to monitor for underground nuclear explosions. However, even chemically inert species, such as the noble gases, have bene observed to exhibit non-conservative behavior when flowing through porous media containing certain materials, such as zeolites, due to gas adsorption processes. This report details the model developed, implemented, and tested in the open source and massively parallel subsurface flow and transport simulator PFLOTRAN for future use in modeling the transport of adsorbing tracer gases.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Ion dynamics in hexagonal boron nitride ionogel electrolytes

Ionogel electrolytes incorporating exfoliated hexagonal boron nitride (hBN) nanoplatelets are promising materials for next-generation energy storage systems. However, detailed understanding of their ion transport properties at the molecular level remains limited. This study employs diffusion and relaxation nuclear magnetic resonance (NMR) techniques, including fast-field cycling (FFC) NMR, to investigate the dynamics of ionic species in hBN-ionogels. By spanning a broad frequency range from 30 kHz using FFC NMR to high-field NMR (500–800 MHz), we reveal distinct relaxation mechanisms governing ion dynamics in ionogels with and without lithium salts. Our results highlight the role of hBN in modulating molecular rotation and translational motion, significantly affecting 1 H and 19 F relaxation profiles. The presence of Li + alters the dynamic behavior in ionogels, enhancing anion mobility at the interface. Notably, 7 Li relaxation reveals strong interactions with the hBN surface that cannot be detected by diffusion NMR. Furthermore, these findings underscore the importance of spanning a broad frequency range in NMR studies of ionogels and provide critical insights into optimizing their design as novel electrolytes.

Electrolytes↗