Search NASA⌕ Search

Engineering topics

Morgan, Dane

Publications and source records attributed to Morgan, Dane.

At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

2 H and 27 Al solid-state NMR study of the local environments in Al-doped 2-line ferrihydrite, goethite, and lepidocrocite

Although substitution of aluminum into iron oxides and oxyhydroxides has been extensively studied, it is difficult to obtain accurate incorporation levels. Assessing the distribution of dopants within these materials has proven especially challenging because bulk analytical techniques cannot typically determine whether dopants are substituted directly into the bulk iron oxide or oxyhydroxide phase or if they form separate, minor phase impurities. These differences have important implications for the chemistry of these iron-containing materials, which are ubiquitous in the environment. In this work, 27 Al and 2 H NMR experiments are performed on series of Al-substituted goethite, lepidocrocite, and 2-line ferrihydrite in order to develop an NMR method to track Al substitution. The extent of Al substitution into the structural frameworks of each compound is quantified by comparing quantitative 27 Al MAS NMR results with those from elemental analysis. Magnetic measurements are performed for the goethite series to compare with NMR measurements. Static 27 Al spin–echo mapping experiments are used to probe the local environments around the Al substituents, providing clear evidence that they are incorporated into the bulk iron phases. As a result, predictions of the 2 H and 27 Al NMR hyperfine contact shifts in Al-doped goethite and lepidocrocite, obtained from a combined first-principles and empirical magnetic scaling approach, give further insight into the distribution of the dopants within these phases.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Local strain effect on the band gap engineering of graphene by a first-principles study

We have systematically investigated the effect of local strain on electronic properties of graphene by first-principles calculations. Two major types of local strain, oriented along the zigzag and the armchair directions, have been studied. We find that local strain with a proper range and strength along the zigzag direction results in opening of significant band gaps in graphene, on the order of 10{sup −1 }eV; whereas, local strain along the armchair direction cannot open a significant band gap in graphene. Our results show that appropriate local strain can effectively open and tune the band gap in graphene; therefore, the electronic and transport properties of graphene can also be modified.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2015

Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re-evaluate thermophysical properties of flibe and flinak. Pacific Northwest National Laboratories has focused on evaluating the fluorinating gas nitrogen trifluoride as a potential salt purification agent. Work there was performed on removing hydroxides and oxides from flinak salt under controlled conditions. Lastly, the University of California Berkeley has spent considerable time designing and simulating reactor components with fluoride salts at high temperatures. Despite the hurdles presented by the innate chemical hazards, considerable progress has been made. The stage has been set to perform new research on salt chemical control which could advance the fluoride salt cooled reactor concept towards commercialization. What were previously thought of as chemical undesirable, but nuclear certified, alloys have been shown to be theoretically compatible with fluoride salts at high temperatures. This preliminary report has been prepared to communicate the construction of the basic infrastructure required for flibe, as well as suggest original research to performed at the University of Wisconsin. Simultaneously, the contents of this report can serve as a detailed, but introductory guide to allow anyone to learn the fundamentals of chemistry, engineering, and safety required to work with flibe salt.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Oxygen surface exchange kinetics and stability of (La,Sr) 2 CoO 4±δ /La 1-x Sr x MO 3-δ (M = Co and Fe) hetero-interfaces at intermediate temperatures

Heterostructured oxide interfaces created by decorating Ruddlesden-Popper phases (A2BO4) or perovskites on perovskites have shown not only pronounced cation segregation at the interface and in the A2BO4 structure but also much enhanced kinetics for oxygen electrocatalysis at elevated temperatures. In this study, we report and compare the time-dependent surface exchange kinetics and stability of (La 0.5 Sr 0.5 ) 2 CoO 4 -decorated (LSC 214 ) La 0.6 Sr 0.4 Co 0.2 Fe 0.8 O 3-δ (LSCF 113 ) and La 0.8 Sr 0.2 CoO 3-δ (LSC 113 ) thin films. While LSC 214 decoration on LSC 113 greatly reduced the degradation in the surface exchange kinetics as a function of time relative to LSC 113 , LSCF 113 with LSC 214 coverage showed comparable surface exchange kinetics and stability to LSCF 113 . This difference can be explained by greater surface stability of LSCF 113 than LSC 113 under testing conditions, and that LSC 214 decoration on LSC 113 reduced the decomposition of LSC 113 to form secondary phases that impedes oxygen exchange kinetics, and thus resulted in enhanced stability. This hypothesis is supported by the observations that annealing at 550 °C led to the formation of Sr-rich secondary particles on LSC 113 while no such particles were observed on LSCF 113 . Density functional theory (DFT) computation provides further support, which revealed greater capacity of surface Sr segregation for LSCF 113 having SrO termination than LSC 113 having (La 0.25 Sr 0.75 )O termination for the experimental conditions, and lower energy gain to move Sr from LSCF 113 into LSC 214 relative to the LSC 214 -LSC 113 system.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Enhanced Mixed Electronic-Ionic Conductors through Cation Ordering

The performance of many energy conversion and storage devices depend on the properties of mixed ionic-electronic conducting (miec) materials. Mixed or ambipolar conductors simultaneously transport ions and electrons and provide the critical interface between chemical and electrical energy in devices such as fuel cells, ion transport membranes, and batteries. Enhancements in storage capacity, reversibility, power density and device lifetime all require new materials and a better understanding of the fundamentals of ambipolar conductivity and surface reactivity.The high temperature properties of the ordered perovksites AA’B 2 O 5+x , where A = rare earth ion, Y and B = Ba, Sr were studied. The work was motivated by the high oxygen transport and surface exchange rates observed for members of this class of mixed ionic and electronic conductors. A combined experimental and computational approach, including structural, electrochemical, and transport characterization and modeling was used. The approach attacks the problem simultaneously at global (e.g., neutron diffraction and impedance spectroscopy), local (e.g., pair distribution function, nuclear magnetic resonance) and molecular (ab initio thermokinetic modeling) length scales. The objectives of the work were to understand how the cation and associated anion order lead to exceptional ionic and electronic transport properties and surface reactivity in AA’B2O5+x perovskites. A variety of compounds were studied by X-ray and neutron diffraction, measurements of thermodynamics and transport and theoretically. These included PrBaCo 2 O 5+x and NdBaCo 2 O 5+x , PrBaCo 2-x FexO 6- δ (x = 0, 0.5, 1.0, 1.5 and 2) and LnBaCoFeO 6- δ (Ln = La, Pr, Nd, Sm, Eu and Gd), Sr 3 YCo 4 O 10.5 , YBaMn 2 O 5+x . A 0.5 A’ 0.5 BO 3 (where A=Y, Sc, La, Ce, Pr, Nd, Pm, Sm; A’= Sr, Ba; and B= Fe, Co, Mn, Ni), Ba 2 In 2 O 5 , and La 1 x Sr x CoO 3-δ /(La 1-y Sry) 2 CoO 4±δ interfaces.

36 MATERIALS SCIENCE↗
DOE OSTI2014

nomalous Interface and Surface Strontium Segregation in (La 1-y Sr y ) 2 CoO 4 /La 1-x Sr x CoO 3- Heterostructured Thin Films

Heterostructured materials have shown unusual physiochemical properties at the interfaces such as two dimensional electron gas systems, high-temperature superconductivity, and enhanced catalysis. Here we report the first atomic-scale evidence of the microscopic structure of a perovskite/Ruddlesden-Popper heterostructure (having La1-xSrxCoO3- /(La1-ySry)2CoO4 ), and anomalous strontium segregation at the interface and in the Ruddlesden-Popper structure using direct X-ray methods combined with ab initio calculations. The remarkably enhanced activity of such heterostructured surfaces relative to bulk perovskite and Ruddlesden-Popper oxides previously shown for oxygen electrocatalysis at elevated temperatures can be attributed to reduced thermodynamic penalty of oxygen vacancies in the oxide structure associated with Sr segregation observed in the heterostructure. Our findings provide insights for the design of highly active catalysts for energy conversion and storage applications.

Feng, Zhenxing↗
DOE OSTI2013

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime. Specifically, we propose four research thrusts, namely Thrust 1: Identify the formation mechanism and evolution for dislocation loops with Burgers vector of a<100> and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Thrust 2: Identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered Martensitic structure under irradiation. Thrust 3: Evaluate the stability of nanometer sized Y- Ti-O based oxide dispersion strengthened (ODS) particles at high fluence/temperature. Thrust 4: Evaluate the extent to which precipitates form and/or dissolve as a function of irradiation temperature and dose, and how these changes are driven by radiation induced segregation and microchemical evolutions and determined by the initial microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Understanding the Irradiation Behavior of Zirconium Carbide

Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450{degree}C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800{degree}C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Modeling Solute Thermokinetics in LiCl-KCl Molten Salt for Nuclear Waste Separation (NEUP 10-418)

Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt at the eutectic composition (58 mol% LiCl, 42 mol% KCl), which is used for treating spent EBR-II fuel. The same process being used for EBRII fuel is currently being studied for widespread international implementation. The methods will focus on first-principles and first- principles derived interatomic potential based simulations, primarily using molecular dynamics. Results will be validated against existing literature and parallel ongoing experimental efforts. The simulation results will be of value for interpreting experimental results, validating analytical models, and for optimizing waste separation by potentially developing new salt configurations and operating conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Modeling Fission Product Sorption in Graphite Structures

The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗