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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 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Recent advances in automotive catalysis for NOx emission control by small-pore microporous materials

The ever increasing demand to develop highly fuel efficient engines coincides with the need to minimize air pollution originating from the exhaust gases of internal combustion engines. Dramatically improved fuel efficiency can be achieved at air-to-fuel ratios much higher than stoichiometric. In the presence of oxygen in large excess, however, traditional three-way catalysts are unable to reduce NOx. Among the number of lean-NOx reduction technologies, selective catalytic reduction (SCR) of NOx by NH3 over Cu- and Fe-ion exchanged zeolite catalysts has been extensively studied over the past 30+ years. Despite the significant advances in developing a viable practical zeolite-based catalyst for lean NOx reduction, the insufficient hydrothermal stabilities of the zeolite structures considered cast doubts about their real-world applicability. During the past decade a renewed interest in zeolite-based lean NOx reduction was spurred by the discovery of the very high activity of Cu-SSZ-13 (and the isostructural Cu-SAPO-34) in the NH3 SCR of NOx. These new, small-pore zeolite-based catalysts not only exhibited very high NOx conversion and N2 selectivity, but also exhibited exceptional high hydrothermal stability at high temperatures. In this review we summarize the key discoveries of the past ~5 years that lead to the introduction of these catalysts into practical application. The review first briefly discusses the structure and preparation of the CHA structure-based zeolite catalysts, and then summarizes the key learnings of the rather extensive (but not complete) characterisation work. Then we summarize the key findings of reaction kinetics studies, and provide some mechanistic details emerging from these investigations. At the end of the review we highlight some of the issues that are still need to be addressed in automotive exhaust control catalysis. Funding A.M.B. and I.L.G. would like to thank EPSRC for funding. F.G., C.H.F.P. and J.Sz. gratefully acknowledge financial support from the US Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy, Vehicle Technologies Program.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2014

DEVELOPMENT OF A HYDROGEN MORDENITE SORBENT FOR THE CAPTURE OF KRYPTON FROM USED NUCLEAR FUEL REPROCESSING OFF-GAS STREAMS

A novel new sorbent for the separation of krypton from off-gas streams resulting from the reprocessing of used nuclear fuel has been developed and evaluated. A hydrogen mordenite powder was successfully incorporated into a macroporous polymer binder and formed into spherical beads. The engineered form sorbent retained the characteristic surface area and microporosity indicative of mordenite powder. The sorbent was evaluated for krypton adsorption capacities utilizing thermal swing operations achieving capacities of 100 mmol of krypton per kilogram of sorbent at a temperature of 191 K. A krypton adsorption isotherm was also obtained at 191 K with varying krypton feed gas concentrations. Adsorption/desorption cycling effects were also evaluated with results indicating that the sorbent experienced no decrease in krypton capacity throughout testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

In situ self-ion (Fe + ) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation

Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used in situ self-ion irradiation (150 keV Fe + ) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe + irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10 15 ions·cm -2 (2 dpa) and 3.4 × 10 15 ions·cm -2 (4 dpa), using three damage dose rates (5 × 10 –5 , 3.3 × 10 –4 , and 2 × 10 –3 dpa·s -1 ). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Regulatory perspective of boiling water reactor peak reactivity credit in spent fuel storage and transportation - 14592

The U.S. Nuclear Regulatory Commission (NRC) performs reviews of applications for spent nuclear fuel transportation packages and storage casks to ensure pertinent safety regulations are met. Among other acceptance criteria, these regulations require that the package be subcritical under a variety of normal, off-normal and accident conditions. Currently, all spent nuclear fuel transportation packages and storage casks assume fresh fuel (i.e., the fuel is unirradiated) in their criticality safety analyses for boiling water reactor (BWR) fuels. Reviews using this conservative assumption are less complex. The fresh fuel assumption for BWR fuel does not allow the presence of gadolinium (often called the 'fresh fuel no-Gad' assumption), an integral burnable absorber present in nearly all BWR fuel assemblies. More recently, applicants began requesting credit for the reactivity reduction due to depletion. This credit is commonly referred to as 'burnup credit' (BUC). As of today, NRC recommends that only analyses for Pressurized Water Reactor (PWR) spent fuel implement BUC. This is due to the fact there is limited directly applicable data available to benchmark codes for depletion and reactivity calculations for BWR BUC analyses, and determining the most reactive irradiation conditions is much less straightforward than for PWR BUC analyses. Similar to what was done for PWR BUC, the NRC staff and its contractors have begun identifying and prioritizing significant technical issues so that a technical basis for the allowance of BWR BUC can be developed. BWR fuel exhibits a peak in reactivity where initial gadolinium is depleted. Crediting burnup at this point is generally referred to as using a 'peak reactivity method'. This is a conservative approach and a method routinely used in BWR spent fuel pool criticality analyses. NRC is implementing a two-phased approach in investigating BWR BUC. Phase 1 investigates peak reactivity credit and Phase 2 evaluates BUC at a typical discharge exposure. This paper discusses peak reactivity credit and its history in regulatory analyses, challenges the NRC expects in reviewing an application requesting peak BWR reactivity credit for a spent fuel transportation package or storage cask, the work that is being performed by NRC contractors to address these challenges, and future work on peak reactivity credit and full BWR BUC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Preliminary Study on TRISO Fuel Cross Section Generation

Cross section self-shielding methodologies for TRISO fuel were assessed to provide accurate multigroup cross sections for a high-fidelity reactor physics code so that the code is able to accurately model and simulate advanced reactors with TRISO fuel. Initially, the two existing methodologies (the SCALE method and the Sanchez-Pomraning method) were studied and implemented to MC2-3 for detailed performance tests. Additionally, a new spatial self-shielding method, named the iterative local spatial self-shielding (ILSS) method, for particulate fuels was developed based on the disadvantage factor and implemented to MC2-3 as well. The new method approximately accounts for the effect of randomly distributed particles on the particle shadowing effect using a homogenized compact region surrounding a particle of interest at the center. The self-shielded cross sections of the particle at the center are determined iteratively since the cross sections of the homogenized compact region are calculated using them. For the energy range above 100 keV where the fuel-to-moderator ratio is more important than the random distribution of particles, a single particle unit-cell model is used by preserving the average amount of moderator per fuel particle in the system. The three self-shielding methods implemented in MC2-3 were tested using numerical benchmark problems made based on fuel compact problems of a prismatic-type very high temperature reactor. Test results indicated that the ILSS method produced slightly better results than the SCALE and Sanchez-Pomraning methods, compared to the Serpent-2 Monte Carlo results obtained with 25 independent random particle configurations. The SCALE and Sanchez-Pomraning methods tend to underestimate the heterogeneity effect by 150 and 100 pcm, respectively, while the new ILSS method overestimates the heterogeneity effect by 70 pcm. In future, the new self-shielding method will be extended to perform pebble calculations and compare results with those from the SCALE and Sanchez-Pomraning methods. Furthermore, the new method will be optimized for practical applications to on-the-fly resonance treatment for lattice or whole-core calculations for advanced reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Twenty years of dry interim storage of spent nuclear fuel (SNF) in Germany- History and developments - 15575

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection. In 1983 a license to store spent nuclear fuel (SNF) and later expanded for vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its actual form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic casks. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years starting with the first cask emplacement. As no opening of the cask is intended during the storage period the condition of the inventory and the cask before loading has to be known in detail. This requires for instance detailed knowledge of the history of the SNF loaded as well as supervision of the manufacturing process of the cask. Within the last twenty years several mishaps influenced the regulatory point of view. During a loading campaign in a boiling water reactor traces of pool water remained in the gasket slot and during the drying procedure crystalline boron salts prevented the required tightness. After that event a long-term experiment was started and is still going on which studies the effects of water and minor damages to the gasket. As well changes in the loading procedure have been introduced. The pressure switch used to monitor the pressure in the space between primary and secondary lid has a self surveillance function. This indicated a malfunction 23 times in the last twenty years. Examinations of all events indicated no systematic failure, but a weak point seems to be the plug-in connection. The manufacturer introduced some modifications to improve the system. Introduction of FEM methods in safety analysis to amend the existing methods of testing allowed a new view on and in the cask safety analysis. Progress in this field is closely connected to the available computing capacity and stipulates experts to ongoing improvement. After expiring of the storage license which will be the case for the first cask loaded in 1992 and stored in Ahaus 2032 an extension or a new license can only be granted if the at this time relevant requirements are met. This will require detailed knowledge about the happenings and accidents during the previous storage period. All experience of the periodic safety assessment (PSA) which started in 2013 and accompanying experiments conducted till now and during the licensed storage time will be needed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Planning Document for an NBSR Conversion Safety Analysis Report

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted. This report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.

42 ENGINEERING↗
DOE OSTI2020

ASSESSING FUEL PROPERTY EFFECTS ON CAVITATION AND EROSION PROPENSITY USING A COMPUTATIONAL FUEL SCREENING TOOL

To advance compression ignition combustion strategies, researchers have evaluated fuel property effects and their impact on achieving higher efficiencies and lower emissions levels relative to current capabilities. Within the Department of Energy's Co-Optima initiative, there has been a recent focus on understanding the influence of fuel properties on fuel injection performance. To help identify candidate fuels that can meet desired injector performance metrics, a computational fuel screening tool is under development that can link fuel properties with the tendency of a given fuel to cavitate and lead to cavitation-induced erosion. In the initial development of this tool, five liquid fuel properties were selected to represent candidate fuels, namely density, viscosity, vapor pressure, surface tension, and heat of vaporization. A design of experiments methodology was employed to generate a set of pseudo fuel cases, which can represent the main effects and interactions among the fuel properties and be related to cavitation erosion predictions. Large eddy simulations were performed using a mixture modeling approach to predict the cavitation and erosion propensity of these pseudo-fuels in terms of the mean fuel vapor mole fraction and stored impact energy from repeated cloud collapse events. The low order regression models generated from this study revealed the importance of liquid fuel density on cavitation formation, whereas liquid viscosity was found to have a strong negative correlation with erosion severity. The surrogate models were then used in the fuel screening tool to rank the cavitation and erosion tendency of four candidate single-component fuels: methyl decanoate, isooctane, ethanol and n-dodecane. While the fuel screening tool was not able to quantitatively predict the cavitation and erosion response metrics, the tool was able to accurately rank the relative cavitation and erosion propensity of the four fuels. Overall, ethanol and iso-octane were observed to produce the highest levels of cavitation and erosion, respectively.

cavitation-induced erosion↗
DOE OSTI2018

Fission gas bubbles and recrystallization-induced degradation of the effective thermal conductivity in U-7Mo fuels

We have developed a mesoscale model to calculate the degradation of the effective thermal conductivity in irradiated U-Mo alloys caused by the fission-induced gas bubbles and recrystallization. The phase-field approach is employed to generate the grain microstructures of U-7Mo fuels with intra- and inter-granular gas bubbles. Based on the phase-field microstructures, the thermal conductivities of U-7Mo as a function of the fission density can be predicted by the developed mesoscale model. The predicted values of effective thermal conductivities are consistent with available experimental data. Results show that the effective thermal conductivity decreases rapidly with recrystallization compared to the one prior to recrystallization, which can be attributed to the sudden increase of grain boundary densities and corresponding intergranular gas bubbles at high fission densities. Smaller grain size fuel structure has a lower thermal conductivity at the same fission density due to the increased grain boundary density. The current study can provide a better understanding of the fission-induced degradation mechanism of the thermal conductivity in U-Mo fuels.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Holistic Fuel Cell Electric Vehicle/Hydrogen Station Optimization Model (CRADA CRD-18-00745 Final Report)

The National Renewable Energy Lab (NREL), Argonne National Lab (ANL), Sandia National Lab (SNL), and Frontier Energy, Inc comprise a team that will perform a review of currently available models that can simulate both a hydrogen station and a fuel cell electric vehicle (FCEV), and whose owners are willing to make them available freely. We will then down select to a single model or model pair, secure agreements for their free use, and validate the resulting model or coupled models. For example, we may evaluate an existing 1-D hydrogen fueling model developed by Kyushu University for the Japanese New Energy and Industrial Technology Development Organization (NEDO). The validation will take place using data from various sources, potentially including the European Joint Research Center (JRC) HyTransfer project, SAE testing performed by Powertech to validate SAE J2601, and whole-station validation using the NREL’s Hydrogen Infrastructure Testing and Research Facility (HITRF). The team will make the model and all the validation data open to the public at the end of the project.

1-D hydrogen fueling model↗
DOE OSTI2026

Ablating Pellets for Areal Density Symmetry Control in Indirectly Driven Inertial Confinement Fusion Target Designs

Asymmetry of the compressed fusion fuel configuration, as characterized by a variation in areal density, in inertial confinement fusion (ICF) implosions is a known performance limitation. While ignition has been achieved on the National Ignition Facility (NIF), implosion symmetry was not perfect, and subsequent experiments that have extended fusion performance still struggle with symmetry control and would, in principle, perform much better if the areal density uniformity of the fuel was improved by eliminating thin regions in the DT (Deuterium-Tritium) fuel that develop by the time of peak compression. In indirect drive (IDD), altering the time-dependent laser cone fraction and cross-beam energy transfer are the usual tactics for controlling implosion asymmetry in IDD target designs, but it may be that those tools have been pushed to their limits. In this Letter, we present a new tactic for controlling the implosion using satellite ablating pellets to manipulate the time-dependent x-ray field around an ICF capsule. In theory and in simulations the action of these pellets is quite effective at reducing the swing of x-ray drive from waist hot to pole hot, thus achieving a more one-dimensional fuel configuration at peak compression. Simulations show that a factor of ∼2 in yield performance can be obtained via this technique, when applied to past record implosions on the NIF. As a result, some speculative alternate uses of the proposed target concept are also mentioned in the conclusion of this Letter.

Implosion symmetry↗
DOE OSTI2013

MR-Guided High-Intensity Focused Ultrasound Ablation of Breast Cancer with a Dedicated Breast Platform

Optimizing the treatment of breast cancer remains a major topic of interest. In current clinical practice, breast-conserving therapy is the standard of care for patients with localized breast cancer. Technological developments have fueled interest in less invasive breast cancer treatment. Magnetic resonance-guided high-intensity focused ultrasound (MR-HIFU) is a completely noninvasive ablation technique. Focused beams of ultrasound are used for ablation of the target lesion without disrupting the skin and subcutaneous tissues in the beam path. MRI is an excellent imaging method for tumor targeting, treatment monitoring, and evaluation of treatment results. The combination of HIFU and MR imaging offers an opportunity for image-guided ablation of breast cancer. Previous studies of MR-HIFU in breast cancer patients reported a limited efficacy, which hampered the clinical translation of this technique. These prior studies were performed without an MR-HIFU system specifically developed for breast cancer treatment. In this article, a novel and dedicated MR-HIFU breast platform is presented. This system has been designed for safe and effective MR-HIFU ablation of breast cancer. Furthermore, both clinical and technical challenges are discussed, which have to be solved before MR-HIFU ablation of breast cancer can be implemented in routine clinical practice.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

U-10Mo Monolithic Fuel Qualification Plan

The Material Management and Minimization (M3) Program’s primary objective within the U.S. Department of Energy/National Nuclear Security Administration is to achieve permanent threat reduction by minimizing and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project, is pursuing a fuel qualification and licensing effort focused on converting high-performance research reactors in the United States from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. The body of this document is focused on defining the activities required for generic qualification of high density, LEU fuel that then allows reactor conversions to proceed for the four high-performance research reactors and the critical assembly that will operate using the U-10Mo monolithic fuel design (see Subsection 1.3): MITR, MURR, NBSR, ATR, and ATRC. Fuel qualification in this report means generic fuel acceptance of the U-10Mo monolithic fuel form by the NRC for use in reactor conversions in the United States, which allows the fuel to be used in subsequent reactor specific licensing requests. The U 10Mo Monolithic Fuel Research, Development, and Qualification Plan utilizes the functions and requirements of the USHPRR Project that was established in the Functions and Requirements Document (F&RD) [1] and expands on these requirements to ensure that planned tests have traceable results that will ensure the requirement has been met. The methods by which data will be collected to show that these requirements have been met are described in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Characterization of Fresh Nuclear Fuel using Time-Of-Flight Neutrons

Neutrons offer excellent characterization capabilities for nuclear fuels and can provide non-destructive insight on micro structure, chemistry and crystallographic phase. Measurements are possible at reactors or pulsed sources, but the inherent energy resolution implicit at pulsed sources offers unique opportunities. Recent advances in 2D neutron detection enabled new capabilities of material characterization using neutron imaging and tomography for nuclear materials. Moreover, time-of-flight (TOF) neutron measurements can enable isotope specific tomography by deriving isotope densities from transmission spectra exhibiting neutron resonances, which are recorded for each pixel. Tomographic algorithms allow 3D reconstructions of the spatial distribution of isotopes, density and features in typical fuel pellet samples. The ability to select radiographs for different energies allows characterizing samples which are opaque for thermal neutrons, such as enriched fuels or fuels containing plutonium, by selecting epi-thermal neutron energies and by knowledge of the total neutron cross sections of the constituents, isotopic enrichment levels can be determined. Nuclear cross sections are typically not calculated directly and databases such as the Evaluated Nuclear Data File (ENDF) database provide this information with cross sections determined by standardized experimental measurements, such as neutron transmission measurements for individual isotopes. Utilizing the information in the ENDF database, the 'inverse' problem can be solved by assuming cross section parameters to be known parameters and areal densities can be determined. This work presents recent advances at the Los Alamos Neutron Science Center (LANSCE) facility at Los Alamos National Laboratory (LANL), which provides the infrastructure to handle radioactive samples, in the development of fuel characterization with TOF neutrons. Examples include recent work on prototype accident tolerant nuclear fuels (ATF). The potential application of these techniques for nondestructive examination of irradiated fuel is under consideration. Non-destructive access to materials features for specimens embedded in containers, particularly for hazardous samples, such as a radioactive nuclear fuel, is indispensable to characterize and ultimately license new fuel forms. Energy-resolve neutron imaging is a new tool for this research field. The results reported here present a new imaging technique, setting a milestone in the capabilities of measuring non-destructively absolute quantities of physical properties. The enabling physics behind this new technique are nuclear resonances. We have demonstrated how absolute areal densities can be extracted from radiographs utilizing pulsed neutron sources. This technique provides information that is otherwise difficult and costly to obtain. In the case of nuclear fuels, the results from energy-dispersive neutron imaging may guide destructive examination of irradiated fuel. This summery does not describe recent analysis, which is ongoing. Particularly, the implementation of calibration standards has improved the determination of enrichment levels without additional calibration. The analysis of the samples was performed with full tomographic reconstructions for isotope specific data and presents a first approach in absolute isotope specific 3D reconstruction. This work was predominantly funded by and received substantial support from the Fuel Cycle R and D program at Los Alamos National Laboratory. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Enabling the Billion-Ton Bioeconomy

The United States is rich in non-food ‎biomass that can fuel the development of a thriving ‎bioeconomy where renewable and sustainable resources power cars and planes instead of petroleum. The ‎transportation and aviation industry is actively seeking ways to reduce its carbon footprint by powering planes with solid municipal waste, woody biomass, purpose-grown crops, and ‎algae. Watch this short video to learn how biomass is being used to make our country greener, provide new employment opportunities, and reduce our dependence on foreign oil.

09 BIOMASS FUELS↗
DOE OSTI2015

USABC Development of 12 Volt Battery for Start-Stop Application: Preprint

Global automakers are accelerating the development of fuel efficient vehicles, as a part of meeting regional regulatory CO2 emissions requirements. The micro hybrid vehicles with auto start-stop functionality are considered economical solutions for the stringent European regulations. Flooded lead acid batteries were initially considered the most economical solution for idle-stop systems. However, the dynamic charge acceptance (DCA) at lower state-of-charge (SOC) was limiting the life of the batteries. While improved lead-acid batteries with AGM and VRLA features have improved battery longevity, they do not last the life of the vehicle. The United States Advanced Battery Consortium (or USABC, a consortium of GM, Ford, and Chrysler) analyzed energy storage needs for a micro hybrid automobile with start-stop capability, and with a single power source. USABC has analyzed the start-stop behaviors of many drivers and has developed the requirements for the start-stop batteries (Table 3). The testing procedures to validate the performance and longevity were standardized and published. The guideline for the cost estimates calculations have also been provided, in order to determine the value of the newly developed modules. The analysis effort resulted in a set of requirements which will help the battery manufacturers to develop a module to meet the automotive Original Equipment Manufacturers (OEM) micro hybrid vehicle requirements. Battery developers were invited to submit development proposals and two proposals were selected for 50% cost share with USABC/DOE.

12 V START-STOP↗