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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 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Electrocatalysts, and fuel cells containing them

Provided are electrocatalysts, fuel cells, methods of making fuel cells, and methods of generating an electric current, each featuring a platinum (Pt)-containing substrate in contact with an aqueous solution comprising Pb2+. Electrocatalysts of the invention are formed via underpotential deposition (UPD) when a trace amount of Pb2+ is present in the electrolyte of a half anodic cell for oxidizing formic acid using Pt as the anode. Surprisingly, the UPD process dramatically enhances the activity of formic acid oxidation, at least as much as 10-fold compared with palladium (Pd) black. In an embodiment, the electrocatalyst comprises a Pt-containing substrate, a submonolayer of lead (Pb) adsorbed onto the Pt-containing substrate, and an aqueous solution comprising Pb2+, wherein the concentration of Pb2+ in the aqueous solution is 10 to 500 μm.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2021

Protective coating for lithium-containing electrode and methods of making the same

A lithium-containing electrode with a protective coating and lithium-containing electrochemical cells including the same are provided herein. The protective coating has a first layer including a first fluoropolymeric matrix and Li—F compounds and a second layer including a second fluoropolymeric matrix. Methods of preparing the protective coating on the lithium-containing electrode by applying a first fluoropolymer and/or a first fluoropolymer precursor and a second fluoropolymer and/or a second fluoropolymer precursor are also provided herein.

Xiao, Xingcheng↗
DOE OSTI2022

Resource loading system and method for use in atmosphere containment scenarios

The invention provides a system for preventing fluid exchange between the interior and exterior of containment enclosures such as process-, hazard-, and research-enclosure systems generally, gloveboxes, containment systems, isolation systems, confinement systems, cleanrooms, negative air systems, and positive air system areas while simultaneously providing material transfer into and out of the enclosures. The invention also provides a method for transporting material into or out of a containment structure.

Dilday, Jr., Daniel R.↗
DOE OSTI2022

Boron injection system for controlling a nuclear reaction by delivering boron into a containment vessel

A nuclear power system includes a reactor vessel that includes a reactor core mounted within a volume of the reactor vessel. The reactor core includes one or more nuclear fuel assemblies configured to generate a nuclear fission reaction. The nuclear power system further includes a containment vessel sized to enclose the reactor vessel such that an open volume is defined between the containment vessel and the reactor vessel. A boron injection system is positioned in the open volume of the containment vessel and includes an amount of boron sufficient to stop the nuclear fission reaction or maintain the nuclear fission reaction at a sub-critical state. The boron injection system is positioned to deliver the amount of boron into the open volume.

Linik, Larry↗
DOE OSTI2013

A Review of Tungsten Heavy Alloy Utilization in Isotope Transport Containers - 13380

A common requirement for radioisotope transport containers is that they provide both durable and efficient shielding of penetrating gamma radiation. This is the case for transport of both spent nuclear fuel as well as intentionally created radioisotopes for medical or other uses. Tungsten heavy alloy (WHA) provides a unique engineering property set for such shielding - easily surpassing more commonly used lead alloys in both strength and attenuation. This family of alloys contains typically 90-98 wt.% W in combination with transition metals such as Ni and Fe. WHA is manufactured in near net shape blanks by liquid phase sintering of compacted powder shapes to full metallurgical density parts. This powder metallurgy approach is described in its ability to provide excellent material utilization and affords efficient manufacturing of various shapes required for gamma shields or collimators. WHAs offer very high density (approaching 19 g/cc) in combination with relatively high thermal conductivity, low thermal expansion, ambient corrosion resistance, and can be provided with mechanical properties comparable to many medium carbon steels. As such, they can be machined to complex, damage resistant geometries using common metal cutting tools and methods. WHA additionally provides a lower toxicity alternative to Pb- or U-based gamma shielding. Given the specialty nature of WHA, specific metallurgical characteristics are reviewed to assist shielding designers who may otherwise encounter difficulties locating important alloy selection and fabrication details. Contained within this materials and applications overview are guidelines for WHA component design, alloy selection, and practical machining, finishing, and assembly considerations. The microstructure of WHA is that of a metal matrix composite. This factor has specific implications in the design of components for stress service as well as their protection in the presence of electrolytes. WHA is also discussed in the broader context of materials compatibility, as it is rarely used in isolated monolithic shapes. Alloy selection for new applications is often made primarily on the basis of density. An alternative strategy to this selection approach is presented which proposes that mechanical requirements for a given shielding use be the primary selection criterion over density. Standard commercial grades of WHA for radiation shielding are defined by specifications such as AMS-T-21014 and ASTM B777. These specifications define 4 density classes of as-sintered WHA. Compositional options, as well as post-sinter processing of WHA, are discussed for shielding components that must exhibit higher levels of ductility or very low magnetic permeability. In addition to the mechanical advantages over Pb-based shielding, the higher linear attenuation of energetic photons for various grades of WHA (as calculated by the NIST XCOM routine) are presented for selected photon energies of interest to illustrate the shield volume reduction generally possible through the use of tungsten-based shielding. While W provides inefficient attenuation of neutrons in a mixed radiation environment, its secondary role in shielding gamma radiation produced as a result of neutron capture is also described. (authors)

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

Use of a Shielded High Resolution Gamma Spectrometry System to Segregate LLW from Contact Handleable ILW Containing Plutonium - 13046

Dounreay Site Restoration Limited (DSRL) have a number of drums of solid waste that may contain Plutonium Contaminated Material. These are currently categorised as Contact Handleable Intermediate Level Waste (CHILW). A significant fraction of these drums potentially contain waste that is in the Low Level Waste (LLW) category. A Canberra Q2 shielded high resolution gamma spectrometry system is being used to quantify the total activity of drums that are potentially in the LLW category in order to segregate those that do contain LLW from CHILW drums and thus to minimise the total volume of waste in the higher category. Am-241 is being used as an indicator of the presence of plutonium in the waste from its strong 59.54 keV gamma-ray; a knowledge of the different waste streams from which the material originates allows a pessimistic waste 'fingerprint' to be used in order to determine an upper limit to the activities of the weak and non-gamma-emitting plutonium and associated radionuclides. This paper describes the main features of the high resolution gamma spectrometry system being used by DSRL to perform the segregation of CHILW and LLW and how it was configured and calibrated using the Canberra In-Situ Object Counting System (ISOCS). It also describes how potential LLW drums are selected for assay and how the system uses the existing waste stream fingerprint information to determine a reliable upper limit for the total activity present in each measured drum. Results from the initial on-site commissioning trials and the first measurements of waste drums using the new monitor are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Performance of Large Neutron Detectors Containing Lithium-Gadolinium-Borate Scintillator

This paper describes the development and testing of a neutron counter, spectrometer, and dosimeter that is compact, efficient, and accurate. A self-contained neutron detection instrument has wide applications in health physics, scientific research, and programs to detect, monitor, and control strategic nuclear materials (SNM). The 1.3 liter detector head for this instrument is a composite detector with an organic scintillator containing uniformly distributed {sup 6}Li{sub 6}{sup nat}Gd{sup 10}B{sub 3}O{sub 9}:Ce (LGB:Ce) microcrystals. The plastic scintillator acts to slow impinging neutrons and emits light proportional to the energy lost by the neutrons as they moderate in the detector body. Moderating neutrons that have slowed sufficiently capture in one of the Lithium-6, Boron-10, or Gadolinium-157 atoms in the LGB:Ce scintillator, which then releases the capture energy in a characteristic cerium emission pulse. The measured captured pulses indicate the presence of neutrons. When a scintillating fluor is present in the plastic, the light pulse resulting from the neutron moderating in the plastic is paired with the LGB:Ce capture pulse to identify the energy of the neutron. About 2% of the impinging neutrons lose all of their energy in a single collision with the detector. There is a linear relationship between the pulse areas of this group of neutrons and energy. The other 98% of neutrons have a wide range of collision histories within the detector body. When these neutrons are 'binned' into energy groups, each group contains a distribution of pulse areas. This data was used to assist in the unfolding of the neutron spectra. The unfolded spectra were then validated with known spectra, at both neutron emitting isotopes and fission/accelerator facilities. Having validated spectra, the dose equivalent and dose rate are determined by applying standard, regulatory damage coefficients to the measured neutron counts for each energy bin of the spectra. Testing at the Los Alamos Neutron Science Center (LANSCE), Edwards Accelerator Laboratory (EAL) at Ohio University and the Radiation Center at University of Massachusetts-Lowell has demonstrated that the instrument can measure neutrons and their spectra over the range between 0.8 MeV and 150 MeV with an uncertainty of only ± 8%. An independent test of the LGB:Ce neutron spectrometer was conducted by a US Defense Threat Reduction Agency (DTRA) team at the Idaho National Laboratory (INL). The results of this evaluation showed that the neutron spectrometer accurately identified bare radioactive isotopes by their spectra. Further, masking and shielding materials alter those spectra in predictable ways that permit an extrapolation from the observed spectra back to the identity of the isotopic spectrum. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Design and Licensing of Irradiated Uranyl Nitrate Liquid Containers and Applicable Loading and Packaging System - 18481

Irradiated highly enriched uranyl nitrate liquid (HEUNL) solutions are typically residual materials resulting from the processing of uranium target materials in the production of radioisotopes. Traditionally, these residual materials have been processed and stored at the site of origin, due to the lack of an option for packaging, transportation and disposition. Modern alternatives to process, stabilize, package and dispose of these materials have been developed, but until recently there has not been a viable Type B package solution that could be used to ship these materials in sufficient quantities to justify the cost of an off-site disposition strategy. For this reason, most facilities that generate irradiated HEUNL materials have implemented on-site storage tanks or containers, resulting in the significant accumulation of HEUNL inventories worldwide. This paper provides an overview of the design and licensing activities for the development of sealed irradiated HEUNL transport containers to permit safe transportation of this material in the NAC Legal Weight Truck (NAC-LWT) cask. A description of the robust HEUNL container, licensing and fabrication is also presented. In addition, NAC has developed a specialized loading system to optimize the process to load, package and un-package HEUNL materials to minimize worker exposure during loading, and execute the packaging process efficiently for safe transport and retrieval. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Richard Hodes Memorial Award Lecture - Development and Certification of New Type-B Containers for the Recovery of High-Activity Sealed Sources - 18710

The US Department of Energy (DOE) National Nuclear Security Administration (NNSA) Off-Site Source Recovery Program (OSRP) has successfully completed a project to design, test, and certify two new Type-B transportation packages. Since about 2011 a shortage of these containers has significantly constrained the recovery and disposal of commercially licensed high activity radioactive sealed sources. These new packages will enable NNSA/OSRP to transport a majority of the high-activity Cs-137 and Co-60 devices encountered in the coming years and decades. The 435-B is an unshielded, leak-tight container, which can be used to ship common shielded devices, such as the teletherapy source drawers and irradiators that contain high-activity Co-60 and Cs-137 sources. The 380-B is a much larger, leak-tight, lead-shielded Type B cask capable of transport by truck, rail, ship, or air. Additionally, to encourage and facilitate the development of transportation capacity, DOE/NNSA intends to make the design documentation available at no cost to commercial vendors. (authors)

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

Recovery and Recrystallization Behavior of Nb-Containing Fe-Cr-Al Alloys

Nuclear-grade wrought Fe-Cr-Al alloys are currently under development as promising accident-tolerant fuel cladding materials for light water reactors (LWRs) due to their superior oxidation and corrosion resistance in high temperature steam environments when compared with current commercial Zr-based alloys. Sufficient amounts of Cr and Al additions in Fe-Cr-Al alloys provide the improved surface protection effects at both the service temperature and elevated temperatures in the accident case scenario. Minor alloying additions (such as Nb in the current study) also promote the second-phase precipitates to strengthen the alloys; e.g. sub-micron size Fe{sub 2}Nb-type Laves phase precipitates are likely to disperse at both inter- and trans-granular in BCC-Fe matrix after applying proper heat treatments. In LWRs, a seamless tube product is required for the fuel cladding which typically has ∼ 4 m length with an outside diameter of ∼ 10 mm and a wall thickness of less than 0.5 mm, and it imposes considerable challenges for tube fabrication process (e.g. tube drawing) of Fe-Cr-Al alloys. Research has been performed to achieve a balance between high-temperature performance and tube fabricability in newly developed Fe-Cr-Al alloys. BCC-Fe materials with Cr and Al additions typically suffer from poor ductility because of relatively high ductile-brittle transition temperature. In this sense, the microstructural stability of refined grain or subgrain structure, even at high temperature, is critical to maintain the deformability of the alloys during tube fabrication process, as well as sufficient mechanical properties at service condition. In this study, the recovery and recrystallization behaviors of Nb-containing Fe-Cr-Al alloys are studied. A specific focus was to evaluate the stability of the subgrain structure of the deformed Fe-Cr-Al alloys, and seek optimized processing parameters for seamless tube fabrication. The present results clearly demonstrated that the Nb addition stabilizes the refined subgrain microstructures of Fe-Cr-Al alloys below 700 deg. C due to Laves phase precipitations along sub-boundaries when compared with the Nb-free alloys, while maintaining sufficient strengths needed for fuel cladding in LWRs. At the same time, the results also showed that Nb-containing alloys can be softened through suitable annealing to improve the processability of tube drawing process. Detailed characterization efforts of subgrains and Laves phase precipitates are currently in progress. Completion of this work will deepen the understanding of the recovery and recrystallized behavior of Nb-containing Fe-Cr-Al alloys and provide guidelines to tube drawing process in the alloys. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2020

GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} for Fukushima Daiichi Water Treatment Secondary Wastes - 20212

The Japanese government is supporting development work implemented jointly by Veolia subsidiaries Kurion Japan, K.K., Veolia Nuclear Solutions (VNS), Inc., and Veolia Nuclear Solutions Federal Services, LLC for treating radioactive waste generated from Fukushima Daiichi Nuclear Power Station (NPS) water treatment using the GeoMelt{sup TM} In-Container Vitrification (ICV){sup TM} technology. The initial work consisted of glass formulation and engineering-scale testing which was completed in 2018, in the frame of an IRID (International Research Institute for Nuclear Decommissioning) program as part of a project subsidized by Japan's Ministry of Economy, Trade and Industry (METI). The Fukushima Daiichi NPS Mid- and Long-Term Road-map requires investigation of methods to stabilize solid wastes (and to immobilize radioisotopes in the wastes) generated as a result of emergency response and decommissioning activities. Cooling water treatment has resulted in a significant amount of solid and slurry secondary wastes (mostly adsorbents and ion-exchange materials) which will require processing at some point. GeoMelt{sup R} ICV{sup TM} is a joule-heated melter technology which uses a refractory-lined single-use container combining the melter and disposal container. There is no pouring required nor concerns with refractory corrosion which allows the process to accommodate a wide range of waste chemistries and high waste loadings. The testing described here consisted of three engineering-scale melts, each processing between 212 kg and 240 kg of waste simulants, glass formers, and non-radioactive cesium (Cs) and strontium (Sr) tracers. Continuous isokinetic stack sampling of off-gas emissions was performed for each test in order to calculate Cs and Sr retention in the glass wasteform. Single-pass retention of Cs in the final glass wasteform ranged from 91.46 to 99.30%, and single-pass retention of Sr ranged from 99.76 to 100%. Planned particulate recycle will increase these retention levels. Melt 1 processed a mixture of KUR-EH (a zeolite-based ion-exchange material), simulated Advanced Liquid Processing System (ALPS) Carbonate and Iron Slurries, and glass additives. Melt 2 processed a mixture of KUR-EH, KUR-TSG (a titanate-based adsorbent), and glass additives. Melt 3 processed a mixture of KUR-EH, simulated barium sulfate/iron ferrocyanide sludge (AREVA sludge), and glass additives. Waste loadings for these melts ranged from 70 weight percent (wt%) to 82 wt%. Vitrification produces a waste form much denser than the stored water treatment secondary waste wastes, resulting in significant volume reduction. Volume reductions for the three tests ranged from 74 to 79 vol%. Vitrification produces a chemically durable wasteform. Pacific Northwest National Laboratory (PNNL) tested three glass samples from each engineering-scale melt) by the Materials Characterization Center 1 (MCC-1) test, an international standard leach test of the chemical durability of nuclear waste glasses. PNNL also obtained one U.S. reference glass (EA Glass) and two Japan reference glasses P0798) and tested these under the same MCC-1 conditions (90 deg. C, 10 m-1, DIW, and 7, 14, 28-day) as the GeoMelt{sup R} ICV{sup TM} glasses. The GeoMelt{sup R} ICV{sup TM} glasses exhibited lower total normalized releases and 14- to 28-day normalized release rates than the three reference glasses. These results suggest that the GeoMelt{sup R} ICV{sup TM} glasses have durabilities on par with high-level waste glasses under standard test conditions. Post-melt process sampling and analysis indicated no Cs migration into the melter refractory materials and very little deposition of Cs or Sr onto the melter hood or off-gas piping internals. The results of the testing indicated good Cs retention in the glass, high volume reduction and waste loadings, and excellent chemical durability. These factors are important to minimize treatment costs and to protect workers and the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

On Temperature Limits for the Outer Surface of a Container in a Disposal Facility - 20313

The temperature in a disposal facility for high-level radioactive waste (HLRW) is an issue for safety. When the recommencement of the search for and selection of a site for a disposal facility for HLRW in Germany was stipulated by the Site Selection Act (StandAG 2017) in 2017, a precautionary temperature limit of 100 deg. C on the outer surface of the containers with high-level radioactive waste in the disposal facility section was set. This precautionary temperature limit shall be applied in preliminary safety analyses provided that the 'maximum physically possible temperatures' in the respective host rocks have not yet been determined due to pending research. Therefore, this issue is addressed and discussed in this paper, contributing to 'pending research' by a review of the literature. This paper briefly discusses a few examples of thermohydraulic, mechanical, chemical and biological processes in a disposal facility, because temperature limits are derived based on safety impacts regarding THMCB-processes. The temperature-dependent processes have been extracted from databases for features, events and processes (FEP-databases). Furthermore, it is discussed if the feasibility to retrieve and recover HLRW is hampered at high temperatures. A design temperature concerning single components of a disposal facility for the preservation of their features can be derived when a safety concept is established. However, the interactions of all relevant processes in a disposal concept must be considered to determine a specific temperature limit for the outer surface of the containers. Therefore, applicable temperature limits may vary for particular safety and disposal concepts in the following host rocks: rock salt, clay stone and crystalline rock. Technical solutions for retrieval and design options for recovery seem to be viable up to temperatures of 200 deg. C with different, sometimes severe, downsides according to expert judgement. Temperature limits regarding the outer surface of the containers can be derived specifically for each safety concept and design of the disposal facility in a host rock. General temperature limits without reference to specific safety concepts or the particular design of the disposal facility may narrow down the possibilities for optimisation of the disposal facility and could adversely affect the site selection process in finding the best suitable site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Superlubricity coating containing carbon nanotubes

A method for producing a structure containing an array of MWCNTs on a metal substrate, comprising: (i) subjecting a metal substrate to a surface oxidation process at a first elevated temperature in an oxygen-containing atmosphere and under a first reduced pressure; (ii) subjecting the metal substrate to a surface reduction process at a second elevated temperature in a reducing atmosphere and under a second reduced pressure of at least 0.01 atm and less than 1 atm to result in reduction of the surface of said metal substrate, wherein the reducing atmosphere contains hydrogen gas; (iii) subjecting the metal substrate to a third reduced pressure of no more than 0.1 atm; and (iv) contacting the metal substrate, while at the third reduced pressure and under an inert or reducing atmosphere, with an organic substance at a third elevated temperature for suitable time to produce the MWCNTs on the metal substrate.

Ihala Gamaralalage, Chanaka Kapila↗
DOE OSTI2015

Assessment of Residual Stresses in 3013 Inner and Outer Containers and Teardrop Samples

This report is an assessment performed by LANL that examines packaging for plutonium-bearing materials and the resilience of its design. This report discusses residual stresses in the 3013 outer, the SRS/Hanford and RFETS/LLNL inner containers, and teardrop samples used in studies to assess the potential for SCC in 3013 containers. Residual tensile stresses in the heat affected zones of the closure welds are of particular concern.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Containment margins in FFTF for postulated failure of in-vessel post-accident heat removal

This study was performed to complete the assessment of potential melt-through consequences for the postulated failure of post-accident heat removal following a hypothetical core disruptive accident (HCDA) in the FFTF. The SPRAY, SOFIRE, CACECO, HAA-3B, and COMRADEX codes were used to determine potential site boundary radiation doses. Parametric comparisons were made to evaluate the effects of natural hydrogen-oxygen recombination, reactor cavity liner failures and containment coolers. Containment pressure relief, vent/purge options, and the effects of exhaust filtration were also included in the study. The potential site boundary radiation exposures for all options considered were within 10 CFR 100 guideline values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Insight Into Sulfur‐Containing Additive to Boost Anti‐Oxidation Ability of the Ether‐Based Electrolyte for Sodium‐Ion Full Batteries

Ether-based electrolytes are considered as promising candidates for sodium-ion batteries (SIBs) due to their high ionic conductivity and good compatibility with hard carbon (HC) anode. However, they suffer from poor anti-oxidative ability with unstable cathode electrolyte interface, inducing successive electrolyte decomposition and rapid capacity fading. Herein, the sulfur-containing additive (1,3-propanediol cyclic sulfate, PCS) is introduced to boost the stability of the electrode-electrolyte interface (EEI) in ether-based electrolyte. PCS largely participates in the inner Na + sheath, causing more diglyme (G2) molecules and fewer PF 6 - anions to occupy the inner Na + solvation sheath, avoiding the decomposition of G2 molecules at high operation voltage. Moreover, PCS is beneficial for simultaneously constructing thin and robust sulfur-containing EEI on both Prussian blue (PB) cathode and HC anode, maintaining the structural stability of PB and alleviating the dissolution of transition metals. These enable PB||HC pouch cells to deliver a capacity retention of 60.3% after 400 cycles, significantly higher than the 32.1% in PCS-free electrolytes. In conclusion, this work discloses the underlying mechanism of PCS to evoke the anti-oxidation ability of ether-based electrolyte and provides a promising method for realizing advanced sodium-ion full cells.

anti-oxidation ability↗
DOE OSTI2017

Metal-containing ionic liquid-based, uncharged–charged diblock copolymers that form ordered, phase-separated microstructures and reversibly coordinate small protic molecules

A metal-containing ionic liquid-based, uncharged-charged block copolymer (BCP) system made via RAFT polymerization is presented. Small-angle X-ray scattering studies show these Co(II) bis(salicylate) anion-containing BCPs can form ordered microstructures (including a gyroid phase) in their neat states. Additionally, these BCPs can reversibly coordinate H 2 O or small alcohols with a noticeable color change.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗