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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
NASA NTRS1974

Composite materials. Volume 3 - Engineering applications of composites. Volume 4 - Metallic matrix composites. Volume 8 - Structural design and analysis, Part 2

This volume discusses a vaety of applications of both low- and high-cost composite materials in a number of selected engineering fields. The text stresses the use of fiber-reinforced composites, along with interesting material systems used in the electrical and nuclear industries. As to technology transfer, a similarity is noted between many of the reasons responsible for the utilization of composites and those problems requiring urgent solution, such as mechanized fabrication processes and design for production. Features topics include road transportation, rail transportation, civil aircraft, space vehicles, builing industry, chemical plants, and appliances and equipment. The laminate orientation code devised by Air Force materials laboratory is included. Individual items are announced in this issue.

Noton, B. R.↗
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↗
DOE OSTI2021

Heat-Iabile Foams 2021 (Project Report)

Foamable materials which can be readily decomposed to yield inert gas in a precise and moderate temperature range are desirable for applications in nuclear waste storage and securement during transportation. In the proposed work, promising lead candidates for such foams were identified from extant peer-reviewed literature, and initial samples synthesized and characterized.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Commercialization of the Transportation-Security, Tracking, and Reporting System (T-STAR)

The Transportation-Security, Tracking, and Reporting System (T-STAR) was developed by the National Nuclear Security Administration, NA-21, Office of Radiological Security (ORS) to provide a transportation security system for detection and tracking during transport of Category 1 and Category 2 radiological material. Few off-the-shelf systems for conveyance tracking offer detection of a cargo compartment breach or a removal of the cargo. Systems that do offer this capability often require permanent installation through modifying of the conveyance itself. This is not sustainable in many countries where ORS is building use, storage, and transport security capacity. The development of T-STAR has moved from fielding robust prototypes deployed in countries ranging from North America, Latin America and Central Asia to a commercially produced product that can now be deployed to provide enhanced security during transit. Each prototype deployment resulted in important lessons learned, which informed the requirements for the final commercial product. T-STAR uses both cellular and Iridium satellite modems to provide redundant communications to provide the configuration, status, and alerts to a server monitoring the shipment, which is accessible using a multilanguage browser-based user interface. A wireless security system employing using Z-wave sensors for intrusion detection located in the conveyance provide low cost but effective solution for a wide range of conveyance types. Additional capabilities include the ability to monitor a vehicles’ CANBUS (Controller Area Network) system, an ethernet port for high throughput sensor information such as video cameras, and the ability to power and use advanced external sensor payloads. These features make the T-STAR a capable and expandable security gateway that can be deployed on a variety of conveyances from box trucks to open trailers. The ability to provide tracking, monitoring, and detection provide a key component in overall best practices designed to protect shipments of radioactive material.

Schultze, Michael [ORNL] (ORCID:0000000283205671)↗
DOE OSTI2025

Domestic and International Consensus Standards for Nuclear Criticality Safety - Overview & Status

The domestic and international consensus standards for nuclear criticality safety (NCS) were developed based on the lessons-learned from process criticality accidents. These consensus standards were developed to reduce the rate of process criticality accidents in facilities that process, store, handle, or transport fissionable materials by hand. After a significant increase in criticality accidents through the mid-1960s, the rate of criticality accidents decreased as a result of these standards, and the criticality accident rate is extremely low from an industrial safety perspective. The last known criticality accident inside the United States was in 1978 (nearly 47 years ago) at the Idaho Chemical Processing Plant, and outside the United States, an accident occurred at Tokai-mura, Japan, in 1999 (almost 26 years ago). The domestic consensus standards for NCS include the American National Standards (ANS) that are prepared and published by the American Nuclear Society and approved by the American National Standards Institute (ANSI). The ANS Standards Board, the NCS Consensus Committee, and the ANS-8 Subcommittee oversee the development and maintenance of these standards. There are currently 18 standards in the ANS-8 series. Nine ANS-8 standards are either in revision mode or planned for revision. A new standard for the use of nondestructive assay measurements (ANSI/ANS-8.28-2024) for NCS was approved in March of 2024. The international consensus standards for NCS calculations, procedures, and practices are maintained and developed within the International Organization for Standardization, Technical Committee 85 “Nuclear Energy,” Subcommittee 5, “Nuclear Fuel Technology,” and Working Group 8, “Nuclear Criticality Safety.” Eleven standards are currently available, four standards are proposed for revision, and four standards are at various stages of development. This paper provides the NCS community with a high-level overview and status report of domestic and international NCS consensus standards to stimulate interest and to support their continued development.

Bowen, Douglas G [ORNL] (ORCID:0000000212460026)↗
DOE OSTI2025

A new capability facilitating nuclear materials research: the Activated Materials Laboratory at the Advanced Photon Source

The Activated Materials Laboratory (AML), located in the Long Beamline Building (LBB) of the Advanced Photon Source (APS) of Argonne National Laboratory (ANL), serves as a centralized radiological facility for preparing radioactive samples for APS beamline experiments. The AML is equipped to receive shipments, handle open-form radioactive materials, encapsulate samples, and transport samples to-and-from beamline end-stations. The AML works closely with users and the APS radiological safety committee to make sure the safe conduct of experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Thermal Neutron Scattering Law (TNSL) Implementation and Testing in FUDGE

Thermal neutron scattering law processing capabilities have been recently implemented in the LLNL code FUDGE (For Updating Data and Generating Evaluations). FUDGE is now capable of producing processed TNSL data for use in Monte Carlo or deterministic transport. To test this new capability, LLNL scientists performed an extensive intercomparison between the Mercury, Ardra, COG and MCNP transport codes. This intercomparison helped probe differences between how TNSL data are handled by the transport codes and by two different processing codes (LLNL’s FUDGE and LANL’s NJOY). This report summarizes recent improvements in FUDGE TNSL capabilities as well as results from the code intercomparison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

ANS Winter 2024 Summary: MCCAFE: The Monte Carlo Constructor for ATR Fuel Elements

The Irradiation Experiment Neutronics Analysis Department at Idaho National Laboratory (INL) has implemented a new analysis workflow for experiments in the Advanced Test Reactor (ATR). One key piece of this workflow is the Monte Carlo Constructor for ATR Fuel Elements, or MCCAFE. For each ATR operating cycle, the Reactor and Nuclear Safety Engineering (RNSE) Department first solves the core in eigenvalue mode and depletes the driver fuel materials. In a separate calculation, neutronics analysts model and deplete the materials of one or more irradiation experiments, usually in a series of fixed-source Monte Carlo N-Particle (MCNP) models of the ATR for neutron transport calculations. It was desirable to use the results of the former calculations to inform the models of the latter. MCCAFE is a Python program developed using American Society of Mechanical Engineers Nuclear Quality Assurance-1 procedures at INL. Its purpose is to take the calculated results from the RNSE depletion solutions and the measured or projected operating parameters from the Nuclear Data Management and Analysis System (NDMAS) to generate fixed-source models of the ATR core at given points in time across one or more cycles.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

FENIX: An Open-Source Multiphysics Integrated Framework Enabling Collaborative Development of Plasma Facing Component Modeling Capabilities

Advanced modeling and simulation tools have a crucial role to play in accelerating fusion energy deployment as a sustainable power source. Multiphysics, high-fidelity computational tools can help understand, model, and quantify the complex interactions between materials performance, plasma and neutron exposure, and engineering processes. As a result, they accelerate the design, safety analysis, and performance evaluation of fusion systems. This webinar introduces the Fusion ENergy Integrated multiphys-X (FENIX) framework, an open-source multiphysics tool for plasma facing component modeling. FENIX leverages the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, which has been developed by the United States Department of Energy Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. FENIX couples various MOOSE capabilities such as heat transfer, thermomechanics, thermal hydraulics, electromagnetics, and plasma kinetics with the MOOSE-based applications Cardinal (neutronics) and TMAP8 (tritium transport). During the webinar, we will present FENIX and discuss how its modularity, openness, software quality assurance processes, and licensing approach supports effective collaborations, including public-private partnerships.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Two-Dimensional Uranyl Borates: From Conventional to Extreme Synthetic Conditions: Two-Dimensional Uranyl Borates: From Conventional to Extreme Synthetic Conditions

A systematic investigation of uranyl borates under different synthetic conditions resulted in five new 2D compounds, namely, (H 3 O)[(UO 2 )(BO 3 )], Li[(UO 2 )(BO 3 )] · (H 2 O), α-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ], β-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ] and K 2.5 [(UO 2 ) 5 (BO 3 ) 2 O 2.5 (OH) 1.5 ] · (H 2 O) 2.5 . (H 3 O)[(UO 2 )(BO 3 )] and Li[(UO 2 )(BO 3 )] · (H 2 O) were obtained from hydrothermal reactions at 220 °C using the same mineralizer. Both materials possess the uranophane sheet topology with different symmetry of the unit cells. In the structure of (H 3 O)[(UO 2 )(BO 3 )] and Li[(UO 2 )(BO 3 )] · (H 2 O), UO 7 pentagonal bipyramids share edges and vertexes with four BO 3 planar triangles. α-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ] and β-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ] are polytypes. α-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ] was synthesized from a high-temperature solid-state reaction under ambient pressure; whereas β-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ] was obtained from a high-temperature/high-pressure (HT/HP) reaction. Both structures have an identical anion topology, but β-K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ] crystallizes in space group with higher symmetry. In K 2.5 [(UO 2 ) 5 (BO 3 ) 2 O 2.5 (OH) 1.5 ] · (H 2 O) 2.5 , which was obtained from a hydrothermal reaction, UO 7 polyhedra share vertexes and edges with two independent BO 3 triangles, forming the most complex uranyl borate layers among all five compounds. The different synthetic routes, novel topologies, thermal behavior and Raman spectra are discussed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗