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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
NASA NTRS1976

Solar heating and cooling technical data and systems analysis

The accomplishments of a project to study solar heating and air conditioning are outlined. Presentation materials (data packages, slides, charts, and visual aids) were developed. Bibliographies and source materials on materials and coatings, solar water heaters, systems analysis computer models, solar collectors and solar projects were developed. Detailed MIRADS computer formats for primary data parameters were developed and updated. The following data were included: climatic, architectural, topography, heating and cooling equipment, thermal loads, and economics. Data sources in each of these areas were identified as well as solar radiation data stations and instruments.

Christensen, D. L.↗
DOE OSTI2019

Materials for learning use of GSAS-II

The General Structure Analysis System (GSAS-II) package provides materials and crystallographic analysis for all types of diffraction data. It was initially made available with very limited capabilities, but over much of the last decade the features have been expanded, so that GSAS-II is now a comprehensive tool for nearly all types of structural and materials characterization studies. The need to provide materials to teach use of GSAS-II, while the software has been undergoing constant revision and expansion, has required new approaches for documentation. This has included providing tutorials, as each major new capability has been added, and context-sensitive help for each section of the program. Comments in the code are also expanded into a software reference guide. Finally, most recently, video versions of more than half of the tutorials were created and others were provided with animated graphics. All GSAS-II documentation is web-based.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Data Processing Package for Cyclic Integrated Reversible Bending Fatigue Testing

A data processing software package has been introduced. The package was developed using MATLAB with the aid of the Curve Fitting Toolbox. The package is made up of four modules: pre-processing, data processing for static testing, data processing for monitoring, and data processing for measurements. CIRFT data are structured with multiple levels of architecture involving group, specimen, session, and scan/block. The degree of complexity of the data structure depends on whether a test is static or cyclic.The test results are presented in figures, scatter plots, and tables. For static testing, the output in tables provides bending mechanical properties and characteristic points of moment–curvature relation: flexural rigidities in linear segments of loading and unloading stages, intersection points between characteristic segments of the curve, and equivalent stress and strain quantities. For cyclic testing, the table output lists control and fatigue life and responsive/dependent quantities including moment, curvature, flexural rigidity, flexural hysteresis, rigidity phase angle, and equivalent stresses and strains. In addition, derivatives such as half-gage length and sensor spacing correction are included. The output also provides standard deviations of the reported quantities when they are applicable or available. The data processing package can serve as a fundamental characterization tool in mechanical study of materials. The package is intended primarily for data processing for the CIRFT process and can also be used in applications for which similar testing requirements exist.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Material Needs and Measurement Challenges for Advanced Semiconductor Packaging: Understanding the Soft Side of Science

This Perspective builds upon insights from the National Institute of Standards and Technology (NIST)-organized workshop, “Materials and Metrology Needs for Advanced Semiconductor Packaging Strategies,” held at the 35th annual Electronics Packaging Symposium in Binghamton, NY, on September 5, 2024. It outlines critical challenges and opportunities related to polymer-based “soft” materials in advanced semiconductor packaging, with emphasis on polymer science, measurement science (metrology), and the strategic development of Research-Grade Test Materials (RGTMs). These efforts, led by the NIST CHIPS team, aim to advance the fundamental understanding of structure-property-processing relationships, promote standardized guidelines and innovative methods for material characterization, and accelerate the development, qualification, and adoption of next-generation packaging materials. The Perspective also distills key insights from the panel discussion with industry experts, emphasizing the need for close collaboration among materials scientists, process engineers, and metrology experts to enable a holistic strategy, further highlighting the importance of cross-sector partnerships among industry, academia, and government to address pressing challenges in packaging materials and processes.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

A Methodology for defining intermediate level waste package characteristics that ensure criticality safety in a geological disposal facility - 14599

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. The basis for ensuring criticality safety of intermediate level waste (ILW) packages is to specify controls on the fissile material content of individual packages through the application of deterministic criticality safety assessments (CSAs). The approach is defined and controlled through RWM's Disposability Assessment process, which aims to ensure that: criticality is prevented during waste package transport to a GDF; the risk of criticality is tolerable and as low as reasonably practicable (ALARP) during GDF operations; and the likelihood and consequences of criticality arising from the evolution of waste packages after GDF closure are low. The most restrictive phase determines the package fissile material limit and, until now, in many cases the GDF post-closure phase has provided the most restrictive limit. Currently, the GDF post-closure limits are derived deterministically using highly stylised and conservative assumptions about fissile material accumulations occurring on very long timescales into the future. RWM, waste producers and regulators do not currently consider this approach to be yielding an appropriate balance of risk. Therefore, RWM is modifying the process for setting fissile material limits on ILW packages by making use of the outcomes of its recent research projects to establish the likelihood of post-closure criticality in a GDF and the consequences of hypothetical post-closure criticality events (presented in companion papers). This paper presents the revised methodology, which is based on the requirement that criticality in a GDF can be demonstrated to be of low and tolerable consequences with regard the performance of a GDF. A probabilistic 'low-likelihood package envelope' is proposed that establishes the packaging and disposal facility conditions under which post-closure criticality is considered to be of low likelihood and low consequence. The bounding conditions for establishing the extent of the package envelope have been evaluated through application of the methodologies and models developed in RWM's Likelihood of Criticality research project. It is expected that the vast majority of ILW packages would meet the package envelope requirements. The methodology also allows the use of other options, such as deterministic CSAs, for waste packages that fall outside the low-likelihood package envelope. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRSDate not supplied

MLtool++ package for machine learning and its applications to materials data

We are developing Mltool++ package of software programs for machine learning (ML). Given the MLtool Python code, we create a faster C++ code with the potential for parallelization. We have extracted materials data from the literature. One dataset contains melting temperatures of stoichiometric 1:1 metallic compounds XZ, composed by elements X={Al, Ti, V, Cr, Zr, Nb, Mo, Hf, Ta, W} and Z={Co, Ni, Cu, Rh, Pd, Ag, Ir, Pt, Au}, and another contains solid-solid symmetry-breaking phase transition temperatures. We studied dependences of temperatures on composition, found several correlations, and parametrized them by analytical functions. Mltool++ package is generic and applicable to any tabulated numeric data.

Pierce M. Pettit↗
NASA NTRS2011

Thermally Conductive Tape Based on Carbon Nanotube Arrays

To increase contact conductance between two mating surfaces, a conductive tape has been developed by growing dense arrays of carbon nanotubes (CNTs, graphite layers folded into cylinders) on both sides of a thermally conductive metallic foil. When the two mating surfaces are brought into contact with the conductive tape in between, the CNT arrays will adhere to the mating surface. The van der Waals force between the contacting tubes and the mating surface provides adhesion between the two mating surfaces. Even though the thermal contact conductance of a single tube-to-tube contact is small, the tremendous amount of CNTs on the surface leads to a very large overall contact conductance. Interface contact thermal resistance rises from the microroughness and the macroscopic non-planar quality of mating surfaces. When two surfaces come into contact with each other, the actual contact area may be much less than the total area of the surfaces. The real area of contact depends on the load, the surface roughness, and the elastic and inelastic properties of the surface. This issue is even more important at cryogenic temperatures, where materials become hard and brittle and vacuum is used, which prevents any gas conduction through the interstitial region. A typical approach to increase thermal contact conductance is to use thermally conducting epoxies or greases, which are not always compatible with vacuum conditions. In addition, the thermal conductivities of these compounds are often relatively low. The CNTs used in this approach can be metallic or semiconducting, depending on the folding angle and diameter. The electrical resistivity of multiwalled carbon nanotubes (MWCNTs) has been reported. MWCNTs can pass a current density and remain stable at high temperatures in air. The thermal conductivity of a MWCNT at room temperature is measured to be approximately 3,000 W/m-K, which is much larger than that of diamond. At room temperature, the thermal conductance of a 0.3 sq cm array of CNTs was measured to be as high as 10 W/K. The high thermal conductivity and the nanoscale size make CNTs ideal as thermal interface materials. The CNT-based thermal tape can be used for the thermal management of microelectronic packages and electronic systems. It also can be integrated with current device technology and packaging. The material would allow for an efficient method to manage excess heat generation without requiring any additional power. Lastly, the CNT tape can be used to enhance thermal contact conductance across two mating surfaces on some NASA missions.

Kashani, Ali↗
NASA NTRS1985

A study of the compatibility of an existing CFD package with a broader class of material constitutions

The flexibility of the PHOENICS computational fluid dynamics package was assessed along two general avenues; parallel modeling and analog modeling. In parallel modeling the dependent and independent variables retain their identity within some scaling factors, even though the boundary conditions and especially the constitutive relations do not correspond to any realistic fluid dynamic situation. PHOENICS was used to generate a CFD model that should exhibit the physical anomalies of a granular medium and permit reasonable similarity with boundary conditions typical to membrane or porous piston loading. A considerable portion of the study was spent prying into the existing code with a prejudice toward rate type and disarming any inherent fluid behavior. The final stages of the study were directed at the more specific problem of multiaxis loading of cylindrical geometry with a concern for the appearance of bulging, cross slab shear failure modes.

French, K. W., Jr.↗
DOE OSTI2012

Code Qualification of Structural Materials for AFCI Advanced Recycling Reactors

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Electric-Drive Vehicle Power Electronics Thermal Management: Current Status, Challenges, and Future Directions

Effective thermal management of traction-drive power electronics is critical to the advancement of electric-drive vehicles and is necessary for increasing power density and improving reliability. Replacing traditional silicon devices with more efficient, higher temperature, higher voltage, and higher frequency wide-bandgap (WBG) devices will enable increased power density but will result in higher device heat fluxes. Compact packaging of high-temperature WBG devices near low-temperature-rated components creates thermal management challenges that need to be addressed for future power-dense systems. This paper summarizes the thermal performance of on-road automotive power electronics thermal management systems and provides thermal performance and pumping-power metrics for select vehicles. Thermal analyses reveal that the package/conduction resistance dominates the total thermal resistance (for existing automotive systems). We model advanced packaging concepts and compare the results with existing packaging designs to quantify their thermal performance enhancements. Double-side-cooled configurations that do not use thermal interface materials are package concepts predicted to provide a low junction-to-fluid thermal resistance (compared to current packages). Dielectric-fluid-cooled concepts enable a redesign of the package to reduce the package resistance, can be implemented in single- and two-phase cooling approaches, and allow for cooling of passive components (e.g., capacitors) and bus bars.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2011

Development of an Air Transport Type A Fissile Package

This paper presents the summary of testing by the Savannah River National Laboratory (SRNL) to support development of a light weight (<140 lbs) air transport qualified Type A Fissile Packaging. The package design incorporates features and materials specifically designed to minimize packaging weight. The light weight package is being designed to provide confinement to the contents when subjected to the normal and hypothetical conditions required of an air transportable Type A Fissile radioactive material shipping package. The objective of these tests was to provide design input to the final design for the LORX Type A Fissile Air Transport Packaging when subjected to the performance requirements of the drop, crush and puncture probe test of 10CFR71. The post test evaluation of the prototype packages indicates that all of the tested designs would satisfactorily confine the content within the packaging. The differences in the performance of the prototypes varied significantly depending on the core materials and their relative densities. Information gathered from these tests is being used to develop the final design for the Department of Homeland Security.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2010

SECOND STATUS REPORT: TESTING OF AGED SOFTWOOD FIBERBOARD MATERIAL FOR THE 9975 SHIPPING PACKAGE

Samples have been prepared from a softwood fiberboard lower subassembly. Physical, mechanical and thermal properties have been measured following varying periods of conditioning in each of several environments. These tests have been conducted in the same manner as previous testing on cane fiberboard samples. Overall, similar aging trends are observed for softwood and cane fiberboard samples, with a few differences. On the positive side, the softwood fiberboard data to date shows less sample-to-sample variation in physical properties than cane fiberboard, and the thermal conductivity decreases at a slower rate at 250F for softwood fiberboard than for cane fiberboard. On the other hand, the softwood fiberboard physical property samples generally show degradation rates greater than cane fiberboard samples in the 185F 30%RH environment. Testing following additional conditioning will continue and the addition of samples in other elevated humidity environment(s) will be pursued to identify the extent of these trends. Post-conditioning data have been measured on samples from a single softwood fiberboard assembly, and baseline data are also available from a limited number of vendor-provided samples. This provides minimal information on the possible sample-to-sample variation exhibited by softwood fiberboard. Data to date are generally consistent with the range seen in cane fiberboard, but some portions of the data trends are skewed toward the lower end of that range. Further understanding of the variability of softwood fiberboard properties will require testing of additional material.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Test Report for DPP-1 Regulatory Compliance Testing. Volume 1 - Main Report

Consolidated Nuclear Security, LLC (CNS), the management and operating contractor for the Y-12 National Security Complex (Y-12) for the National Nuclear Security Administration (NNSA), is currently responsible for the regulatory testing of a new Type B fissile material shipping package called the Defense Programs Package (DPP)-1. The DPP-1 is a drum packaging with an inner liner and a removable lid. Both the drum body and the lid are filled with an impact-limiting, thermal-insulating material that protects the inner containment vessel (CV) during normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as defined by 10 US code of Federal Regulations (CFR) 71. In fiscal year 2021, the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP), located at the National Transportation Research Center (NTRC), conducted Type B performance tests as prescribed in Title 10 Code of Federal Regulations Part 71 (10 CFR 71) on seven DPP-1 prototype units. This test report documents test unit (TU) preparation, pretest condition, conditioning of each TU, NCT and HAC testing, and posttest measurements and observations made of the damage resulting from these tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Resolution of Qualification Issues for Existing Structural Materials

This report gives a detailed assessment of several key technical issues that needs resolution for the existing structural materials with emphasis on application in liquid metal reactors (LMRs), in particular, sodium cooled fast reactors. The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for the Advanced Fuel Cycle Initiative (AFCI) Reactor Campaign. The report is the second deliverable in FY09 (M2505050201) under the work package “Advanced Materials Code Qualification”. The overall objective of the Advanced Materials Code Qualification project is to evaluate the key technical requirements for the qualification of currently available and future advanced materials for application in sodium reactor systems and the resolution of issues that the U.S. Nuclear Regulatory Commission (NRC) has raised in the past on structural materials in support of the design and licensing of the LMR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. However, the development of sufficient database and qualification of these materials for application in LMRs require considerable amount of time and resources. In the meantime, the currently available materials will be used in the early development of fast reactors. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants licensing. As the LMR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Section III Subsection NH (Class 1 Components in Elevated Temperature Service). Assessment of materials performance issues and high temperature design methodology issues pertinent to the LMR were presented in an earlier report (Natesan et al. 2008). In a subsequent report (Majumdar et al. 2009), we addressed the needs in high temperature methodologies for design of various high temperature components in sodium cooled fast reactor. The present report addresses several key technical issues for the currently available structural materials such as Type 304 and 316 austenitic stainless steels and ferritic steels such as 2.25Cr-1Mo and modified 9Cr-1Mo. The 60-year design life for the LMR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high temperature structural design methodology. The current Subsection NH is applicable to the design life only up to 34 years. No experimental data contain test durations of 525,000 hours, and it is impractical to conduct such long-term tests in any types of testing. So far the longest creep tests for Grade 91 and Grade 92 steels have run up to 100,000 hours. It has been noted that there is a large drop in creep rupture strength in long-term tests for these high-Cr creep-resistant steels, which may result in overestimation of the long-term creep strength and allowable stress. The report addresses in detail the need for a mechanistic understanding of the structural materials, from the standpoint of the effects of thermal aging, creep deformation, creep fracture, fatigue and creep-fatigue, creep-fatigue predictive models, fatigue and creep crack growth, and fracture toughness. Based on an in-depth assessment of the available data and mechanistic understanding, key technical issues are identified and discussed for each of the property areas. Furthermore, we have proposed viable approaches to resolve the issues and prioritized our recommendations.

Natesan, K.↗
DOE OSTI2021

Cold Testing for Characterization of Oxide Feed Impurities in Support of the ARIES Program

The Advanced Recovery and Integrated Extraction System (ARIES) Program converts surplus plutonium metal components into an oxide powder form and packages the material for final disposition. In support of the ARIES program, over a metric ton (MT) of plutonium has been removed from surplus nuclear weapons since 1998 and packaged for long-term storage in 3013 cans. The oxide production rate is anticipated to increase substantially over the next decade in support of dilute and dispose goals. Analytical chemistry data from destructive analysis of the resulting 70+ blend lots to date has been reported for contaminants in the product oxide. Changes in the Pu product specification for Pu oxide product, induced by a different pit mix, the LANL swap program and processing Alternate (non-Pit Pu) Feed Stock (AFS) metals, will affect the level and the mix of these contaminants in the upcoming years. With these changes in feed material, it will be important to provide fast and efficient measurement of elemental impurities in the oxide to be certain that the process is under control and to monitor the anticipated changes from the historical blend lot data. Non-destructive methods need to be evaluated for the characterization of the plutonium oxide product and the quantification of the impurities at the current and anticipated levels. A feasibility study has begun to investigate the applicability of new spectroscopic methods for oxide impurity characterization. To this end, here we describe the preparation of surrogate materials to support these investigations, and the results of a feasibility study investigating the use of LIBS (Laser Induced Breakdown Spectroscopy) to characterize the surrogate materials as calibration standards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Load planning in the dark: RH-TRU waste container load planning at Sandia National Laboratories/New Mexico - 15116

Sandia National Laboratories (Sandia) performed several sets of experiments during the 1970's and 1980's with fresh and irradiated mixed oxide (MOX) and uranium oxide (UO{sub 2}) materials designed to simulate various fuel responses under varying reactor conditions. These research materials were from light water and breeder mixed oxide reactors. The experimental vessels and extraneous piece parts resulting from the experiments were identified as eligible for disposal at the Waste Isolation Pilot Plant (WIPP) as remote-handled (RH) transuranic (TRU) waste. Once Sandia received the approval to dispose at WIPP, discussions began with the Carlsbad Field Office (CBFO) and Nuclear Waste Partnerships, LLC (NWP) to provide Central Characterization Program (CCP) support for certified acceptable knowledge, visual examination, and dose-to-curie for the repackaging effort. In addition to the WIPP support, Sandia was required to prepare a load plan describing the configuration of the parent containers and how they could be repackaged to meet WIPP and other regulatory requirements. However, all of these containers were packaged as material twenty (20) or thirty (30) years ago and the records of the packaging process and configuration were not documented as waste. Therefore, it seemed that the load plans were prepared with limited information regarding parent container configuration, hence in the dark. The design of the Auxiliary Hot Cell Facility (AHCF) in technical area (TA) V, where the repackaging was to take place, along with the shipping requirements for the RH 72-B cask, limited the options for repackaging. This configuration required that 0.114 m{sup 3} (30-gallon) containers be packaged in the hot cell, removed and placed into 0.208 m{sup 3} (55-gallon) containers to mitigate contamination issues. The material had been tracked as accountable nuclear material and Sandia had verified Material Control and Accountability (MC and A) data on each parent container. Based on the MC and A data, scaling factors could be determined and sampling and analysis was not required. However, it meant that parent containers could not be split into two daughter containers, though two or more parent containers could be combined into one daughter. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗