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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Materials for defense/aerospace applications (NON-SV)

Through this effort, Sandia and Lockheed Martin Aeronautics Company (LM Aero) sought to assess the feasibility of (1) applying special materials to a defense application; (2) developing a piezoelectric-based micro thermophotovoltaic (TPV) cell; and (3) building and delivering a prototype laboratory emission measurement system. This project supported the Stockpile Research & Development Program by contributing to the development of radio frequency (RF) MEMS- and optical MEMS-based components - such as switches, phase shifters, oscillators, and filters - with improved performance and reduced weight and size. Investigation of failure mechanisms and solutions helped to ensure that MEMS-based technology will meet performance requirements and long term reliability goals in the specified environments dictated by Lockheed Martin's commercial and defense applications. The objectives of this project were to (1) fabricate and test materials for military applications; (2) perform a feasibility study of a piezoelectric-based micro TPV cell; and (3) build and deliver a prototype laboratory emission measurement system. Sandia fabricated and tested properties of materials, studied options for manufacturing scale-up, and delivered a prototype IR Emissometer. LM Aero provided material requirements and designs. Both participated in the investigation of attachment methods and environmental effects on material performance, a feasibility study of piezoelectric TPV cells, an investigation and development of new approaches to implement the required material functionality, and analysis and validation of material performance physics, numerical models, and experimental metrology.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Retention and surface morphology evaluation of fine-grain dispersion-strengthened tungsten for plasma-facing component applications

This study exposed novel fine-grain dispersion-strengthened tungsten (W) to high fluence, low energy deuterium (D) and helium (He) plasmas to evaluate how material microstructure and composition affect hydrogen retention and surface morphology. Tested materials included fine-grain dispersion-strengthened tungsten (DSW) with 3 wt% zirconium carbide (ZrC) dispersoids, fine-grain dense W without any dispersoids (FGW), and coarse-grained polycrystalline ‘ITER-grade’ W. Samples were exposed to D 2 + and He + plasmas at fusion-relevant fluences (∼10 25 m -2 ) and ion energies (75 eV) over a range of temperatures (200 °C, 300 °C, 450 °C for D, 850 °C for He). Helium ion microscopy was performed on the exposed samples to evaluate surface morphology changes and material integrity. After D plasma exposure, the ZrC dispersoids showed near-surface degradation at exposure temperatures above 300 °C, but no detrimental morphology changes were observed for the adjacent W grains. After He plasma-exposure, nano-structured fuzz formation was observed in the tungsten matrix of all samples. The ZrC dispersoids maintained their integrity despite the surrounding fuzz growth, with clear delineation between the W fuzz and dispersoid regions. Thermal desorption spectroscopy showed that ZrC DSW consistently retained more D than the FGW by about a factor of 2 across all temperatures. At 200 °C and 300 °C, the ITER-W displayed lower D retention than both the DSW and FGW, however at 450 °C ITER-W showed the highest retention, about 50% more than DSW. He retention was comparable across all samples, with the highest retention observed in the fine-grain W, only 26% higher than in ITER-W. These insights on retention behavior will inform further optimization of these novel fine-grained tungsten materials with and without dispersoid additives.

Dispersion-strengthened tungsten↗
DOE OSTI2019

Pulse Mode Calibration of the Micro Pocket Fission Detector

The Micro-Pocket Fission Detector (MPFD) is a neutron sensor currently under development in the High Temperature Test Laboratory (HTTL) for use as a supporting instrument during material testing experiments performed at TREAT. The detector has the capability to observe the real-time neutron population while surviving the harsh radiation, and temperature conditions, which sets it apart from most neutron sensitive devices. The next step in the instrument development process is to define a method of calibrating pulse-mode operations. This procedure will allow the detector to be calibrated to allow for absolute measurements of localized neutron population in real time.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2011

Advanced Test Reactor Tour

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

System Upgrades at the Advanced Test Reactor Help Ensure that Nuclear Energy Research Continues at the Idaho National Laboratory

Fully operational in 1967, the Advanced Test Reactor (ATR) is a first-of-its-kind materials test reactor. Located on the Idaho National Laboratory’s desert site, this reactor remains at the forefront of nuclear science, producing extremely high neutron irradiation in a relatively short time span. The Advanced Test Reactor is also the only U.S. reactor that can replicate multiple reactor environments concurrently. The Idaho National Laboratory and the Department of Energy recently invested over 13 million dollars to replace three of ATR’s instrumentation and control systems. The new systems offer the latest software and technology advancements, ensuring the availability of the reactor for future energy research. Engineers and project managers successfully completed the four year project in March while the ATR was in a scheduled maintenance outage. “These new systems represent state-of-the-art monitoring and annunciation capabilities,” said Don Feldman, ATR Station Manager. “They are comparable to systems currently used for advanced reactor designs planned for construction in the U.S. and in operation in some foreign countries.”

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Challenges and Solutions for Fast Neutron Irradiation of Bulk Material Specimens

Reactor developers continue to recognize opportunities for further enhancing fast spectrum reactor designs with advanced core materials, but all the material test reactors currently available to the United States are thermal spectrum designs. Fortunately, the Advanced Test Reactor and High Flux Isotope Reactor are versatile high flux facilities where spectral modification strategies can be used to reduce undesirable thermal neutron capture transmutation damage and augment fast flux delivered to specimens. New opportunities to leverage high flux regions and specially designed fast flux boosting experiment configurations can be used to achieve meaningful fast fluences on large specimens in ATR. New optimization potentials can be employed to achieve even higher fluences, albeit for smaller specimens, using thermal neutron filters in HFIR test positions. These capabilities, while not true fast reactors, can provide highly relevant environments for researchers needing to study the effects of fast neutron damage in bulk material specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Exxon Mobil-NETL Testing of DAC Materials

This joint initiative is aimed at extending our understanding of real DAC testing conditions at the NETL DAC testbed facility. Of particular focus will be: • Small Scale testing of powdered and formulated materials (supplied by ExxonMobil) to evaluate various performance metrics under DAC process cycles • Pilot scale testing of formulated materials at larger scales ExxonMobil will work with NETL to shake down equipment, validate testing methods, and define best practices for data analysis. Three tasks are proposed: 1. Validation of multi-cycle test data on powdered and formulated materials to benchmark various performance metrics of these materials. 2. Steam regeneration of powdered samples at small scale to define baseline performance under commercially relevant conditions. 3. Pilot scale testing of larger formulated materials to test commercially relevant samples under actual process cycles.

36 MATERIALS SCIENCE↗
DOE OSTI2015

International Round-Robin Study on Thermoelectric Transport Properties of n-type Half-Heusler from 300 K to 773 K

International transport property measurement round-robins have been conducted by the Thermoelectric Annex under the International Energy Agency (IEA) Implementing Agreement on Advanced Materials for Transportation (AMT). The previous round-robins used commercially available bismuth telluride as the testing material, with the goals of understanding measurement issues and developing standard testing procedures. The current round-robin extended the measurement temperature range to 773 K. It was designed to meet the increasing demands for reliable transport data of thermoelectric materials for power generation applications. Eleven laboratories from six IEA-AMT member countries participated in this study. Half-Heusler (n-type) material prepared by GMZ Energy was selected for the round-robin. The measured transport properties showed narrower distribution on uncertainties compared to previous round-robin efforts. The study intentionally included multiple testing methods and instrument types. Over the full temperature range, the measurement discrepancies on the figure of merit, ZT, in this round-robin were ±1.5 to ±16.4% from the averages.

36 MATERIALS SCIENCE↗
DOE OSTI2012

ATR LEU Monolithic Foil-Type Fuel with Integral Cladding Burnable Absorber – Neutronics Performance Evaluation

The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The burnable absorber - 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and to improve the peak ratio of the inner/outer heat flux. The present work investigates the LEU Monolithic foil-type fuel with 10B Integral Cladding Burnable Absorber (ICBA) design and evaluates the subsequent neutronics operating effects of this proposed fuel designs. The proposed LEU fuel specification in this work is directly related to both the RERTR LEU Development Program and the Advanced Test Reactor (ATR) LEU Conversion Project at Idaho National Laboratory (INL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Integrated Performance Testing Workshop - Supplemental Materials (Scripts and Procedures)

A variety of performance tests are described relating to: Material Transfers; Emergency Evacuation; Alarm Response Assessment; and an Enhanced Limited Scope Performance Test (ELSPT). Procedures are given for: nuclear material physical inventory and discrepancy; material transfers; and emergency evacuation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2011

The ORNL High Flux Isotope Reactor and New Advanced Fuel Testing Capabilities

The U.S. Department of Energy s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Relevant Additive Manufacturing Materials for Wind Industry

The purpose of this report is to equip wind industry professionals with the relevant material testing and properties of large format extrusion-based additive manufacturing processes. Discussion on the printability and processing of different classes of materials (i.e. low temperature, high temperature, thermoplastics, elastomeric, foam, etc.) is provided as well as relevant test standards and methods. Existing results for materials and properties are detailed. A companion document titled “Additive Manufacturing Design Guidelines for Wind Industry” covers the seven families of additive manufacturing, the relevance of each family to the wind industry, and design guidelines and considerations for the most relevant processes.

17 WIND ENERGY↗
DOE OSTI2015

Characterization of unsaturated hydraulic conductivity in fractured media using the multistep outflow method - 15461

The multistep outflow method is routinely used to characterize the unsaturated hydraulic conductivity of soils. The technique involves placing a soil sample in a pressure plate apparatus, subjecting the sample to multiple gas pressures in discrete steps through time, and measuring the transient volume of pore fluid extracted. Unsaturated hydraulic property values are estimated through inverse modeling of the experimental conditions. In this study the multistep outflow concept was applied to micro-cracked cementitious materials to assess the efficacy of the technique for these materials. The cementitious materials tested were salt-waste simulant grout samples artificially damaged through oven-drying. Compared to typical soils, fractured media exhibit higher saturated conductivity and lower air-entry pressure, and the volume of fluid extractable from fractures is much lower than soil porosity. To accommodate these material differences the standard test apparatus was modified to incorporate a higher conductivity ceramic pressure plate, a high-precision digital balance for logging outflow mass, a low volume (diameter) effluent line, multiple in-line high-precision gas regulators, and a high-precision low-range pressure gauge. Testing to date indicates that the modified apparatus can provide a viable means to measure the unsaturated hydraulic properties of micro-fractured cementitious materials. However the accuracy/uniqueness of inverse modeling is limited by the inherent characteristics of fractured media: high saturated conductivity, low air-entry pressure, and strong non-linearities. Hydraulic property results in the form of van Genuchten / Mualem curves are presented for three fractured grout specimens. DOE Performance Assessments often involve cementitious barriers and/or waste forms that are predicted or assumed to degrade over time due to various mechanisms such as carbonation-influenced reinforcing steel corrosion, external sulfate attack, differential settlement, and seismic activity. Physical degradation typically takes the form of small-scale cracking / fracturing, and the affected materials reside in unsaturated hydrogeologic zones. In these cases, unsaturated hydraulic properties are needed for fractured cementitious materials to simulate moisture movement and contaminant transport within and around the facility. The outflow extraction method, as implemented in the present study, provides a suitable method for estimating these material properties. (authors)

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

Performance Test of Mini LVDT - ELVIS

The Institute for Energy (IFE) Technology has been a pioneer in the development of Linear Variable Differential Transformers (LVDTs) for in-pile testing, deploying over 2,200 units in various reactor environments with less than a 10% failure rate after five years of operation. This report focuses on the performance testing of IFE’s Mini LVDT, a compact sensor ideal for material test reactor experiments. The Mini LVDT, with a limited range of +/- 1.5 mm, offers excellent performance comparable to larger LVDTs, making it valuable in space-constrained applications. The development of an Enhanced Linear Variable Intrinsic Sensor (ELVIS) with internal temperature monitoring capabilities represents a significant advancement, addressing the critical need for real-time, accurate measurements in high-radiation and high-temperature environments. Two ELVIS prototypes were evaluated in terms of both temperature and displacement, showcasing promising results, though challenges with noise during temperature measurements were identified. This report summarizes the rigorous testing performed at Idaho National Laboratory (INL) and highlights the potential applications of Mini LVDTs and ELVIS in nuclear and other high-precision industries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Stabilization of Residual Contamination with Alternative Materials - 19404

Since 2015, Savannah River National Laboratory (SRNL) in conjunction with Florida International University (FIU) have been investigating the applicability of intumescent coatings as radiological contamination fixatives for stabilization of residual materials after decontamination efforts have concluded. Intumescent coatings are designed to prevent heat transfer to a protected substrate and are typically used in both the aerospace industry to protect components from thermal shock and in building design to protect building joints that may fail in the event of a fire. For this work, these materials are examined for their ability to reduce worker risk by fixating residual contamination and lower the impact of a release in a scenario that breaches containment of a contaminated building and subsequently induces a large-scale fire by crediting their fire-retardant attributes. Initial studies of intumescent coatings and commercial fixatives included performance in direct flame, varied environmental conditions (20-110 F or -6.67 C, 10-95% relative humidity), adhesion studies to stainless steel after curing in varied environmental conditions, and radiological exposure. Of all the materials tested, many were found to be inadequate in one or more of these areas. These materials failed in one or more of three characteristic ways: 1) through delamination at reduced temperatures (≤40 F or 4.44 C) or from even moderate handling, 2) melting off the substrate at elevated temperatures (≥90 F or 32.22 C), 3) proving to be flammable and promote smoke propagation. Failure in these ways results in the potential for a subsequent release of the fixated contamination in the event of water presence/absence or direct flame. Using these results as a platform for the best in class identified material that did not experience failure in any of these tests, SRNL and FIU targeted the plutonium-238 contaminated hot cells within Savannah River Site (SRS) Building 235-F as a characteristic test bed to benchmark an intumescent coating used as a contaminant fixative. Through discussion with the SRS 235-F Risk Reduction personnel, appropriate hazards and performance metrics were discussed and a cold demonstration was performed in a hot cell mockup located at SRS. The application methods were proven functional in the cold mock up, and these intumescent coatings were subsequently utilized in the Savannah River Site Building 235-F's Plutonium Fuel Form (PuFF) Facility to aid in stabilization of plutonium-238 in both a contaminated hot cell and Entry Hood in September 2018. Presented in this work is the overview of the application process, associated working considerations for introduction of equipment and materials into the facility, lessons learned, and lifetime monitoring data to date. (authors)

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

Technical Manual for Automated J-R Curve Analysis Program Based on the ASTM E1820-18 Normalization Method

The J-integral vs. crack growth resistance curve (J-R curve) is a useful tool for evaluating materials fracture toughness in the ductile region. The widely applied standard for conducting such tests is American Society for Testing Materials (ASTM) Standard E1820-18 [1], in which the normalization analysis method is described in Annex 15. The normalization method requires only load-displacement data and crack size measurements before and after testing, which greatly simplifies the fracture toughness testing and alleviates many issues that arise when testing in extreme conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Accelerated materials evaluation for nuclear applications

This report addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62]

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗