Status of solar simulation at Lewis
Evaluation of solar simulators using enclosed sources, in-house measurement facility for testing proposed devices to produce solar simulation, and related research programs
SOLAR SIMULATOR↗
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Results for “Material Testing”
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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Experimental investigation of a charring ablative material exposed to combined convective and radiative heating.
Charring ablative materials effectiveness in combined convective and radiative heating as reentry shield
MATERIAL TESTING↗
Study of asbestos for electrochemical cells First summary report
Fuel cell asbestos test program
ASBESTOS↗
A Module Experimental Process System Development Unit (MEPSDU)
A cost effective process sequence and machinery for the production of flat plate photovoltaic modules are described. Cells were fabricated using the process sequence which was optimized, as was a lamination procedure. Insulator tapes and edge seal material were identified and tested. Encapsulation materials were evaluated.
Source record↗
The Pack Method for Compressive Tests of Thin Specimens of Materials Used in Thin-Wall Structures
The strength of modern lightweight thin-wall structures is generally limited by the strength of the compression members. An adequate design of these members requires a knowledge of the compressive stress-strain graph of the thin-wall material. The "pack" method was developed at the National Bureau of Standards with the support of the National Advisory Committee for Aeronautics to make possible a determination of compressive stress-strain graphs for such material. In the pack test an odd number of specimens are assembled into a relatively stable pack, like a "pack of cards." Additional lateral stability is obtained from lateral supports between the external sheet faces of the pack and outside reactions. The tests seems adequate for many problems in structural research.
Aitchison, C S↗
Further Studies of Materials Compatibility in High-Test Hydrogen Peroxide
Assessment of the compatibility of high-test hydrogen peroxide (HTP) with materials, in particular new materials such as composites, is critical to the development of new propulsion systems meeting requirements of reduced cost and environmental impact. While compatibility with HTP has been addressed previously, newer materials were not considered. The focus of this project was to develop a scheme for evaluation of HTP with all materials. In the previous summer, methods were developed for production of HTP on site, and preliminary steps were taken to evaluate materials. Methods investigated this summer have included accelerated aging by heating, coupled with assay of concentration and stabilization loss, observation of reactivity by means of Isothermal Microcalorimetry, and evaluation of changes to the materials by Short Beam Shear testing and by Photoacoustic-Fourier Transform Infrared Spectroscopy. Various metals, polymers, and composites were examined in this study.
Gostowski, R.↗
The Columbia Accident Investigation and The NASA Glenn Ballistic Impact Laboratory Contributions Supporting NASA's Return to Flight
On February 1, 2003, the Space Shuttle Columbia broke apart during reentry, resulting in loss of the vehicle and its seven crewmembers. For the next several months, an extensive investigation of the accident ensued involving a nationwide team of experts from NASA, industry, and academia, spanning dozens of technical disciplines. The Columbia Accident Investigation Board (CAIB), a group of experts assembled to conduct an investigation independent of NASA, concluded in August, 2003 that the most likely cause of the loss of Columbia and its crew was a breach in the left wing leading edge Reinforced Carbon-Carbon (RCC) thermal protection system initiated by the impact of thermal insulating foam that had separated from the orbiters external fuel tank 81 seconds into the mission's launch. During reentry, this breach allowed superheated air to penetrate behind the leading edge and erode the aluminum structure of left wing, which ultimately led to the breakup of the orbiter. The findings of the CAIB were supported by ballistic impact tests, which simulated the physics of External Tank Foam impact on the RCC wing leading edge material. These tests ranged from fundamental material characterization tests to full-scale Orbiter Wing Leading Edge tests. Following the accident investigation, NASA spent the next 18 months focused on returning the shuttle safely to flight. In order to fully evaluate all potential impact threats from the many debris sources on the Space Shuttle during ascent, NASA instituted a significant impact testing program. The results from these tests led to the validation of high-fidelity computer models, capable of predicting actual or potential Shuttle impact events, were used in the certification of STS-114, NASA s Return to Flight Mission, as safe to fly. This presentation will provide a look into the inner workings of the Space Shuttle and a behind the scenes perspective on the impact analysis and testing done for the Columbia Accident Investigation and NASA's Return to Flight programs. In addition, highlights from recent Shuttle missions are presented.
Melis, Matthew E.↗
Application of Direct Current Potential Drop for the J-integral vs. Crack Growth Resistance Curve Characterization
The direct current potential drop (DCPD) technique has been applied to derive the J-integral vs. crack growth resistance curve (J-R curve) for fracture toughness characterization of structural materials. The test matrix covered three materials including type 316LN stainless steels, Ni-based alloy 617, and one ferritic-martensitic steel, three specimen configurations including standard compact, single edge bending, and disk-shaped compact specimens, and temperatures ranging from 20 C to 650 C. When compared with baseline J-R curves derived from the ASTM normalization method, the original J-R curves from the DCPD technique yielded much smaller Jq values due to the influence of crack blunting, plastic deformation, etc. on potential drop. To counter these effects, a new procedure for adjusting DCPD J-R curves was proposed. After applying the new adjustment procedure, the average difference in Jq between the DCPD technique and the normalization method was only 5.2% and the difference in tearing modulus was 7.4%. The promising result demonstrates the applicability of the DCPD technique for the J-R curve characterization especially in extreme environments, such as elevated temperatures, where the conventional elastic unloading compliance method faces considerable challenges.
DCPD↗
Preserving physics knowledge at the fast flux test facility
One of the goals of the Dept. of Energy's Office of Nuclear Energy, initiated under the Fuel Cycle Research and Development Program (FCRD) and continued under the Advanced Reactor Concepts Program (ARC) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMRs) that could support the development of an environmentally and economically sound nuclear fuel cycle. The Fast Flux Test Facility (FFTF) is the most recent LMR to operate in the United States, from 1982 to 1992, and was designed as a fully instrumented test reactor with on-line, real time test control and performance monitoring of components and tests installed in the reactor. The 10 years of operation of the FFTF provided a very useful framework for testing the advances in LMR safety technology based on passive safety features that may be of increased importance to new designs after the events at Fukushima. Knowledge preservation at the FFTF is focused on the areas of design, construction, and startup of the reactor, as well as on preserving information obtained from 10 years of successful operating history and extensive irradiation testing of fuels and materials. In order to ensure protection of information at risk, the program to date has sequestered reports, files, tapes, and drawings to allow for secure retrieval. A disciplined and orderly approach has been developed to respond to client's requests for documents and data in order to minimize the search effort and ensure that future requests for this information can be readily accommodated. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Scaling for Nuclear Reactor System: Overview
A scaling analysis is essential for predicting the behavior of a reactor system under various operating conditions, as well as for ensuring safe and reliable experimentation and adequate dataset generation for supporting the system code model development and assessment in accordance with USNRC guidelines 1.203. Testing and modeling can be carried out at various scales, from small test units to full-scale reactors, to validate the predictions produced by the analysis. However, due to the cost and challenges associated with full-scale testing, the nuclear industry uses scaled test facilities—integral effect test (IET) and separate effect test (SET)—to identify potential design issues and optimize the design to ensure regulatory compliance for safer operation and anticipated accident management. In general, IET facilities are non-nuclear thermal hydraulics test facilities. In contrast, SET facilities can be non-nuclear testing (e.g., steam generator, containment, pump system SETs) and nuclear testing (mostly fuel-material irradiation testing). IETs are used to study the system-level thermal-hydraulic behavior of the reactor coolant system (RCS) under different anticipated operational occurrences (AOOs) and design basis accidents (DBAs) such as loss-of-coolant-accident (LOCA) and loss-of-power (i.e., station blackout) and loss-of-flow (i.e., failure of RCS pump). Several LOCA scenarios could happen in reactor DBAs, such as main steam line break (MSLB), SG tube rapture (SGTR), break in direct vessel injection (DVI), and other small breaks LOCA. Scaling analysis determines the appropriate size and operating conditions for the test facility. This involves analyzing the system's thermal-hydraulic behavior at different scales and identifying the key parameters that affect the thermal-hydraulic behavior of the reactor. These key parameters can be identified so that the scaled facility can replicate the expected conditions in a prototypical setting. The scaled facility can simulate the same flow rates, pressure gradients, heat transfer coefficients, and fluid properties as the prototype. Scaling analysis is used to ensure that the results from these tests can be scaled up to predict the behavior of a full-scale prototype with an acceptable range of scaling distortions.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Relative toxicity of pyrolysis gases from materials - Effects of chemical composition and test conditions
Relative toxicity test data on 270 materials are presented, based on test procedures developed at the University of San Francisco. The effects of chemical composition, using data on 13 types of synthetic polymers and eight types of fabrics, are discussed. Selected materials were evaluated using nine test conditions with the USF method, and using methods developed at the FAA Civil Aeromedical Institute, Douglas Aircraft Company and San Jose State University.
Hilado, C. J.↗
Reusable Surface Insulation (RSI) Material Samples
Test specimen data sheets are presented for 48 high temperature and 40 low temperature reusable surface insulation tiles. Tabular data show dimensions, density, coating weight, and final tile weight. Codes indicate basic material, surface coating, and hydrophobic treatment.
Banas, R. P.↗
ANALYTICAL NEUTRONIC STUDIES CORRELATING FAST NEUTRON FLUENCE TO MATERIAL DAMAGE IN CARBON, SILICON, AND SILICON CARBIDE
This study evaluates how fast neutron fluence >0.1 MeV correlates to material damage (i.e., the total fluence spectrum folded with the respective material’s displacements-per- atom [dpa] damage response function) for the specific material fluence spectra encountered in Next Generation Nuclear Plant (NGNP) service and the irradiation tests conducted in material test reactors (MTRs) for the fuel materials addressed in the white paper. It also reports how the evaluated correlations of >0.1 MeV fluence to material damage vary between the different spectral conditions encountered in material service versus testing.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Promoting the regulatory acceptance of combined ion and neutron irradiation for material degradation in nuclear reactors
The Advanced Materials and Manufacturing Technologies (AMMT) program within the Department of Energy (DOE) Office of Nuclear Energy has developed its current recommendation for promoting the use of combined ion irradiation and neutron irradiation for the accelerated qualification of materials to be deployed in nuclear reactors. This plan is intended to provide a collaborative path forward that can be adopted by academia, national laboratories, and industry, and has been developed with input from the regulatory research arm of the U.S. Nuclear Regulatory Commission (NRC). To deploy new materials or materials manufactured with new technologies, such as additive manufacturing, materials must be evaluated for reactor-induced degradation from the combination of harsh temperatures, corrosive environments, and radiation fields. However, rapid deployment of materials necessitates accelerated testing methods rather than relying on years of neutron irradiation in a material test reactor. Ion irradiation has demonstrated success in reproducing material microstructure and select property evolution resulting from neutron irradiation with three to four orders of magnitude reduction in time and cost, making it an ideal candidate for accelerated irradiation testing. This presentation provides context governing both the scientific and regulatory aspects of the proposed goal. The discussion is aimed at a broad audience including researchers from industry, national laboratories, and academia. The recommended path forward is presented as a conceptual framework of specific steps. In brief, the strategy entails developing an integrated ion and neutron irradiation test plan for the material property of interest based on the fundamental tenet of the linkage of microstructure and properties in materials. Physics-based modeling interprets ion irradiation data and predicts neutron irradiation microstructure and properties with uncertainty bounds. The first round of testing is sufficient for an initial licensing application using a risk-informed approach, while a minimum required neutron irradiation test plan reduces cost and time requirements. A surveillance program with witness specimens in-reactor provides additional data over time to improve model predictions to higher damage levels and further reduce uncertainty bounds, which can be used for license extensions or longer lifetimes in new license applications.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Performance evaluation of DAAF as a booster material using the onionskin test
Initiation of insensitive high explosive (IHE) formulations requires the use of a booster explosive in the initiation train. Booster material selection is crucial, as the initiation must reliably function across some spectrum of physical parameters. The interest in Diaminoazoxyfurazan (DAAF) for this application stems from the fact that it possesses many traits of an IHE but is shock sensitive enough to serve as an explosive booster. A hemispherical wave breakout test, termed the onionskin test, is one of the methods used to evaluate the performance of a booster material. The wave breakout time-position history at the surface of a hemispherical IHE charge is recorded and the relative uniformity of the breakout can be quantitatively compared between booster materials. A series of onionskin tests were performed to investigate breakout and propagation diaminoazoxyfurazan (DAAF) at low temperatures to evaluate ignition and detonation spreading in comparison to other explosives commonly used in booster applications. Some wave perturbation was observed with the DAAF booster in the onionskin tests presented. The results of these tests will be presented and discussed.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Liquid salt environment stress-rupture testing
Disclosed herein are systems, devices and methods for stress-rupture testing selected materials within a high-temperature liquid salt environment. Exemplary testing systems include a load train for holding a test specimen within a heated inert gas vessel. A thermal break included in the load train can thermally insulate a load cell positioned along the load train within the inert gas vessel. The test specimen can include a cylindrical gage portion having an internal void filled with a molten salt during stress-rupture testing. The gage portion can have an inner surface area to volume ratio of greater than 20 to maximize the corrosive effect of the molten salt on the specimen material during testing. Also disclosed are methods of making a salt ingot for placement within the test specimen.
36 MATERIALS SCIENCE↗
Baseline Evaluation of Eddy Current Testing for PWSCC Susceptible Materials
This technical letter report (TLR) summarizes technical efforts focused on assessments of eddy current testing (ET) examinations using various mockups for obtaining baseline surface and subsurface conditions on materials susceptible to primary water stress corrosion cracking (PWSCC). The scope of this work included nondestructive evaluation characterization of weld surfaces on mockups that were identified for application of various surface peening approaches. All of the work reported here was conducted prior to planned peening of these surfaces to evaluate pre-peening conditions, and subsequently compare the baseline ET results with post-peening conditions. The overall objective of this work was to demonstrate the effectiveness of employing ET for detecting near/sub-surface defects on mockups in support of peening assessments for surface stress improvement and mitigation of PWSCC in these materials. The U.S. Nuclear Regulatory Commission has communicated their expectations to industry for eddy current inspections of peened, inlayed, and onlayed dissimilar metal welds and is interested in better understanding detection performance and sensitivity of ET techniques to support the development of a technical foundation for detection and sizing limits, effective ET qualification requirements, alternate inspections or programs for evaluating PWSCC mitigation processes, and development of ET acceptance standards that can be referenced in the appropriate American Society of Mechanical Engineers Section XI Code Case (N-770-X). This TLR provides a technical discussion focused on the ET data obtained, the analysis, and conclusions drawn from surface examinations of various plate standards and actual components extracted from cancelled nuclear power plants.
36 MATERIALS SCIENCE↗
Prioritizing Off-Gas Metrics: A Guide for Comparable Off-Gas Capture Testing
The Material Recovery and Waste Form Development (MRWFD) off-gas team had a workshop, hosted by Idaho National Laboratory (INL), to align goals and expectations for off-gas research. The workshop included team members from four national laboratories. The workshop focused on defining distinct R&D phases with specific metrics, outlining standard test and measurement protocols for Iodine and Krypton/Xenon sorbents, brainstorming approaches to future disruptive technologies, and recognizing parameters with more inherent risk, requiring more rigorous evaluation. This report will serve as a guide for future off-gas work. Its purpose is to foster efficient collaboration across diverse research facilities and invite direct comparison of materials and results.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗