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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Molten salt environment creep testing extensometry system

Disclosed herein are systems, devices and methods for creep testing selected materials within a high-temperature molten salt environment. Exemplary creep testing systems include a load train for holding a test specimen under a load within a heated inert gas vessel. An extensometry system can be included to measure elongation of the test specimen while under load. The extensometry system can include fixed members and axially translating member that move along with the elongation of the test specimen, and the system can include a sensor to measure the relative axial motion between such components to measure elongation of the test specimen over time. The test specimen can include a cylindrical gage portion having an internal void filled with a molten salt during creep testing to simulate the corrosive effect of the molten salt on the specimen material during testing.

Ren, Wenlin↗
DOE OSTI2018

The Development of a DC Breakdown Voltage Test for Photovoltaic Insulating Materials

The ability of electrical insulating materials within a module to act as insulators is a key safety requirement for PV technology. Direct current breakdown voltage is therefore now specified for relied-upon insulator materials in the IEC 61730-1 safety standard. To fulfill that requirement, a new test method has been developed within the IEC TS 62788-2 backsheet standard for the measurement of breakdown voltage. The development of the test will be described, including the verification of the most critical parameters relative to factors such as defect population(s), dielectric medium, electrode size, electrode surface roughness, maximum current limit, moisture conditioning, number of replicate specimens, rate of voltage rise, specimen thickness, test polarity, and test temperature. Many of these parameters were specifically explored in discovery experiments as the test method was developed. An interlaboratory round-robin (R-R) experiment was conducted to further validate the test method by quantifying its repeatability and reproducibility. The materials examined in the R-R include the backsheet materials: polyethylene terephthalate (PET, two thicknesses), polyvinyl fluoride (PVF), and laminated PVF/PET/PVF ('TPT') as well as the encapsulants poly(ethylene-co-vinyl acetate) (EVA), and polyvinyl butyral (PVB). The precision of the test method as well as key factors contributing to the measurement will be described.

14 SOLAR ENERGY↗
DOE OSTI2020

Graphite Licensing (ASME) and Testing (ASTM) Technical Assistance

DOE ART Graphite R&D program is currently involved in developing the structural code for licensing the new High Temperature Reactor (HTR) designs through the ASME Boiler Pressure Vessel Code (BPVC). The basic structural design code for graphite core components has been written, approved, and within the BPVC since 2012. However, several areas within the graphite code remain to be improved, specifically the addition of material property changes resulting from environmental degradation (neutron flux and oxidation). The incorporation of this data, which can significantly affect the graphite components and alter the structural integrity of the core during operation, is proving to be an important activity. How this data will be incorporated within the code and how it will be used for license applications will be critical to a number of HTR applicants in the future. These modifications to the code must take into account the unique irradiation and oxidation response of the wide spectrum of available nuclear graphite grades currently under consideration for HTR applications. Additionally, the ASME BPVC requires material property testing within the expected operational conditions of the HTR core including neutron radiation flux, elevated temperatures, and after chronic/acute oxidation of the graphite components. Currently, no standardized graphite testing methods are approved for these environmental conditions. New American Society for Testing and Materials (ASTM) standardized tests for elevated temperature testing, small sample testing, and testing after oxidation must be developed to support the requirements within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multi-lab study on the pure-gas permeation of commercial polysulfone (PSf) membranes: Measurement standards and best practices

Gas-separation membranes are a critical industrial component for a low-carbon and energy-efficient future. As a result, many researchers have been testing membrane materials over the past several decades. Unfortunately, almost all membrane-based testing systems are home-built, and there are no widely accepted material standards or testing protocols in the literature, making it challenging to accurately compare experimental results. Further, in this multi-lab study, ten independent laboratories collected high-pressure pure-gas permeation data for H 2 , O 2 , CH 4 , and N 2 in commercial polysulfone (PSf) films. Equipment information, testing procedures, and permeation data from all labs were collected to provide (1) accepted H 2 , O 2 , CH 4 , and N 2 permeability values at 35°C in PSf as a reference standard, (2) statistical analysis of lab-to-lab uncertainties in evaluating permeability, and (3) a list of best practices for sample preparation, equipment set-up, and permeation testing using constant-volume variable-pressure apparatuses. Results summarized in this work provide a reference standard and recommended testing protocols for pure-gas testing of membrane materials.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2010

SUMMARY PLAN FOR BENCH-SCALE REFORMER AND PRODUCT TESTING TREATABILITY STUDIES USING HANFORD TANK WASTE

This paper describes the sample selection, sample preparation, environmental, and regulatory considerations for shipment of Hanford radioactive waste samples for treatability studies of the FBSR process at the Savannah River National Laboratory and the Pacific Northwest National Laboratory. The U.S. Department of Energy (DOE) Hanford tank farms contain approximately 57 million gallons of wastes, most of which originated during the reprocessing of spent nuclear fuel to produce plutonium for defense purposes. DOE intends to pre-treat the tank waste to separate the waste into a high level fraction, that will be vitrified and disposed of in a national repository as high-level waste (HLW), and a low-activity waste (LAW) fraction that will be immobilized for on-site disposal at Hanford. The Hanford Waste Treatment and Immobilization Plant (WTP) is the focal point for the treatment of Hanford tank waste. However, the WTP lacks the capacity to process all of the LAW within the regulatory required timeframe. Consequently, a supplemental LAW immobilization process will be required to immobilize the remainder of the LAW. One promising supplemental technology is Fluidized Bed Steam Reforming (FBSR) to produce a sodium-alumino-silicate (NAS) waste form. The NAS waste form is primarily composed of nepheline (NaAlSiO{sub 4}), sodalite (Nas[AlSiO{sub 4}]{sub 6}Cl{sub 2}), and nosean (Na{sub 8}[AlSiO{sub 4}]{sub 6}SO{sub 4}). Semivolatile anions such as pertechnetate (TcO{sub 4}{sup -}) and volatiles such as iodine as iodide (I{sup -}) are expected to be entrapped within the mineral structures, thereby immobilizing them (Janzen 2008). Results from preliminary performance tests using surrogates, suggests that the release of semivolatile radionuclides {sup 99}Tc and volatile {sup 129}I from granular NAS waste form is limited by Nosean solubility. The predicted release of {sup 99}Tc from the NAS waste form at a 100 meters down gradient well from the Integrated Disposal Facility (IDF) was found to be comparable to immobilized low-activity waste glass waste form in the initial supplemental LAW treatment technology risk assessment (Mann 2003). To confirm this hypothesis, DOE is funding a treatability study where three actual Hanford tank waste samples (containing both {sup 99}Tc and {sup 125}I) will be processed in Savannah River National Laboratory's (SRNL) Bench-Scale Reformer (BSR) to form the mineral product, similar to the granular NAS waste form, that will then be subject to a number of waste form qualification tests. In previous tests, SRNL have demonstrated that the BSR product is chemically and physically equivalent to the FBSR product (Janzen 2005). The objective of this paper is to describe the sample selection, sample preparation, and environmental and regulatory considerations for treatability studies of the FBSR process using Hanford tank waste samples at the SNRL. The SNRL will process samples in its BSR. These samples will be decontaminated in the 222-S Laboratory to remove undissolved solids and selected radioisotopes to comply with Department of Transportation (DOT) shipping regulations and to ensure worker safety by limiting radiation exposure to As Low As Reasonably Achievable (ALARA). These decontamination levels will also meet the Nuclear Regulatory Commission's (NRC's) definition of low activity waste (LAW). After the SNRL has processed the tank samples to a granular mineral form, SRNL and Pacific Northwest National Laboratory (PNNL) will conduct waste form testing on both the granular material and monoliths prepared from the granular material. The tests being performed are outlined in Appendix A.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of Real-Time, In-pile Creep Test Rigs for Characterizing the Structural Materials of Nuclear Components

New and improved materials are being considered for supporting both existing and next-generation nuclear reactors. Reactor materials can significantly degrade with time, thus limiting or altering their properties in harsh reactor environments. To accurately understand such material degradation, real-time data obtained under prototypic irradiation conditions are required. In particular, understanding the creep behavior of materials exposed to irradiation and elevated temperatures is essential for safety concern evaluations. To provide these capabilities, Idaho National Laboratory (INL)’s High Temperature Test Laboratory (HTTL) developed several instrumented test rigs for obtaining real-time data from specimens in well-controlled pressurized-water reactor (PWR) coolant conditions at the Materials Test Reactor. This technical report focuses on INL’s efforts to evaluate and enhance the former creep test rig prototype that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig can detect changes in the length of creep specimens, which is useful for measuring thermal expansion and creep deformation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Examination of an Optical Transmittance Test for Photovoltaic Encapsulation Materials (Presentation)

The optical transmittance of encapsulation materials is a key characteristic for their use in photovoltaic (PV) modules. Changes in transmittance with time in the field affect module performance, which may impact product warranties. Transmittance is important in product development, module manufacturing, and field power production (both immediate and long-term). Therefore, an international standard (IEC 62788-1-4) has recently been proposed by the Encapsulation Task-Group within the Working Group 2 (WG2) of the International Electrotechnical Commission (IEC) Technical Committee 82 (TC82) for the quantification of the optical performance of PV encapsulation materials. Existing standards, such as ASTM E903, are general and more appropriately applied to concentrated solar power than to PV. Starting from the optical transmittance measurement, the solar-weighted transmittance of photon irradiance, yellowness index (which may be used in aging studies to assess durability), and ultraviolet (UV) cut-off wavelength may all be determined using the proposed standard. The details of the proposed test are described. The results of a round-robin experiment (for five materials) conducted at seven laboratories to validate the test procedure using representative materials are also presented. For example, the Encapsulation Group actively explored the measurement requirements (wavelength range and resolution), the requirements for the spectrophotometer (including the integrating sphere and instrument accessories, such as a depolarizer), specimen requirements (choice of glass-superstrate and -substrate), and data analysis (relative to the light that may be used in the PV application). The round-robin experiment identified both intra- and inter-laboratory instrument precision and bias for five encapsulation materials (encompassing a range of transmittance and haze-formation characteristics).

14 SOLAR ENERGY↗
DOE OSTI2011

Use of Melt Flow Rate Test in Reliability Study of Thermoplastic Encapsulation Materials in Photovoltaic Modules

Use of thermoplastic materials as encapsulants in photovoltaic (PV) modules presents a potential concern in terms of high temperature creep, which should be evaluated before thermoplastics are qualified for use in the field. Historically, the issue of creep has been avoided by using thermosetting polymers as encapsulants, such as crosslinked ethylene-co-vinyl acetate (EVA). Because they lackcrosslinked networks, however, thermoplastics may be subject to phase transitions and visco-elastic flow at the temperatures and mechanical stresses encountered by modules in the field, creating the potential for a number of reliability and safety issues. Thermoplastic materials investigated in this study include PV-grade uncured-EVA (without curing agents and therefore not crosslinked);polyvinyl butyral (PVB); thermoplastic polyurethane (TPU); and three polyolefins (PO), which have been proposed for use as PV encapsulation. Two approaches were used to evaluate the performance of these materials as encapsulants: module-level testing and a material-level testing.

14 SOLAR ENERGY↗
DOE OSTI2012

Bridging the Gap: Linking Simulation and Testing

The Materials Genome Initiative (MGI) which is a key enabler for the Advanced Manufacturing Partnership, announced in 2011 by U.S. President Barack Obama, was established to accelerate the development and deployment of advanced materials. The MGI is driven by the need to "bridge the gap" between (I) experimental results and computational analysis to enable the rapid development and validation of new mateirals, and (II) the processes required to convert these materials into useable goods.

bridge the gap↗
DOE OSTI2015

Investigation of the Effect of In-Plane Fiber Waviness in Composite Materials through Multiple Scales of Testing and Finite Element Modeling

Defects in materials can reduce strengths and lifetimes of manufactured parts. The number of possible defects increases with the complexity inherent in composite materials. The wind industry uses composite wind turbine blades in which the manufacturing process induces a number of defects. In order for the wind industry to continue sustainable expansion, the effects of defects must be better understood. In-plane (IP) fiber waviness is the focus of this work. The three main parts of this work include testing on the coupon level, modeling on the coupon level, and testing of beams in four-point bending (with and without defects). The coupon level testing includes partial IP waves, similar to those in manufactured parts, rather than full width IP waves. This allows investigation into complex interactions and varying failure mechanisms caused by the fiber misalignment gradient. Partial waves are also modeled to both validate testing as well as to increase robustness of a previously developed progressive damage modeling method. Lastly, a sandwich beam test specimen for testing in 4-point bending is developed to investigate the effects of fiber waviness in both tension and compression when loaded in flexure.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Preliminary Procedures and Acceptance Criteria for in-situ Structural Materials Surveillance for MSR

This report describes the initial development of the technical basis for a materials surveillance technology that, when fully developed and validated, can be used by stakeholders to develop and implement a materials surveillance program to manage materials degradation during reactor operations. The procedures described in this report aim to bound the detrimental effects of the reactor environment on the creep-fatigue life of the reactor components through materials surveillance. Specifically, the report describes procedures for selecting critical locations within the reactor components for surrogate materials surveillance and the sizing and placements of passively-actuated materials surveillance test articles under development by the US Department of Energy, Advanced Reactor Technologies Program. These test articles apply cyclic thermomechanical load to a surrogate sample of the component structural material passively through thermal expansion mismatch within the articles, without requiring any penetrations in the reactor coolant boundary. The procedures can be used to size materials surveillance test articles so that they will fail before the corresponding structural component. These articles would be exposed to the component operating environment, located to experience a bounding radiation fluence and representative coolant exposure. They can be monitored during reactor operations to ensure that they have not failed under the combination of creep-fatigue damage and environmental degradation. The report describes the key technical decisions and the corresponding rational for using these procedures to provide advanced warning of any impending material failure. An appendix describes the technical basis and the procedure for sizing a family of passively-actuated test articles. Finally, the report describes future work needed to complete the technical basis underlying this materials surveillance technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Chemical and Radiological Compatibility Testing of 3D Printed Materials

To dramatically increase the adaptability, performance, and safety of processes in support of the Defense Waste Processing Facility (DWPF), Savannah River National Lab (SRNL) plans to perform chemical and radiological compatibility testing on a wide variety of 3D printed materials of interest. The 3D printing process provides numerous strategic operational benefits such as rapid prototyping of complex designs and geometry specific to the needs of the nuclear waste disposition process, as well as on-demand rapid prototyping and iteration with materials that aren’t as accessible through traditional manufacturing methods. Reaction chemistry in simulated waste batches can be matched closely to its radioactive counterpart, but glass reactor vessels have limitations. Vessel geometry can play a big factor in mixing transport limitations, process chemistry, and degradation reaction kinetics. In addition, additive manufacturing allows for much more detailed vessel design than traditional alternatives. Waste processing techniques in DWPF also encounter extreme chemical environments including high pH, strong acids, abrasive slurries, and significant irradiation. To meet these challenges, a matrix of various polymer, ceramic, and metal additive manufacturing materials have been exposed to a suite of chemical environments of interest as well as radioactive dose (such as gamma radiation from 60 Co) to properly test their durability under these conditions. Mass change has been monitored over a period of up to a week in these conditions, as well as added characterization for surface modification through Scanning Electron Microscopy/Electron Dispersive X-ray analysis (SEM/EDX). Further chemical characterization has been monitored through Fourier-Transform InfraRed Spectroscopy (FTIR), with planned investigation via thermal and tensile strength degradation. While the direct product of this research is identification of material(s) that can withstand specific hazardous environments encountered by the mercury water wash tank in DWPF process simulation experiments, the reference base of materials will be used for many other nuclear processes in the pursuit of rapidly developed, cost-efficient, and highly specific devices for environmental remediation and much more.

Wilson, Nathan W. [Savannah River National Laborat↗
DOE OSTI2022

Final Status Report Confirming Completion of Mechanical Testing of HFIR Irradiated Nanocomposite Materials

This report provides a final status report on the in-cell mechanical testing of High Flux Isotope Reactor (HFIR)-irradiated nanodispersion-strengthened materials at the Irradiated Materials Examination and Testing (IMET) facility. All tension tests were performed at room temperature with a nominal strain rate of 0.018 mm/mm/min using shoulder-loading grip sets by following the standard testing procedure in ASTM E8/E8M. Round 1 (unirradiated) and round 2 (0.7 dpa) testing was completed in FY 2021. This report confirms completion of round 3 (1.4 dpa) and round 4 (2.1 dpa) tensile testing in FY 2022 with examples of test data collected on select sample conditions. Additionally, the report includes IDs of samples selected for shipment to Idaho National Laboratory (INL) for additional characterization.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Promoting regulatory acceptance of combined ion and neutron irradiation testing of nuclear reactor materials: Modeling and software considerations

As the needs for the nuclear energy industry continue to evolve in the 21st century, timely adoption of new technological solutions acceptable to regulatory agencies is critical. Quantitative prediction of radiation damage in materials and its impact on mechanical properties is a key component of licensing and regulatory decisions regarding nuclear power plants. Accelerated testing methodologies such as combined ion and neutron irradiation data sets are crucial for the development and deployment of new materials and new manufacturing methods (e.g., additive manufacturing). However, regulatory acceptance of accelerated testing methodologies is necessary for their adoption. Further, the present work discusses the fundamental basis for comparing ion- and neutron-induced material microstructures, the theory behind interpreting radiation damage across length and time scales and radiation types, and the codes, standards, and quality assurance concerns surrounding different modeling methods and software. In particular, recommendations are given as to the path forward that will enable national laboratories, academia, and industry to develop the modeling and software basis for regulatory acceptance of the combined use of ion and neutron irradiation for material performance evaluation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Testing of LWR Surveillance Materials from the High Fluence UCSB ATR-2 Irradiation

The reactor pressure vessel (RPV) in a light water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. Of the many significant issues discussed regarding irradiation effects, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. As also stated in a companion paper by Odette, et. al., the primary objective of the DOE-sponsored Light Water Reactor Sustainability (LWRS) Program RPV task is to develop robust predictions of transition temperature shifts (TTS) at high fluence (φt) to at least 10{sup 20} n/cm{sup 2} (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the discovery and experimental verification of 'late blooming' Mn-Ni-enriched precipitates (MNPs) with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. The new and existing databases will be combined to support development of physically based models of TTS for high fluence-low flux (φ < 10 {sup 11}n/cm{sup 2}-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 10{sup 20} n/cm{sup 2} (>1 MeV). The RPV task of the LWRS Program has worked with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar alloys irradiated under high flux conditions in test reactors, such as the UCSB ATR-2 experiment conducted at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Report on FY19 Testing in Support of Integrated EPP-SMT Design Methods Development

Experiments in support of the development of the integrated Elastic–Perfectly Plastic (EPP) analysis and Simplified Model Test (SMT) design methodology continued in FY 2019. Major progress was made in the verification of the newly developed two-bar SMT and single-bar SMT (SBSMT) methods by performing tests on Alloy 617 and SS 316H at high temperatures. Comparable results were achieved when the materials were tested under same conditions using both techniques. Because of its simple test configuration, SBSMT was the preferred testing technique used to evaluate elastic follow-up effect and generate data in support of SMT-based design curves. Evaluation of elastic follow-up was started with testing SS 316H at 815°C using SBSMT in this reporting period. As a result of this effort, the basic procedure and test protocol to provide data in support of the development of SMT-based design curves were established. This report also summarizes recent test results for the pressurization SMT using test articles with key features on Alloy 617 at intermediate strain ranges and at a low temperature of 850°C. Although more test results are needed in order to fully understand the effect of primary load on SMT creep-fatigue (CF) life, the limited test data indicate that the primary load reduced the SMT life at intermediate strain ranges and high temperatures. Recognizing the advantages of the newly developed SBSMT method, a new test geometry was designed to allow the elevation of the primary load effect on SMT CF life with the SBSMT technique. Additionally, selected tests were performed on SS 316H (heat 101076) to provide basic information about the heat of material. The tests performed included high-temperature tensile, fatigue, creep fatigue, and thermomechanical experiments. The results will be used to support the verification of the material constitutive model developed at Argonne National Laboratory.

36 MATERIALS SCIENCE↗