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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Geometrical gauge factor of directional electric potential drop sensors for creep monitoring

Directional electric potential drop measurements can be exploited for in-situ monitoring of creep in metals. The sensor monitors the variation in the ratio of the resistances measured simultaneously in the axial and lateral directions using a square-electrode configuration. This technique can efficiently separate the mostly isotropic common part of the resistivity variation caused by reversible temperature variations from the mostly anisotropic differential part caused by direct geometrical and indirect material effects of creep. Initially, this ratio is roughly proportional to the axial creep strain, while at later stages, the resistance ratio increases even faster with creep strain because of the formation of directional discontinuities such as preferentially oriented grain boundary cavities and multiple-site cracks in the material. Similarly to ordinary strain gauges, the relative sensitivity of the sensor is defined as a gauge factor that can be approximated as a sum of geometrical and material parts. This work investigated the geometrical gauge factor by analytical and experimental means. We found that under uniaxial stress square-electrode sensors exhibit geometrical gauge factors of about 4 and 5 in the elastic and plastic regimes, respectively, i.e., more than twice those of conventional strain gauges. Experimental results obtained on 304 stainless steel using a square-electrode electric potential drop creep sensor agree well with our theoretical predictions.

36 MATERIALS SCIENCE↗
DOE OSTI2018

In Situ High Pressure Hydrogen Tribological Testing of Common Polymer Materials Used in the Hydrogen Delivery Infrastructure

High pressure hydrogen gas is known to adversely affect metallic components of compressors, valves, hoses, and actuators. However, relatively little is known about the effects of high pressure hydrogen on the polymer sealing and barrier materials also found within these components. More study is required in order to determine the compatibility of common polymer materials found in the components of the hydrogen fuel delivery infrastructure with high pressure hydrogen. As a result, it is important to consider the changes in physical properties such as friction and wear in situ while the polymer is exposed to high pressure hydrogen. In this protocol, we present a method for testing the friction and wear properties of ethylene propylene diene monomer (EPDM) elastomer samples in a 28 MPa high pressure hydrogen environment using a custom-built in situ pin-on-flat linear reciprocating tribometer. Representative results from this testing are presented which indicate that the coefficient of friction between the EPDM sample coupon and steel counter surface is increased in high pressure hydrogen as compared to the coefficient of friction similarly measured in ambient air.

hydrogen↗
DOE OSTI2024

Report on RIA Relevant Modified Burst Testing of ATF Cladding Materials

The mechanical performance of accident-tolerant fuel (ATF) cladding candidates in light-water reactors (LWRs) must be similar to or better than that of current conventional nuclear fuel claddings to reduce dose to the public and ensure a that the core coolable geometry is maintained during a postulated reactivity-initiated accident (RIA) in light-water reactors (LWRs). During an RIA event, the rapid thermal expansion of nuclear fuel can deform the cladding once the fuel–cladding gap closes. In some cases, the pellet–cladding mechanical interaction (PCMI) can induce mechanical failure in ATF candidates. Thus, the mechanical response of ATF cladding must be investigated by mimicking the conditions of RIA and potentially performing Transient Reactor Test (TREAT) experiments to establish or verify the safety envelope. The work presented in this report investigated the failure behavior of as-received, hydrided, and chromium-coated (Cr-coated) Zircaloy-4 (Zry-4) cladding tube under strain-driven mechanical conditions, mimicking postulated RIA loading conditions. Mechanical testing was performed at 300°C via modified burst test (MBT) equipment with pulse width control previously developed under the Department of Energy’s (DOE’s) Advanced Fuel Campaign (AFC). The mechanical strains were determined using 2D digital image correlation (DIC) techniques. The base Zry-4 acquired by Cameco Inc. was in stress-relieved annealed (SRA) condition. Because of the observed large deformation of the cladding tubes, the failure strain definition was updated for the MBT, which can also be applied to other tube tests where significant bulging (out-of-plane deformation) is present. The failure strain was determined to be affected by the speed of the test or the RIA event. As the RIA-like event duration decreased from 75 to 15 ms, the failure strain decreased 5, 7 and 1% for as-received, hydrided, and Cr-coated specimens, respectively. Fractography on the Cr-coated specimens indicated the presence of two failure mechanisms: (i) crenulation at the outer surface of the coating due to the tensile tractional forces along with coatings grain microstructure and (ii) formation of critical defect at the coating/cladding interface that initiated coating rupture after severe plastic deformation of the Zry- 4 substrate. Based on the MBT results and fractography observations, performing mechanical property testing at high strain rates—in particular on Cr-coated tubes—and semi-integral TREAT experiments are required future efforts to ensure ATF cladding performance during transients. The testing recommended would also inform the development of long-term generalized cladding technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A probabilistic creep model incorporating test condition, initial damage, and material property uncertainty

Uncertainty is prevalent in the creep resistance of alloys, where at elevated temperature and low pressure, rupture can range across logarithmic decades. In this study, a probabilistic continuum-damage-mechanics (CDM)-based model is derived to capture the uncertainty of creep resistance. To meet this objective, creep data for alloy 304 Stainless Steel is gathered. A constitutive model, “Sinh”, is calibrated deterministically to determine the statistical variability of the material properties. Three sources of uncertainty are injected into the model: test condition (stress and temperature), initial damage, and material properties. Probabilistic simulations are carried out by (a) calibrating probability distribution functions (pdfs) for each source of uncertainty (b) randomly sampling the pdfs using Monte Carlo methods and (c) executing simulations to replicate the uncertain creep behavior. A sensitivity analysis is performed to evaluate the relative effect of each source of uncertainty. In full probabilistic simulations, the cumulative uncertainty of creep behavior is evaluated. The probabilistic model accurately predicts the creep deformation and rupture of the available experiments. The probabilistic model is validated for interpolation but lacks extrapolation ability. Several future works are proposed to further improve the model.

36 MATERIALS SCIENCE↗
DOE OSTI2018

An advanced experimental design for modified burst testing of nuclear fuel cladding materials during transient loading

Here an advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

EXPERIMENTAL INVESTIGATION OF IMPACT AND ABRASION EROSIONS AT LOW VELOCITIES AND COMPARISON TO CFD SIMULATIONS

Wear and erosion damage of materials due to interaction of the surface with solid particles is a major issue in various industries. This phenomenon is common in mining, and oil and gas production, causing damages worth millions of dollars annually. Erosion is becoming a critical issue in renewable technologies as well such as concentrated solar thermal power (CSP) systems. In some CSP systems, solid particles can be used to absorb solar energy and enhance the heat storage. However, these particles may cause significant amount of wear to system components while falling down the power tower at high temperatures. This damage can be costly, and therefore, investigating solid particle erosion in CSP systems is of utmost importance. In collaboration with Sandia National Laboratories, the erosion durability of particle-based Generation 3 (GEN3) CSP system are being investigated both experimentally and numerically. In order to develop models to predict the life time of such systems, erosion under low temperature and low particle impact velocity conditions are investigated. Impact erosion and abrasion erosion experiments are conducted for low velocity and low temperature conditions for various candidate particles and containment materials. The impact erosion testing is conducted for three different containment material including Stainless Steel 316, special refractory material, and Inconel 740H with CARBOBEAD HSP 40/70 ceramic particles and UNIWEST 430 silica quartz erodent particles. All this data will enable us to better model the erosion behavior of these materials and provides a data base to investigate erosion at high temperatures experienced in CSP systems. At low temperatures and low velocities the softer refractory material exhibited greater erosion by an order-of-magnitude when compared to the harder Inconel 740H and Stainless Steel 316 materials. The overall erosion ratio for all the candidate materials was of the order of 10 -7 kg/kg. We also investigated wear in containment materials from sliding of particles over the surfaces at very low angles, also referred to as abrasion. For the velocities expected in a typical 1MW CSP plant, visible surface wear was noted, but the overall mass loss per unit surface area was found to be dependent on the hardness of the containment material tested. In this manuscript, we also present Computational Fluid Dynamics (CFD) simulations and erosion prediction of low velocity impact erosion. These results are compared with the experimental data to evaluate the ability of available erosion models for predicting erosion under these conditions.

14 SOLAR ENERGY↗
DOE OSTI2022

Small Punch Testing of Molybdenum-99 Targets

Northstar Medical Radioisotopes is developing an accelerator-based method to produce 99 Mo, which is a parent isotope of the commonly used 99m Tc medical isotope. The Mo targets being designed for the accelerator will be produced from enriched 100 Mo, also known as aMo . Pressed and sintered powder feedstock is used to fabricate aMo targets, producing 29 mm disk-shaped targets. The targets are subjected to 1–6 days in line of an electron beam with subsequent dissolution of the disk to retain the 99 Mo, which decays to 99m Tc at radio-pharmacies. The press and sinter method is advantageous because the inherent porosity produced by this method enables increased surface area and therefore increased flow of dissolution media, decreasing the dissolution time and reducing the need for a highly acidic media. Although porosity aids in dissolution, it reduces the mechanical strength and ductility. Targets require good mechanical integrity when subjected to the conditions in the accelerator. Therefore, Northstar is seeking methods to rapidly test disk samples after fabrication to assure mechanical performance metrics are achieved. This report details the design and testing of a small punch test (SPT) that accommodates the 29 mm disk. Initial data were used to relate the SPT data to tensile properties, such as the yield strength (YS), ultimate tensile strength (UTS), and total elongation to failure. Although the SPT has been established as a somewhat reliable method for testing metallic materials, few studies have applied the SPT to refractory materials such as Mo. Based on tensile testing performed at Oak Ridge National Laboratory on different Mo samples, correlation between the Mo tensile and SPT properties could be performed, establishing a standard calibration that could be applied to other Mo samples. To test the efficacy of the SPT with Mo, multiple different disk batches were fabricated under different conditions (e.g., pressure, lubricant) with commercially available pure Mo powder. Generally, only a UTS could be well defined because the press and sinter disks failed under brittle fracture, making it difficult to determine the YS and elongation. Compared with disks fabricated with aMo powder, the aMo samples underperformed their pure nat Mo counterparts. This report summarizes the current status of the SPT, but further evaluation is needed before it can be applied as a reliable quality assurance tool.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2022

The Nuclear Fuel Cycle: Safe Management of Spent Nuclear Fuel

The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objective s are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum SNF from research reactors to corrosion in poor quality water (wet storage). In dry storage, aluminum SNF can release hydrogen via radiolysis of the hydrated oxides on the aluminum cladding. Austenitic stainless steel canisters used for dry storage are susceptible to chloride-induced stress corrosion cracking (outside-in attack) that threaten the confinement boundary provided by the canister. This paper further describes these challenges, among others, and the formulated solutions to support extended safe storage of SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Adsorption of a catalytically accessible polyoxometalate in a mesoporous channel-type metal–organic framework

A Keggin-type polyoxometalate (H 3 PW 12 O 40 ) was incorporated into a mesoporous Zr-based MOF (NU-1000) via an impregnation method in aqueous media, resulting in the hybrid material, PW 12 @NU-1000. The POM@MOF composite was characterized by a suite of physical methods, indicating the retention of crystallinity and high porosity of the parent MOF. The hybrid material was also stable to leaching in aqueous media at varying pH. Lastly, the material was tested as a heterogeneous catalyst for the oxidation of 2-chloroethyl ethyl sulfide using hydrogen peroxide as the oxidant. PW 12 @NU-1000 was shown to have a higher catalytic activity than either of the individual constituents alone.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of Surveillance Test Articles with Reduced Dimensions and Material Volumes to Support MSR Materials Degradation Management

This report details the efforts toward developing new surveillance test article designs with reduced dimensions and material volumes to support materials surveillance technology development for advanced reactors. Two fabrication methods for the surveillance test articles are described. Welded test articles were fabricated with 316H and A617 materials, and interlocking test articles were fabricated using A709 and titanium-zirconium-molybdenum (TZM). The preliminary results demonstrate the successful design and testing of the flat surveillance test articles. The report also describes ongoing efforts to use an induction heating test setup to increase the heat up and cool down rates in testing the surveillance test articles. A brief description of the planned FY-24 work is provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

DEVELOPMENT OF A SENSOR NETWORK TEST BED FOR ISD MATERIALS AND STRUCUTRAL CONDITION MONITORING

The P Reactor at the Savannah River Site is one of the first reactor facilities in the US DOE complex that has been placed in its end state through in situ decommissioning (ISD). The ISD end state consists of a grout-filled concrete civil structure within the concrete frame of the original building. To evaluate the feasibility and utility of remote sensors to provide verification of ISD system conditions and performance characteristics, an ISD Sensor Network Test Bed has been designed and deployed at the Savannah River National Laboratory. The test bed addresses the DOE-EM Technology Need to develop a remote monitoring system to determine and verify ISD system performance. Commercial off-the-shelf sensors have been installed on concrete blocks taken from walls of the P Reactor Building. Deployment of this low-cost structural monitoring system provides hands-on experience with sensor networks. The initial sensor system consists of: (1) Groutable thermistors for temperature and moisture monitoring; (2) Strain gauges for crack growth monitoring; (3) Tiltmeters for settlement monitoring; and (4) A communication system for data collection. Preliminary baseline data and lessons learned from system design and installation and initial field testing will be utilized for future ISD sensor network development and deployment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2017

Creep Fatigue Testing -- Rev. 9

This practice is to be used by the Materials Science and Engineering Department (B120) in determining the fatigue and creep-fatigue properties of nominally homogeneous materials from test specimens subjected to uniaxial forces. This practice is restricted to the testing of uniform gage section test specimens (from the Scope Section of ASTM Standard E 606-04).

36 MATERIALS SCIENCE↗
DOE OSTI2024

Transfer Tube Material Properties Obtained from Test Data

A series of tension and compression tests were performed on typical transfer tubes made of 304L corrosion-resistant steel. The results were compiled and converted to true stress and true plastic strain for comparison to numerical models. The parameters of a Johnson-Cook constitutive model and fracture criterion were calibrated to mimic the test data. Comparisons were made between the converted data and the empirical formulas in the Johnson-Cook model. Simulations were performed to evaluate the implementation of the formula in finite element analysis, and the parameters were adjusted to improve the correlation.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Response of nickel surface to pulsed fusion plasma radiations

Nickel based alloys are being projected as suitable materials for some components of the next generation fusion reactor because of compatible thermal, electrical and mechanical properties. Pure nickel material is tested here for possibility of similar application purpose. Nickel samples (> 99.5 % purity) are exposed here to plasma radiations produced due to D-D fusion reaction inside an 11.5 kJ plasma focus device. The changes in the physical properties of the nickel surface at microscopic level which in turn change the mechanical properties are analyzed using scanning electron microscope, optical microscope, glancing incident X-ray diffractometer and Vicker's hardness gauge. The results are reported here.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Corrosion Testing Trade Study on Testing Methods in Molten Salts

In assessing molten salt compatibility with structural materials, static testing is often reported because it is easy to conduct. However, in order to assess mass transfer, where material dissolves in the salt in the hot section and may precipitate out of the liquid in colder regions, a flowing salt experiment in a temperature gradient is needed. While a forced convection or pumped loop produces prototypic flowing conditions, it also is a multi-million dollar investment. Among the less expensive dynamic testing options, the thermal convection loop (TCL) was favored during the 1960’s development of molten salt reactor technology. Because the workforce and core facilities are currently available to conduct TCL experiments, it is the least expensive strategy to evaluate candidate structural alloys in flowing molten salt conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermal Gradient and Neutron Irradiation Experiment Design for Fusion Reactor Materials in the Advanced Test Reactor

This work outlines a hypothetical coupled thermal gradient and neutron irradiation experiment in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Although the ATR is a thermal spectrum test reactor and doesn’t inherently produce a flux spectrum dominated by the high-energy neutrons typical in a fusion reactor, it’s multitude of experiment positions and dynamic flux environment make it a suitable platform for investigating fusion related issues.

36 MATERIALS SCIENCE↗