Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “containers”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Iodine Removal from Carbonate-Containing Alkaline Liquids Using Strong Base Resins, Hybrid Resins, and Silver Precipitation

The ability of several material types to remove aqueous iodine from a mildly alkaline, carbonate-rich nuclear waste stream was evaluated: strong base anion exchange resins (SBAR), hybrid resins, Ag-containing materials and Bi-containing hybrid resins. A combination of batch testing and flow through column testing were used in the evaluation. In batch testing, hybrid resins CHM-20, SIR-110-CE and RTBI were shown to have high efficiency for the removal of both iodide and iodate simultaneously, while Ag-containing materials and SBAR demonstrated high capacity for iodide removal. One example of each material type (CHM-20, A532E and Ionex 400) were further evaluated for their sorption isotherms and column performance. The Langmuir isotherm, or a Langmuir-Freundlich hybrid isotherm, best described the sorption of iodide to the CHM-20 hybrid resin and Purolite A532E. The Freundlich isotherm best described the uptake of iodate to CHM-20 and A532E and for both iodide and iodate to the Ag-containing Ionex-400. In column testing the Purolite A532E had exceptional performance for overall iodide removal. With the capacity demonstrated the A532E resin would exceed Class C waste classification before breakthrough initiated, and column change outs in processing would be dictated by eventual waste classification, not breakthrough. The Ionex 400, a Ag-zeolite, was observed to degrade over time in the column in the mild alkaline conditions whereas the hybrid CHM-20 was limited in the single pass through design and would be best suited for applications where iodide and iodate are present and recirculation of the column effluent is feasible. Here this work highlights the feasibility of commercially available materials to separate radioiodine from liquid environments.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Highly fluorescent purine-containing conjugated copolymers with tailored optoelectronic properties

Conjugated copolymers containing electron donor and acceptor units in their main chain have emerged as promising materials for organic electronic devices due to their tunable optoelectronic properties. In this work, we describe the use of direct arylation polymerization to create a series of fully π-conjugated copolymers containing the highly tailorable purine scaffold as a key design element. To create efficient coupling sites, dihalopurines are flanked by alkylthiophenes to create a monomer that is readily copolymerized with a variety of conjugated comonomers, ranging from electron-donating 3,4-dihydro-2H-thieno[3,4-b][1,4]dioxepine to electron-accepting 4,7-bis(5-bromo-3-hexylthiophen-2-yl)benzo[c][1,2,5]thiadiazole. The comonomer choice and electronic nature of the purine scaffold allow the photophysical properties of the purine-containing copolymers to be widely varied, with optical bandgaps ranging from 1.96–2.46 eV, and photoluminescent quantum yields as high as Φ = 0.61. Frontier orbital energy levels determined for the various copolymers using density functional theory tight binding calculations track with experimental results, and the geometric structures of the alkylthiophene-flanked purine monomer and its copolymer are found to be nearly planar. The utility of direct arylation polymerization and intrinsic tailorability of the purine scaffold highlight the potential of these fully conjugated polymers to establish structure–property relationships based on connectivity pattern and comonomer type, which may broadly inform efforts to advance purine-containing conjugated copolymers for various applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Airborne Release Fractions from Surrogate Nuclear Waste Fires Containing Lanthanide Nitrates and Depleted Uranium Nitrate in 30% Tributyl Phosphate in Kerosene

Airborne contaminants from fires containing nuclear waste represent significant health hazards and shape the design and operation of nuclear facilities. Much of the data used to formulate DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities,” from the U.S. Department of Energy, were taken over 40 years ago. The objectives of this study were to reproduce experiments from Pacific Northwest Laboratories conducted in June 1973 employing current aerosol measurement methods and instrumentation, develop an enhanced understanding of particulate formation and transport from fires containing nuclear waste, and provide modeling and experimental capabilities for updating current standards and practices in nuclear facilities. A special chamber was designed to conduct small fires containing 25 mL of flammable waste containing lutetium nitrate, ytterbium nitrate, or depleted uranium nitrate. Carbon soot aerosols showed aggregates of primary particles ranging from 20 to 60 nm in diameter. In scanning electron microscopy, ~200-nm spheroidal particles were also observed dispersed among the fractal aggregates. The 200-nm spherical particles were composed of metal phosphates. Airborne release fractions (ARFs) were characterized by leaching filter deposits and quantifying metal concentrations with mass spectrometry. The average mass-based ARF for 238 U experiments was 1.0 × 10 -3 with a standard deviation of 7.5 × 10 -4 . For the original experiments, DOE-HDBK-3010-94 states, “Uranium ARFs range from 2 × 10 -4 to 3 × 10 -3 , an uncertainty of approximately an order of magnitude.” Therefore, current measurements were consistent with DOE-HDBK-3010-94 values. ARF values for lutetium and ytterbium were approximately one to two orders of magnitude lower than 238 U. Furthermore, metal nitrate solubility may have varied with elemental composition and temperature, thereby affecting ARF values for uranium surrogates (Yb and Lu). In addition to ARF data, solution boiling temperatures and evaporation rates can also be deduced from experimental data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

tert-butoxides as precursors for atomic layer deposition of alkali metal containing thin films

Alkali metal containing materials have become increasingly attractive in a world hunting for sustainable energy materials and green functional devices. Lithium- and sodium battery technology, lead-free piezo- and ferroelectric devices, and record-breaking alkali doped tandem perovskite solar cells are among the applications where alkali metal-containing thin films get increasing attention. Atomic layer deposition (ALD) is one of the enabling thin film deposition techniques that offer chemical and geometrical versatility to realize the implementation of such thin films on an applicable scale. The drawback has until recently been a lack of available precursor chemistry that offers self-limiting growth that is fundamental to ALD. The alkali metal tert-butoxides have been shown to exhibit the necessary properties to facilitate saturating growth for Li-, Na-, K-, and Rb-containing compounds. However, the behavior of the tert-butoxides in ALD-growth has been considered difficult to unravel, with processes exhibiting limited control and low reproducibility. Very little has been reported on trends in reaction mechanisms as the mass of the alkali metal increases. Herein, we summarize the existing literature on the use of alkali metal tert-butoxides as precursors in ALD. We consider differences in the structure and behavior of the tert-butoxides as the alkali metal cation becomes heavier. In addition, we present precursor synthesis routes and key information on precursor structure, stability, and mechanistic behavior. Finally, we provide the first ever report of Cs-containing films by ALD to complement previous work on its lighter counterparts.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Chromium and Vanadium Incorporation in Sulfate-Containing Sodium Aluminoborosilicate Glass

In the composition space for Hanford low-activity nuclear waste glass, the waste loading of some formulations is limited by poor incorporation of sulfate. In particular, certain elements identified in previous studies tend to raise or lower sulfate solubility in borosilicate glass. The aim of the current study is to understand how vanadium and chromium each affect a sulfate-containing sodium aluminoborosilicate glass structure. Glass transition temperature, mass density, visible absorption, and Raman scattering were measured to investigate changes in the glass structure. Both Cr6+ (CrO42-) and Cr3+ were found in the chromium-containing glasses while vanadium primarily existed in the 5+ oxidation state in the vanadium-containing glasses. Electron probe microanalysis determined elemental compositions to assess retention of components. Increasing Cr2O3 caused saturation of Cr concentration within the glass and formation of crystalline eskolaite (Cr2O3). On the other hand, all the targeted V was incorporated into the glass. A fraction of batched S was not incorporated in any glass, presumably due to volatilization during melting. Apparently, the details of incorporation of Cr, and to an extent V, are different depending on whether sulfur is added as in this study or in oversaturated conditions as described in the literature, where Cr partitions to a sulfate-containing salt phase.

Borosilicate glass, Waste Glass, vanadium chemistr↗
DOE OSTI2020

Pool Fire Analysis Methodology for Assessing Damage to Waste Containers

This document provides guidelines, whereby the physical parameters associated with postulated flammable/combustible liquid pool fires are determined (thermal analysis) in order to estimate the number of waste containers impacted. Pool fire damage to waste containers may occur due to a container being fully engulfed (within the periphery of the pool), partially engulfed, or within the critical heat flux distance. The methodology described here for estimating the size of the pool fire is developed using spill characteristics outlined in Chapter 65 of the SFPE Handbook (Ref. 7) and testing done by Hansen (Ref. 25). For exposures, the critical flux height is determined using the detailed method of Shokri and Beyler (Ref. 8) from the flame height, calculated as described by Heskestad (Ref. 7) and the average black body emissive power, determined using the method of Shokri and Beyler (Ref. 8). Three types of liquid pool fires should be considered; unconfined crash/instantaneous spill fires, unconfined continuously flowing (also called metered leak) spill fires, and confined spill fires. All fuel sources should be considered consistent with the facility Fire Hazard Analysis (FHA). However, the most likely liquid fuel sources present in TRU waste facilities are those present on TRU waste handling vehicles (forklifts, pallet movers, transport trucks, etc.). This method can also be used for assessing damage to items other than waste containers or other methods of analysis may be used with appropriate justification.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Surveillance Report on SAVY 4000 and Hagan Nuclear Material Storage Containers Update for FY2020

A Surveillance Program is in place to assess how nuclear material storage containers at LANL are aging in-service. This program is guided by the LANL Surveillance Plan which is required by DOE M441.1-1. The plan is modified as necessary to ensure that any issues identified during surveillance or laboratory studies are examined in future surveillances, and that any lifetime implications are taken into account. Under the plan, overall container integrity is evaluated by a combination of visual inspections and photographs. Multiple measurements are made to assess the SAVY 4000 performance, including, helium leakage rate, O-ring hardness (durometer), O-ring compression set, filter performance and filter water resistance. The containers for surveillance are chosen annually based on the previous surveillance program results and observations. The surveillance plan targets items believed to provide the greatest challenge to the SAVY 4000 container integrity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers” for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field’s Metal eutectic alloy with a melting point of 62 °C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Fire-Induced Pressure Response and Failure Characterization of PCV/SCV/3013 Containers - Phase 3

Several Department of Energy (DOE) facilities have materials stored in robust, welded, stainless - steel containers with presumed fire - induced pressure response behaviors. Lack of test data related to fire exposure requires conservative safety analysis assumptions for container response at these facilities. This conservatism can in turn result in the implementation of challenging operational restrictions with costly nuclear safety controls. To help address this issue for sites that store DOE 3013 stainless - steel containers, a series of ten tests were undertaken at Sandia National Laboratories. The goal of this test series was to obtain the response behavior for various configurations of DOE 3013 containers with various payload compositions when exposed to one of two ASTM fire conditions. Key parameters measured in the test series included identification of failure - specific characteristics such as pressure, temperature, and whether or not a vessel was breached during a test . Numerous failure - specific characteristics were identified from the ten tests. This report describes the implementation and execution of the test series performed to identify these failure - specific characteristics. Discussions on the test configurations, payload compositions, thermal insults, and experimental setups are presented. Test results in terms of pressurization and vessel breach (or no - breach) are presented along with corresponding discussions for each test.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Pit depth analysis using automated software for three containers with complete imaging

The 3013 surveillance program is tasked with ensuring the long term safety of plutonium storage in 3013-compliant containers. One component of this task is to understand corrosion features found on the interior sidewall of the inner container closure weld region (ICCWR). To this end, three containers were selected to have laser confocal microscope (LCM) 20X imaging of the entire ICCWR. This report describes the analysis of the LCM pit depth data using the automated corrosion analysis software. The analysis included determining maximum pit depths and distributions of pit depths, volumes, and diameters for all pit features identified on the three containers. Results provided insight into possible corrosion mechanisms and were consistent with previous pit growth model predictions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Finite Element Simulation of the Acoustic Pressure Inside a Beverage Container for Non-Thermal, Ultrasound-based Pasteurization

The purpose of this effort is to investigate whether large acoustic pressure waves can be transmitted inside beverage containers to enable pasteurization. Acoustic waves are known to induce large nonlinear compressive forces and shock waves in fluids, suggesting that compression waves may be capable of damaging bacteria inside beverage containers without appreciably increasingly the temperature or altering the freshness and flavor of the beverage contents. Although a combined process such as thermosonication (e.g., sonication with heating) is likely more efficient, it is instructive to compute the acoustic pressure field distribution inside the beverage container. The COMSOL simulations used two and three-dimensional models of beverage containers placed in a water bath to compute the acoustic pressure field. A limitation of these COMSOL models is that they cannot determine the bacterial lysis efficiency, rather the models provide an indirect metric of bacterial lysis based on the magnitude of the pressure field and its distribution.

42 ENGINEERING↗
DOE OSTI2023

Packaging Capacity Calculation: Pu Oxide Packaging Options with the 2 Quart SAVY Container

This calculation supports programmatic efforts to Dilute and Dispose Pu. Metal is to be oxidized at LANL, then transported to SRS for dilution at SRS. Ultimately, the diluted oxide will be disposed of at the Waste Isolation Pilot Plant (WIPP). Currently 3013 containers are used to package the oxide, which are placed in to 9975 Type B shipping containers. The Oxide Packaging Technology Maturation Plan determined that a transition to the used of 2-qt SAVY containers placed in 9977 would increase efficiency in shipping and handling and increase programmatic flexibility. The 2-quart SAVY container meets DOE M 441.1 and TA-55 Documented Safety Analysis requirements for handling and storage of Pu oxide at LANL.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Improved Compaction Experiments and Modeling of Waste Isolation Pilot Plant Standard, Non-degraded, Waste Containers

A credible simulation of disposal room porosity at the Waste Isolation Pilot Plant (WIPP) requires a tenable compaction model for the 55-gallon waste containers within the room. A review of the legacy waste material model, however, revealed several out-of-date and untested assumptions that could affect the model’s compaction behavior. For example, the legacy model predicted non-physical tensile out-of-plane stresses under plane strain compression. (Plane strain compression is similar to waste compaction in the middle of a long drift.) Consequently, a suite of new compaction experiments were performed on containers filled with surrogate, non-degraded, waste. The new experiments involved uniaxial, triaxial, and hydrostatic compaction tests on quarter-scale and full-scale containers. Special effort was made to measure the volume strain during uniaxial and triaxial tests, so that the lateral strain could be inferred from the axial and volume strain. These experimental measurements were then used to calibrate a pressure dependent, viscoplastic, constitutive model for the homogenized compaction behavior of the waste containers. This new waste material model’s predictions agreed far better with the experimental measurements than the legacy model’s predictions, especially under triaxial and hydrostatic conditions. Under plane strain compression, the new model predicted reasonable compressive out-of-plane stresses, instead of tensile stresses. Moreover, the new model’s plane strain behavior was substantially weaker for the same strain, yet substantially stronger for the same porosity, than the legacy model’s behavior. Although room for improvement exists, the new model appears ready for prudent engineering use.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Engineered Solutions Group's SMR Containment Cable and Electrical Penetration Assembly System

GAIN Project CRADA Number NFE-21-08839, entitled “SMR Containment Cable and EPA System” was initiated by the partnership of Oak Ridge National Laboratory (ORNL) and Engineered Solutions Group (ESG) to test an ESG-designed Electrical Penetration Assembly and Containment Cabling System Qualified for not only legacy LWR designs, but also Small Modular Reactors designs currently being designed by several different suppliers. This project was undertaken to fill the equipment gap of EPA and Cabling Systems that require much more severe environmental requirements than legacy plant applications present due to their smaller containment volumes that result in high energy densities compared to legacy designs. This high energy density results in severe accident environments and more severe normal operating conditions as well. We developed two approaches to qualify Electrical Penetration Assembly (EPA) and Containment Cabling Systems for SMRs and Advanced Reactors. We take into consideration the more severe environmental parameters found with SMR designs. The system will need to meet a qualification test program addressing wear/cyclic aging, potential radiation exposure, thermal aging, vibration aging, thermal cycling, seismic qualification, electrical fault testing (per IEEE 317) and accident simulation. The equipment must meet the requirements of 10CFR50.49, GDC 50 in 10CFR50 Appendix A, and 10CFR50 Appendix J. NRC Regulatory Guides (RGs) identify an acceptable way of meeting regulatory requirements. RGs frequently endorse a standard for meeting these requirements. Specific to this review, equipment would be qualified in accordance with the following IEEE Standards. • IEEE 317-2013 (Electrical Penetration Assemblies), which is endorsed by RG 1.63, Rev. 3, • IEEE 323-2003 and the more current IEC/IEEE 60780-323 (Environmental Qualification of 1E Equipment) which is endorsed by RG 1.89. IEEE 323-2003 is endorsed by RG 1.209, • IEEE 344-2020 (Seismic Qualification. The 2013 version endorsed by RG 1.100, Rev. 4, with exceptions), • IEEE 383-2015 (Electrical Cables) (which is endorsed by RG 1.189 Rev. 4 and the -2003 version endorsed by RG 1.211 rev. 0), • IEEE 572-2019 (Electrical Connectors and Assemblies), which is endorsed by 1.156 Rev 1, and • [IEEE 1202 (endorsed by RG 1.189) would normally be applicable but the advanced cable designs are impervious to this cable flame test.] The primary goal of such a program is to provide an EPA design that can meet the qualification requirements for all legacy light water reactor plants currently operating as well as new plant designs including light water Small Modular Reactors. Thus, these requirements are applicable to plants licensed under 10CFR50 and 10CFR52. Other reactor designs may be evaluated, and this test system and qualification method applied to those applications if the requirements would satisfy the requirements of the intended plant. A secondary benefit of this work is to document some of the history and background in these requirements as there have been recent delays in an SMR licensing process due to NRC Requests for Additional Information in this subject matter area.

42 ENGINEERING↗
DOE OSTI2026

Improbability of Post-Closure Criticality in Array of Transuranic Waste Containers after Compaction by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, post-closure nuclear criticality is improbable when room closure compacts containers disposing transuranic (TRU) waste emplaced at the Waste Isolation Pilot Plant (WIPP), an operating repository in bedded salt in southeastern New Mexico. As described in the original WIPP certification, a qualitative estimate of the probability of post-closure criticality in TRU waste produced from atomic energy defense activities has been low either because remote-handled TRU waste canisters are neutronically isolated by the bedded salt or because the low fissile mass in an array of contact-handled TRU waste drums cannot be compacted sufficiently by room closure from salt creep. These situations are still valid for the majority of TRU waste that is disposed at WIPP without any disposal constraints, as updated herein. This report also qualitatively evaluates the probability of criticality after disposal of TRU waste in pipe overpack containers (POC) where every POC in a shipment may have the maximum 200 fissile gram equivalent of 239Pu content. The probability of criticality for a disposal room filled with POCs is estimated during four representative phases of repository evolution: (1) a large salt block falls onto POCs in the first 20 years, (2) salt creep closes a disposal room in the first 1000 years without brine seepage and subsequent gas generation, which permits maximum compaction, (3) some brine seepage occurs into the closed room, which initiates consumption of the fiberboard (cellulose) impact absorber in the POC in the second 1000 years, and (4) full brine inundation of a room and consumption of all fiberboard thereafter. Salt-block fall in the first phase does not greatly disrupt three tiers of POCs. The compacted spacing of POCs in the later three repository conditions is calculated through high-fidelity, geomechanical modeling. Criticality evaluation of compacted 200-g 239 Pu spheres at the compacted spacing shows that neither 12-inch nor 6-inch POCs are critical after the first 1000 years, the second 1000 years, or thereafter as the sea of reflector material changes to represent the three repository conditions. Specifically, fiberboard and iron isolates 239 Pu while dry, and brine reduces the reactivity when a room is partially and fully inundated. Because POC behavior bounds behavior of other standard TRU waste containers, post-closure criticality caused by room closure compacting containers is omitted in the performance assessments for the 2019 and 2026 WIPP compliance re-certification applications to the US Environmental Protection Agency.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Detecting the Presence of Intrusive Drilling in Secure Transport Containers Using Non-Contact Millimeter-Wave Radar

We employ a 77–81 GHz frequency-modulated continuous-wave (FMCW) millimeter-wave radar to sense anomalous vibrations during vehicle transport at highway speeds for the first time. Secure metallic containers can be breached during transport by means of drilling into their sidewalls but detecting a drilling signature is difficult because the large vibrations of transport drown out the small vibrations of drilling. For the first time, we demonstrate that it is possible to use a non-contact millimeter-wave radar sensor to detect this micron-scale intrusive drilling while highway-speed vehicle movement shakes the container. With the millimeter-wave radar monitoring the microdoppler signature of the container’s vibrating walls, we create a novel signal-processing pipeline consisting of range–angle tracking, time–frequency analysis, horizontal stripe image convolution, and principal component analysis to create a robust and powerful detection statistic to alarm if drilling is present. To support this pipeline, we develop a statistical model combining the vibrating container and the random vibrations induced by vehicle movement to explore the robustness of the sensor’s detection capabilities. The presented results strongly support the inclusion of a millimeter-wave radar vibration sensor into a transport security system.

Wagner, Samuel↗
DOE OSTI2020

Methods of forming aromatic containing compounds

Methods that include acylating an aromatic containing compound by reacting the aromatic containing compound with an anhydride containing compound to form an acylated aromatic containing compound.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Stretchable electronic systems with containment chambers

The present invention provides electronic systems, including device arrays, comprising functional device(s) and/or device component(s) at least partially enclosed via one or more containment chambers, such that the device(s) and/or device component(s) are at least partially, and optionally entirely, immersed in a containment fluid. Useful containment fluids for use in containment chambers of electronic devices of the invention include lubricants, electrolytes and/or electronically resistive fluids. In some embodiments, for example, electronic systems of the invention comprise one or more electronic devices and/or device components provided in free-standing and/or tethered configurations that decouple forces originating upon deformation, stretching or compression of a supporting substrate from the free standing or tethered device or device component.

Rogers, John A.↗