#
Westinghouse Electric Corporation
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The organization ** Westinghouse Electric Corporation** represents an institution, an association, or corporate body that is associated with resources found in **University of Oklahoma Libraries**.

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Westinghouse Electric Corporation
Resource Information

The organization

**Westinghouse Electric Corporation**represents an institution, an association, or corporate body that is associated with resources found in**University of Oklahoma Libraries**.- Label
- Westinghouse Electric Corporation

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- Optimization of distribution transformer efficiency characteristics
- George Westinghouse, 1846-1914
- Annual report
- Transformer requirements for the year 2000
- The book of record of the time capsule of cupaloy, : deemed capable of resisting the effects of time for five thousand years, preserving an account of universal achievements, embedded in the grounds of the New York World's Fair, 1939
- Annual report for 1978 on research, development and demonstration of nickel-iron batteries for electric vehicle propulsion
- Study of lead-acid battery systems for peaking power : final report for the period ending Oct. 1976
- Iron-air battery devdelopment program
- Westinghouse extension course
- Electrical transmission and distribution reference book,
- Sensor application survey : technical report, task 1
- Electrical transmission and distribution reference; : Supplement.

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- A pressure-deflection relation for a clamped circular plate in the plastic range
- A simple method for calculating the maximum size of ductile rupture in pressurized systems
- A simple method for calculating the maximum size of ductile rupture in pressurized systems
- A small perturbation approach to the study of parallel channel boiling flow oscillations
- A survey of numerical methods in the solution of diffusion problems
- A table of the integral [psi] (x, t)=[1 over 2 times the square root of pi t times the integral from minus infinity to infinity of (e to the minus (x-y)2 over 4t divided by 1+y2) dy]
- ABRAC : an IBM-704 three dimensional nuclear-thermal depletion program with distributed void effects
- ART : a program for the treatment of reactor thermal transients on the IBM-704
- ART-04 : a modification of the ART program for the treatment of reactor thermal transients on the IBM-704
- ATBAC, an IBM-704 code for reactor thermal transients
- Absolute power calibration of a flexible survey assembly
- An alternating direction method for operators with an application to the transport equation
- An application of Specer's method to the Stieltjes and Hamburger moment problems
- An appraisal of the Mim and art differencing methods employing M0076, a digital frequency analysis program
- An experiment to measure effective delayed neutron fractions
- Analysis of check valve disc motion during a flow transient
- Analysis of the neutron capture cross section and resonance integral of hafnium
- Applications of the simplified spherical harmonics equations in spherical geometry
- Approximate spatially separable flux calculations
- BAFL-1 : a program for the solution of thin elastic plate equations on the PHILCO-2000 computer
- BAND-1 : a data reduction program for the IBM-704
- BKSB : a revision of the BKS system for the Philco-2000 computer
- BOUND 1, 2, 3 : IBM-704 programs for the analytics solution of the one energy P3 approximation in cylindrical geometry
- BRIC : an IBM-704 two-dimensional nuclear-thermal depletion program with distributed void effects
- Before the National war labor board : Westinghouse electric & manufacturing co. and United electrical, radio & machine workers of America, C.I.O., Case, no. 111-8213-D, General electric company and United electrical,radio & machine workers of America, C.I.O., Case no. 111-8214-D
- Behavior of highly restrained partial penetration welds in austentic stainless steel
- Benzene turbine cycles
- Bettis FORTRAN programming : auxiliary subroutines
- Bettis technical review
- Bibliography on two-phase heat transfer
- Block iterative methods for two-cyclic matrix equations with special applications to the numerical solution of the second-order self-adjoint elliptic partial differential equation in two dimensions
- Brine migration test for Asse Mine, Federal Republic of Germany : final test plan : technical report
- Brine migration test for Asse Mine, Federal Republic of Germany : final test plan : technical report
- CHIC programs for thermal transients
- CINDER : a one-point depletion and fission product program
- CLIP 1 : an IBM-704 program to solve the P-3 equations in cylindrical geometry
- CTS-3 : a multi-group transport program for infinite cylinders
- CURF 1 : a least-squares polynomial fitting program for the Philco-2000 computer
- Calculated equilibrium constants for metal-oxygen reactions in molten uranium alloys
- Calculation of fission product activity in PWR from a seed plate failure
- Calculation of the background neutron source in new, uranium-fueled reactors
- Calculation of the concentrations and fissions of heavy isotopes in time-varying irradiations of uranium and plutonium
- Calculation of thermal neutron fluxes in primary shields
- Cleaning zircaloy components for pressure bonding
- Cofit, a least squares cosine fitting program for the IBM-704
- Conference on Bearing Development for Water Lubricated Application : summary outline of talks
- Conversion of Rapid City pilot plant : phase II, preliminary detailed design
- Corrosion testing of Zircaloy-2 and Zircaloy-3
- Crime prevention through environmental design : the commercial demonstration in Portland, Oregon : executive summary
- Crime prevention through environmental design : the school demonstration in Broward County, Florida : executive summary
- Critical experiments in a uranium-zirconium water-moderated core with plate fuel elements and slab geometry
- Critical experiments with a seed and blanket slab geometry assembly
- DART : a digital-analog system for computing reactor start-up transients
- DRACO : a three-dimensional few-group depletion code for the IBM-704
- Deflections and load distributions in linear elastic structures : CTAC and MODE codes
- Description of Shippingport Atomic Power Station site and surrounding area with radiation background and meteorological data
- Description of the Shippingport Atomic Power Station
- Design and evaluation of fluidized bed heat recovery for diesel engine systems
- Design and evaluation of fluidized bed heat recovery for diesel engine systems
- Development of Shippingport Atomic Power Station operating procedures
- Differential thermal analysis of some irradiated materials
- Dimensional inspection of irradiated PWR Core 1 blanket fuel
- ECESS : an IBM-704 program computing transport equation coefficients for a monatomic gas moderator in the thermal energy region
- EURIPUS-3 and DAEDALUS : Monte Carlo density codes for the IBM-704
- Effect of hydrogen and hydride morphology on the tensile properties of zircaloy-2
- Effect of neutron bombardment upon the properties of ASTM type SA212B steel
- Effect of surface treatment on the corrosion resistance of zircaloy-2
- Effects of silicon, nitrogen, and oxygen on the corrosion and hydrogen absorption performance of zircaloy-2
- Ensign code
- Equations and programs for solutions of the neutron group diffusion equations by synthesis approximations
- Evaluation and demonstration of the capillary suction sludge dewatering device
- Evaluation of crud deposits of PWR blanket bundles
- Evaluation of isotopic content of irradiated natural uranium dioxide fuel rods PWR Core 1
- Evaluation of silicone fluid for replacement of PCB coolants in railway industry
- Evaluation of the PWR Core 1 failed element detection and location system
- Examination of PWR Core 1 blanket fuel rods for microstructure changes, hydrogen pickup, burst strength, and fission gas release at the end of the second seed life
- Examination of PWR Core 1 blanket fuel rods for microstructure, hydrogen pickup, and burst strength
- Exponential circuit response
- Extrapolation techniques for real symmetric matrices
- Extrusion of zircaloy 1 by Ugine-Sejournet glass process
- F0010 : a two-dimensional IBM-704 code for the calculation of thermal stresses in a long, hollow cylinder with internal heat generation
- F0020 : an IBM-704 thermal transient analysis code
- FLIP : an IBM-704 code to solve the PL and double-PL equations in slab geometry
- Fabrication and characteristics of experimental radiographic amplifier screens
- Fabrication defects observed in PWR core 1 blanket fuel rods
- Fabrication development of boron carbide and boron carbide+silicon carbide mixtures for possible application as lumped burnable poisons in PWR-2
- Far field monitor for instrument landing systems
- Few group constants for delayed neutrons
- Final report : economic assessment photovoltaic/battery systems
- Fission fragment damage to crystal structures
- Fission gas release in PWR Core 1 blanket fuel rods upon conclusion of seed 1 life
- Flux from homogeneous cylinders containing uniform source distributions
- Forced-convection heat transfer burnout studies for water in rectangular channels and round tubes at pressures above 500 PSIA
- Fort Hood solar total energy project
- Fortran unit record simulation techniques (FURST)
- Further work on the diffusion of krypton-85 from uranium dioxide powder
- Gas cooled fuel cell systems technology development : final report for the first logical unit of work, contract period, May 1982-May 1983
- Gas pressure bonding of production size PWR core 2 plate type fuel elements containing ceramic fuel
- Graphs of coefficients for determining multigroup P-1 constants from differential scattering cross sections
- HEAT : a one-dimensional heat transfer equation code for the IBM-704
- HERD 1, 2, and 3 : IBM-704 codes used to solve the one-dimensional, one-velocity transport equation with isotropic scattering
- HYDRA-2 : an IBM-704 code calculating daughter isotope fission product activity in coolant
- Hazards to the area surrounding PWR due to atomospheric diffusion of radioactivity
- High-temperature water and steam-corrosion behavior of zirconium-uranium alloys
- Hydrogen absorption in zircaloy during aqueous corrosion : effect of environment
- Hydrogen flammability data and application to PWR loss-of-coolant accident
- Hydrogen redistribution in thin plates of zirconium under large thermal gradients
- Hydrogen redistribution in zircaloy-2 under thermal and mechanical stress gradients
- In-pile effective thermal conductivity of oxide fuel elements to high fission depletions
- Inelastic deflection and buckling of fuel plates due to fluid velocity and compressive edge loads
- Input preparation for diffusion-depletion programs on the Philco-2000 computer
- Interim report : the strengthening effect of beryllium on zircaloy-3
- Interim report of creep behavior of Zircaloy-2 and -3
- Internal zirconium hydride formation in zircaloy fuel element cladding under irradiation
- Investigation of burnout heat flux in rectangular channels at 2000 psia
- Iron, carbon, and nitrogen impurities in PWR-2 seed fuel
- Irradiation behavior of bulk B4C and B4C-SiC burnable poison plates
- Isotopic analyses of irradiated natural uranium dioxide fuel rods from PWR Core 1 : preliminary results
- Isotopic concentrations for a time-varying irradiation of uranium or plutonium
- KATE-1 : a program for calculating Wigner-Wilkins and Maxwellian averaged thermal constants on the Philco-2000
- KOAD-1 : a digital program to calculate stresses and deflections in linear elastic structures under thermal distortion, pressure and applied loads
- LOGIC : a flux data processing code for the IBM-650
- Large counting losses in neutron detection channels
- M0150 : a FORTRAN program to solve the double P-3 equations in slab geometry
- M0176 : a FORTRAN program to solve several P-approximations to the few group neutron transport equation in slab geometry
- M0322 and M0332 : FORTRAN programs for calculating neutron absorption in spheres distributed randomly
- M0564 : a FORTRAN program for symmetirc and asymmetric resonance integrals
- MARC : a multigroup Monte Carlo program for the calculation of capture probabilities
- MUFT-4 : a fast neutron spectrum code for the IBM-704
- MUFT-5, a fast neutron spectrum program for the Philco-2000
- MULE : a FORTRAN program for the calculation of three types of overtone modes
- Measurement of change in [capture to fission ratio] and fission rate with temperature for U-235 : (LWB-LSBR development program)
- Measurement of the natural neutron source in two cores
- Metallurgical examination of PWR Core 1 blanket fuel rods at the end of the third seed life
- Metallurgical examination of fuel rod from PWR-1 blanket at the end of the second seed life
- Multigroup Fourier transform calculation description of MUFT-III code
- Neutron bursts in a ring reactor
- Neutron degradation in an absorbing hydrogenous moderator
- Neutron flux mapping by large detectors
- Nuclear analysis of thermal reflected cylindrical homogeneous critical assemblies
- Nuclear reactor depletion programs for the Philco-2000 computer
- Numerical methods and techniques used in the two-dimensional neutron-diffusion program PDQ-5
- Numerical solution of the two-group diffusion equations in x-y geometry
- On estimating rates of convergence in multigroup diffusion problems
- On the equilvalence of the spherical harmonics method and the discrete ordinate method using Gauss quadrature for the Boltzman equation
- Operational consequences of literacy gap
- Over-all evaluation of blanket fuel removed from PWR Core 1 during the first refueling of the seed
- P-3 multigroup calculations of neutron attenuation in metal-hydrogenous shields
- P1MG : a one-dimensional multigroup P1 code for the IBM-704
- P1MG : a one-dimensional multigroup P1 code for the IBM-704
- P3MG1 : a one-dimensional multigroup P-3 program for the Philco-2000 computer
- PDQ : an IBM-704 code to solve the two-dimensional few-group neutron-diffusion equations
- PDQ-3 : a program for the solution of the neutron-diffusion equations in two dimensions on the IBM-704
- PDQ-4 : a program for the solution of the neutron-diffusion equations in two dimensions on the Philco-2000 computer
- PDQ-5: A FORTRAN program for the solution of the two-dimensional neutron-diffusion problem, Part 1, Steady-state version
- PROP and JET : a program for the synthesis and survey of three-dimensional power shapes on the IBM-704
- PWR FLECHT-Set phase B1 evaluation report
- PWR fuel element specifications
- PWR hazards summary report
- PWR hazards summary supplement
- PWR loss-of-coolant accident-core meltdown calculations
- PWR plant container sizing criteria : studies of transient temperature and pressure in plant container following primary coolant system rupture
- PWR reactivity accidents
- Pickling of the zircaloys prior to corrosion exposure
- Plastic-sass : a computer program for stresses and deflections in a reactor subassembly under thermal, hydraulic, and fuel expansion loads
- Polyphemus : a Monte Carlo study of neutron penetrations through finite water slabs
- Polyphemus : a Monte Carlo study of neutron penetrations through finite water slabs, addendum
- Post-irradiation evaluation of a plate-type UO2 fuel element
- Preliminary report on status and evaluation of carbon steel
- Pressure vessel and piping codes applicable to the PWR reactor plant
- Pressure, thermal, and vibrational loads and deflections in linear elastic structures : FORTRAN II programs CATAC and more
- Primary coolant system activities due to a burst source of fission products
- Procedure for generation of boron trifluoride gas
- Properties of a hafnium control rod after exposure during three seed lives in PWR Core 1
- Properties of a hafnium control rod after exposure during two seed lives in pwr core 1
- RAM-1 : a FORTRAN program to transform and average cross sections from the ROC library tape for use in multigroup neutron transport programs
- RANCH : an IBM-704 program used to solve the one-dimensional, single energy neutron transport equation with anisotropic scattering
- RECAP-1 : a Monte Carlo program for estimating epithermal capture rates in slabs
- RECAP-2 : a Monte Carlo program for estimating epithermal capture rates in rod arrays
- RECAP-3 : a Monte Carlo program for estimating epithermal capture rates in rectangular or 60Â° parallelogram geometry
- ROC-1 : a FORTRAN program for storage of evaluated nuclear data
- Radiation damage exposure and embrittlement of reactor pressure vessels
- Radiation-induced reduction of chromium (VI) solutions
- Radiochemistry of fifth PWR fuel material test (X-1-f) X-1 loop NRX reactor
- Radiochemistry of the PWR fuel material cycling tests (WAPD-29-1 and -2) in the WAPD-29 VH-3 loop of materials testing reactor
- Radiochemistry of third PWR fuel material test X-1 loop NRX reactor
- Radiographic amplifier screens : fabrication process and characteristics
- Recoil range of fission fragments in zirconium
- Redundancy techniques for computing systems.
- Reference manual for the Bettis open shop system (BOSS-2) as used with the Philco-2000
- Resume of uranium alloy data, XI
- Resume of uranium alloy data, XII
- Resume of uranium dioxide oxide data, IX
- Resume of uranium dioxide oxide data, V
- Resume of uranium dioxide oxide data, VI
- Resume of uranium dioxide oxide data, VII
- ResumÃ© of uranium dioxide oxide data, IV
- Review of carbon steel corrosion data in high-temperature, high-purity water in dynamic systems
- SHLOG : a data retrieval program for the Philco-2000
- SLOP-1 : a thermal multigroup program for the IBM-704
- SN5001 : an IBM-650 code for steady-state thermal evaluation of an instrumented multi-fuel-plate subassembly
- SO131 : an IBM-650 code to solve pressure and thermal stress problems in core subassembly plates
- SPAN-2 : an IBM-704 code to calculate uncollided flux outside a circular cylinder
- SPIC-1 : an IBM-704 code to calculate the neutron distribution outside a right-circular cylindrical source
- STDY-3 : a program for the thermal analysis of a pressurized water nuclear reactor during steady-state operation
- Safeguards aspects of PWR reactor coolant chemistry
- Salt bath heat treatment of PWR core 2 fuel elements
- Saxton Plutonium Program : nuclear design of the Saxton partial plutonium core
- Saxton Plutonium Program semiannual progress report for the period ending June 30, 1967
- Saxton Plutonium Program semiannual progress report for the period ending June 30, 1967
- Saxton plutonium program semiannual progress report for the period ending June 30, 1966
- Saxton plutonium program semiannual progress report for the period ending June 30, 1966
- Seal-shell : a digital program to determine stresses and deflections in an axisymmetric shell of revolution
- Seal-shell-2 : a computer program for the stress analysis of a thick shell of revolution with axisymmetric pressures, temperatures, and distributed loads
- Selection and application of materials for the PWR reactor plant
- Shippingport Atomic Power Station : operating experience, developments, and future plans
- Shippingport Atomic Power Station inspection and test program
- Shippingport Atomic Power Station operating experience, developments, and future plans : prepared for the U.S. - Japan Atomic Industrial Forum, Tokyo, Japan, December 5-8, 1961
- Shock testing instrumentation
- Simplified reactor theory lecture series , Lecture I, Bulk conservation of neutrons
- Simplified reactor theory lecture series , Lecture II, Absorption, leakage and moderation
- Simplified reactor theory lecture series , Lecture IV, Stability of reactors - thermal
- Some information on the metallographic appearance of hafnium
- Space- and time-dependent flux oscillations (and instability) in thermal reactors due to nonuniform formation of depletion of xenon
- Span-3 : a shield design program for the Philco-2000 computer
- Steam jet pump analysis and experiments
- Stochastic fluctuations in a power reactor
- Stress analysis with applications to pressurized water reactors
- Stresses and deflections in thick, curved plates
- Structure design notes : calculation of stresses, forces, and deflections in linear-elastic skeleton structures with temperature, pressures, applied loads and redundant loads
- Study-by simulator techniques-of transient pressures in high pressure water systems utilizing a surge tank
- Summary report of reactor hazards associated with the high temperature test facility
- Summary report on the hazards associated with the pressurized water reactor : flexible assembly
- Surface modified nuclear optical model : description of the SUMNUM code for the NORC computer
- TEMP-2 : a one dimensional transient thermal stress program for the IBM 704
- TITE : a digital program for the prediction of two-dimensional two-phase hydrodynamics
- TKO : a three-dimensional neutron diffusion equation program for the IBM-704
- TRAC-1 : a Monte Carlo Philco-2000 program for the calculation of neutron capture probabilities
- TRIP-1 : a two-dimensional P-3 program in X-Y geometry for the IBM-704
- TURBO : a two-dimensional few-group depletion code for the IBM-704
- TUT-T5 : a two-dimensional Monte Carlo calculation of capture probabilities
- Technical bases for establishing a salt test facility
- Technical information, preliminary hazards summary report, Volume II, License application
- Temperature coefficients of a highly enriched core in slab geometry
- The BKS system for the Philco-2000 computer
- The IBM-704 SIMPL codes
- The ORNL GCR-3, a 750-Mw(e) gas-cooled clad-fuel reactor power plant : a joint design study
- The Peaceman-Rachford method for small mesh increments
- The Shippingport Pressurized Water Reactor Project catalog of document abstracts
- The Shippingport Pressurized Water Reactor Project catalog of document abstracts
- The accommodation coefficients of helium and krypton on zircaloy-2
- The behavior of electrolytic solutions at elevated temperatures as derived from conductance measurements
- The calculation of thermal constants averaged over a Wigner-Wilkins flux spectrum : description of the SOFOCATE code (704 obtaining constants at thermal energies)
- The corrosion of zircaloy base fuel alloys in high temperature water
- The design of a continuously operated 1-keV deuterium-ion extractor
- The diffusion of krypton-85 from uranium dioxide powder
- The effect of buckling on the multigroup diffusion theory group constants calculated by the MUFT code
- The effect of lithium hydroxide and related solutions on the corrosion rate of zircalloy in 680 F water
- The effects of heat treatments on the hardness and tensile properties of cold rolled iodide hafnium
- The effects of stress on nuclear power plant operational decision making and training approaches to reduce stress effects
- The electrical conductivity and thermoelectric power of uranium dioxide
- Uncollided flux from finite circular and rectangular cylinders
- WANDA, a one-dimensional few-group diffusion equation code for the IBM-704
- WANDA-5 : a one dimensional neutron diffusion equation program for the Philco-2000 computer
- Weight-shape conversion tables for Zircaloy-2
- Welding procedure specification for welding of austenitic stainless steel to carbon steel by shielded metal-arc process
- XACT : a buffered tape read-write subroutine for the Philco-2000
- XITE : a digital program for the analysis of two-dimensional boiling flow transients with fluid expansion
- ZIP-2 : a one-dimensional few-group synthesis nuclear reactor depletion program for the Philco-2000 computer
- Zirconium highlights
- Zirconium highlights
- Zirconium, zircaloy, and hafnium safe practice guide for shipping, storing, handling, processing, and scrap disposal
- Zirconium-water reaction data and application to PWR loss-of-coolant accident
- 54 group library for P-1 programs
- A 50 kW on site concentrating solar photovoltaic power system
- A FORTRAN code for gamma penetration in a finite slab
- A FORTRAN computer program for data reduction of photographs of two-phase flow phenomena
- A FORTRAN program to solve the P-3 gamma ray equations in slab geometry
- A code to invert the fourier cosine transform - FTI-1
- A code to invert the fourier sine or cosine transform - FTI-2
- A digital program to determine the operational characteristics of a pilot-actuated water-relief valve : Bettis Code M0317
- A dimensional analysis of the departure from nucleate boiling heat flux in forced convection
- A few group one-dimensional code for IBM-650
- A lattice of slightly enriched UO2 fuel rods partly immersed in light water
- A metallographic and x-ray study of the UO2-ZrO2 system
- A numerical solution for plane elasticity problems
- A numerical solution for plate bending problems
- Two-region experiments in slightly enriched water-moderated uranium and uranium dioxide lattices
- Two-region reactivity worth method for analysis of fuel-poison subassemblies
- WANDA, a one-dimensional few-group diffusion equation code for the IBM-704
- The emanation of radon-220 from sintered UO2 powders and plates
- The evolution of xenon-133 from slightly irradiated zirconia-urania plates
- The performance of base-form ion exchangers for pH control and removal of radioisotopes from a pressurized water reactor system
- The physics and mathematical analysis for the TUT-T5 Monte Carlo Code
- The properties of hafnium from a PWR Core 1 control rod after one seed life exposure
- The pyrolytic carbon coating of ceramic fuel for the second Shippingport reactor core
- The release of helium from slightly irradiated boron carbide and boron carbide-silicon carbide plates
- The sampling estimate of the parameter variance/mean in reactor fluctuation measurements
- The set codes : IBM 704 codes for the calculation of the stresses in a pressure vessel with an ellipsoidal head
- The transformation kinetics of uranium-zirconium alloys containing 50 and 60 WT PCT uranium
- The two-dimensional quadruple P0 and P1 approximations
- The use of isomeric states for neutron threshold detection
- The use of microhardness in the determination of the diffusivity of oxygen in alpha zirconium
- The zircaloy-2 in-pile tube for the NRX Central Thimble
- Theory of local atomic displacements in solid solutions of uranium alloys
- Thermal stability evaluation of elastomeric seal materials
- Thermodynamic investigation of the benzene turbine cycle
- Time-dependent fission-product thermal and resonance absorption cross sections
- Time-dependent fission-product thermal and resonance absorption cross sections : (data revisions and calculational extensions)
- Transformations in uranium-base alloys : summary report, December 14, 1955 - March 31, 1957
- Transport delay simulation circuits
- Two region studies in slightly enriched water moderated uranium and uranimum dioxide lattices

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