Sandia National Laboratories
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The organization Sandia National Laboratories represents an institution, an association, or corporate body that is associated with resources found in Indiana State Library.
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Sandia National Laboratories
Resource Information
The organization Sandia National Laboratories represents an institution, an association, or corporate body that is associated with resources found in Indiana State Library.
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- Sandia National Laboratories
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291 Items by the Organization Sandia National Laboratories
5 Items that are about the Organization Sandia National Laboratories
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- A Review of techniques for propagating data and parameter uncertainties in high-level radioactive waste repository performance assessment models
- A Review of the South Texas Project probabilistic safety analysis for accident frequency estimates and containment binning
- A bibliography of marine radiation ecology prepared for the seabed program
- A cause-defense approach to the understanding and analysis of common cause failures
- A combined analytical study to characterize uranium soil and sediment contamination : the case of the Naturita UMTRA site and the role of grain coatings
- A comparison of parameter estimation and sensitivity analysis techniques and their impact on the uncertainty in ground water flow model predictions
- A comparison of software which solves systems of nonlinear equations
- A demonstration experiment of steam-driven, high-pressure melt ejection : the HIPS-10S test
- A finite element analysis of a reactor pressure vessel during a severe accident
- A performance assessment methodology for high-level radioactive waste disposal in unsaturated, fractured tuff
- A performance assessment methodology for low-level waste facilities
- A precision programmable step generator for use in automated test systems
- A self-teaching curriculum for the NRC/SNL low-level waste performance assessment methodology
- A simplified model of aerosol removal by containment sprays
- A simplified model of aerosol removal by natural processes in reactor containments
- A simplified model of aerosol scrubbing by a water pool overlying core debris interacting with concrete : draft report for comment
- A simplified model of aerosol scrubbing by a water pool overlying core debris interacting with concrete : final report
- A standard problem for HECTR-MAAP comparison : in-cavity oxidation
- A standard problem for HECTR-MAAP comparison : incomplete burning
- A summary of the fire testing program at the German HDR test facility
- A thermodynamic model of fuel disruption in ST-1
- A user's manual for the postprocessing program PSTEVNT
- A woman's place is where she wants to be : a photographic history of women at Sandia
- Aerosol Sampling and Transport Efficiency Calculation (ASTEC) and application to surtsey/DCH aerosol sampling system : code version 1.0 : code description and user's manual
- Aging and loss-of-coolant accident (LOCA) testing of electrical connections
- Aging management and performance of stainless steel bellows in nuclear power plants
- Aging of cables, connections, and electrical penetration assemblies used in nuclear power plants
- Aging, condition monitoring, and loss-of-coolant accident (LOCA) tests of class 1E electrical cables
- Aging, loss-of-coolant accident (LOCA), and high potential testing of damaged cables
- Ampacity derating and cable functionality for raceway fire barriers
- An analysis of operational experience during low power and shutdown and a plan for addressing human reliability assessment issues
- An analytical study of seismic threat to containmnent integrity
- An approach for estimating the frequencies of various containment failure modes and bypass events
- An experimental investigation of internally ignited fires in nuclear power plant control cabinets
- An investigation of the effects of thermal aging on the fire damageability of electric cables
- Analysis of bellows expansion joints in the Sequoyah containment
- Analysis of core damage frequency from internal events, Vol. 2, Expert judgment elicitation
- Analysis of core damage frequency, Vol. 1, Internal events methodolgy
- Analysis of core damage frequency, Vol. 3, Surry Power Station, Unit 1
- Analysis of core damage frequency, Vol. 4, Peach Bottom, Unit 2
- Analysis of core damage frequency, Vol. 5, Sequoyah, Unit 1
- Analysis of core damage frequency, Vol. 6, Grand Gulf, Unit 1
- Analysis of core damage frequency, Vol. 6, Grand Gulf, unit 1
- Analysis of diffusion flame tests
- Analysis of shell-rupture failure due to hypothetical elevated-temperature pressurization of the Sequoyah Unit 1 steel containment building
- Analysis of the LaSalle unit 2 nuclear power plant : risk methods integration and evaluation program (RMIEP)
- Assessing compliance with the EPA high-level waste standard : an overview
- Assessment of emergency response planning and implementation for large scale evacuations
- Assessment of the DCH issue for plants with ice condenser containments
- Assessment of the combustion model in the HECTR code
- Assessment of the impact of degraded shear wall stiffnesses on seismic plant risk and seismic design loads
- BORE : a computer program for the calculation of wellbore heat loss
- Background information for the development of a low-level waste performance assessment methodology
- Background information for the development of a low-level waste performance assessment methodology
- Bending stress and frequency calculations for the hearing-pack shaft
- Beta and gamma dose calculations for PWR and BWR containments
- Building the bombs : a history of the nuclear weapons complex
- Building the bombs : a history of the nuclear weapons complex
- Building the bombs : a history of the nuclear weapons complex
- CHARM, a model for aerosol behavior in time varying thermal-hydraulic conditions
- CORCON-MOD3, an integrated computer model for analysis of molten core-concrete interactions : user's manual
- Cable insulation resistance measurements made during cable fire tests
- Calculation of failure importance measures for basic events and plant systems in nuclear power plants
- Capacity of steel and concrete containment vessels with corrosion damage
- Characterization of retardation mechanisms in soil
- Circuit analysis : failure mode and likelihood analysis
- Circuit bridging of components by smoke
- Cobalt-60 simulation of LOCA radiation effects
- Code manual for CONTAIN 2.0 : a computer code for nuclear reactor containment analysis
- Code manual for MACCS2
- Comparison of strongly heat-driven flow codes for unsaturated media
- Components of an overall performance assessment methodology
- Conceptualization of a hypothetical high-level nuclear waste repository site in unsaturated, fractured tuff
- Containment event analysis for postulated severe accidents
- Containment integrity research at Sandia National Laboratories
- Core-concrete interactions using molten steel with zirconium on a basaltic basemat : the SURC-4 experiment
- Core-concrete interactions using molten urania with zirconium on a limestone concrete basemat : the SURC-1 experiment
- Core-concrete interactions with overlying water pools : the WETCOR-1 test
- Correlation of electrical reactor cable failure with materials degradation
- DCM3D, a dual-continuum, three-dimensional, ground-water flow code for unsaturated, fractured, porous media
- DOSFAC2 user's guide
- Damaged fuel experiment DF-1 : results and analyses
- Demonstrating the feasibility and reliability of operator manual actions in response to fire : final report
- Demonstration of a performance assessment methodology for high-level radioactive waste disposal in basalt formations
- Design, instrumentation, and testing of a steel containment vessel model
- Development of evacuation time estimate studies for nuclear power plants
- Direct containment heating experiments at low reactor coolant system pressure in the Surtsey test facility
- EM international
- EPRI/NRC-RES fire PRA methodology for nuclear power facilities : final report
- Economic risk of contamination cleanup costs resulting from large nonreactor nuclear material licensee operations
- Effects of adsorption constant uncertainty on contaminant plume migration : one- and two-dimensional numerical studies
- Effects of smoke on functional circuits
- Elicitation and use of expert judgement in performance assessment for high-level radioactive waste repositories
- End of a war, beginning of a laboratory : Z Division, 1945-1949
- Enhancements to data collection and reporting of single and multiple failure events
- Enhancements to the accident precursor methodology
- Entry/exit control components for physical protection systems
- Environmental testing of an experimental digital safety channel
- Environmental, health, and safety assessment of photovoltaics
- Equipment operability during station blackout events
- Equipment qualification (EQ) - risk scoping study
- Evaluation of a performance assessment methodology for low-level radioactive waste disposal facilities
- Evaluation of behavior and the radial shear strength of a reinforced concrete containment structure
- Evaluation of generic issue 57 : effects of fire protection system actuation on safety-related equipment
- Evaluation of materials of construction for the reinforced concrete reactor containment model
- Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, unit 1
- Evaluation of seals for mechanical penetrations of containment buildings
- Evaluation of severe accident risks and the potential for risk reduction
- Evaluation of severe accident risks, Vol. 1, Methodology for the containment, source term, consequence, and risk integration analyses
- Evaluation of severe accident risks, Vol. 2, Quantification of major input parameters
- Evaluation of severe accident risks, Vol. 2, Quantification of major input parameters
- Evaluation of severe accident risks, Vol. 3, Surry Unit 1
- Evaluation of severe accident risks, Vol. 4, Peach Bottom, Unit 2
- Evaluation of severe accident risks, Vol. 5, Sequoyah, Unit 1
- Evaluation of severe accident risks, Vol. 6, Grand Gulf, Unit 1
- Evaluation of severe accident risks, Vol. 6, Grand Gulf, unit 1
- Evaluation of suppression methods for electrical cable fires
- Experimental results from containment piping bellows subjected to severe accident conditions
- Experimental results from pressure testing a 1:6-scale nuclear power plant containment
- Experimental results pertaining to the performance of thermal igniters
- Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant
- Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the SURTSEY test facility
- Experiments to investigate the effect of flight path on direct containment heating (DCH) in the Surtsey test facility : the limited flight path (LFP) tests
- FLAME facility : the effect of obstacles and transverse venting on flame acceleration and transition to detonation for hydrogen-air mixtures at large scale
- Field-portable gas chromatograph : Electronic Sensor Technology model 4100
- Field-portable gas chromatograph : Perkin-Elmer Photovac Voyager
- Field-portable gas chromatograph : Sentex Systems, Inc. Scentograph Plus II
- Field-portable gas chromatograph/mass spectrometer : Inficon, Inc. HAPSITE
- Fire modeling of the Heiss Dampf Reaktor containment
- Fission product release and fuel behavior of irradiated light water reactor fuel under severe accident conditions : the ACRR ST-1 experiment
- HECTR analyses of the Nevada Test Site (NTS) premixed combustion experiments
- Handbook for battery energy storage in photovoltaic power systems
- Handbook of parameter estimation for probabilistic risk assessment
- Health effects models for nuclear power plant accident consequence analysis : low LET radiation
- Health effects models for nuclear power plant accident consequence analysis : low LET radiation
- High performance computing for U.S. industry
- Historical case analysis of uranium plume attenuation
- Human radiation exposures related to nuclear weapons industries
- Human reliability data bank for nuclear power plant operations
- Hydrogen-steam jet-flame facility and experiments
- Hydrogen:air:steam flammability limits and combustion characteristics in the FITS vessel
- IMPACTS-BRC, version 2.0 : program user's manual
- IMPACTS-BRC, version 2.1 : code and data verification
- Identification and analysis of factors affecting emergency evacuations
- Impact of distributed energy resources on the reliability of a critical telecommunications facility
- In-pile observation of fuel and clad relocation during LMFBR core-disruptive accidents
- Instrumentation and process control development for in situ coal gasification
- Integrated fuel-coolant interaction (IFCI 6.0) code : user's manual
- Integrated risk assessment for the LaSalle unit 2 nuclear power plant : phenomenology and risk uncertainty evaluation program (PRUEP)
- Interim reliability evaluation program : analysis of the Millstone Point Unit 1 Nuclear Power Plant
- Interior intrusion detection systems
- Large-scale molecular dynamics simulations of metal sorption onto the basal surfaces of clay minerals
- Leak and structural test of personnel airlock for LWR containments subjected to pressures and temperatures beyond design limits
- Least squares fitting of discrete data by polynominals in several variables
- Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables : U.S./French cooperative research program
- Loss of essential service water in LWRs (GI-153) : scoping study
- Low-cost photovoltaic cell mount study
- Lower head failure experiments and analyses
- MACCS version 1.5.11.1 : a maintenance release of the code
- MELCOR 1.8.0 : a computer code for nuclear reactor severe accident source term and risk assessment analyses
- MELCOR Accident Consequence Code System (MACCS)
- MELCOR analyses for accident progression issues
- MELCOR computer code manuals
- MELCOR computer code manuals
- MELPROG-PWR/MOD1, a two-dimensional, mechanistic code for analysis of reactor core melt progression and vessel attack under severe accident conditions
- Melt progression, oxidation, and natural convection in a severely damaged reactor core
- Metallographic preparation of titanium diboride coatings
- Methods for external event screening quantification : risk methods integration and evaluation program (RMIEP) methods development
- Microcomputer applications of, and modifications to, the modular fault trees
- Minimum bed parameters for in situ processing of oil shale
- Mitigation of direct containment heating and hydrogen combustion events in ice condenser plants : analyses with the CONTAIN code and NUREG-1150 PRA methodology
- Modeling one-dimensional radionuclide transport under time-varying fluid-flow conditions
- Nuclear power plant design concepts for sabotage protection
- Overpressurization test of a 1:4-scale prestressed concrete containment vessel model
- PARTITION, a program for defining the source term/consequence analysis interface in the NUREG-1150 probabilistic risk assessments : user's guide
- Parametric evaluation of seismic behavior of freestanding spent fuel dry cask storage systems
- Perspectives on reactor safety
- Perspectives on reactor safety
- Photoacoustic infrared monitor : Innova AirTech Instruments type 1312 multi-gas monitor
- Photovoltaic reliability R&D toward a solar-powered world
- Platinum catalytic igniters for lean hydrogen-air mixtures
- Posttest analysis of a 1:6-scale reinforced concrete reactor containment building
- Posttest analysis of the NUPEC/NRC 1:4 scale prestressed concrete containment vessel model
- Posttest analysis of the steel containment vessel model : /
- Posttest destructive examination of the steel liner in a 1:6-scale reactor containment model
- Preliminary study of the remote sensing of a modified in situ oil shale report by electromagnetic methods
- Preliminary thermal expansion screening data for tuffs
- Pressurized melt ejection into water pools
- Pretest analyses of the steel containment vessel model
- Pretest analysis of a 1:4-scale prestressed concrete containment vessel model
- Pretest round robin analysis of a prestressed concrete containment vessel model
- Prioritization of reactor control components susceptible to fire damage as a consequence of aging
- Probabilistic accident consequence uncertainty analysis : dispersion and deposition uncertainty assessment
- Probabilistic accident consequence uncertainty analysis : early health effects uncertainty assessment
- Probabilistic accident consequence uncertainty analysis : uncertainty assessment for deposited material and external doses
- Probabilistic accident consequence uncertainty analysis : uncertainty assessment for internal dosimetry
- Procedure for analysis of common-cause failures in probabilistic safety analysis
- Procedures for the external event core damage frequency analyses for NUREG-1150
- Proceedings of the CSNI Workshop on PSA Applications and Limitations : held in Santa Fe, New Mexico, September 4-6, 1990
- Proceedings of the Fourth Workshop on Containment Integrity : held in Arlington, Virginia, June 14-17, 1988
- Project Plowshare
- Putting weapons to the test
- Quality assurance (QA) plan for computer software supporting the U.S. Nuclear Regulatory Commission's high-level waste management program
- Quality assurance procedures for the CONTAIN severe reactor accident computer code
- RADTRAD : a simplified model for R̲A̲D̲ionuclide T̲ransport and R̲emoval A̲nd D̲ose estimation
- RADTRAD, a simplified model for R̲A̲D̲ionuclide T̲ransport and R̲emoval A̲nd D̲ose estimation
- Readiness Program
- Reexamination of spent fuel shipment risk estimates
- Reexamination of spent fuel shipment risk estimates
- Reference manual for the CONTAIN 1.1 code for containment severe accident analysis
- Relocation of metallic constituents in core debris beds
- Requirements for a transformerless power conditioning system
- Resolution of the direct containment heating issue for all Westinghouse plants with large dry containments or subatmospheric containments
- Resolution of the direct containment heating issue for combustion engineering plants and Babcock & Wilcox plants
- Results and insights on the impact of smoke on digital instrumentation and control
- Review and assessment of thermodynamic and transport properties for the CONTAIN code
- Review of NUREG-0654, supplement 3, "Criteria for protection action recommendations for severe accidents"
- Risk assessment for the intentional depressurization strategy in PWRs
- Risk comparisons of scheduling preventive maintenance for boiling water reactors during shutdown and power operations
- Risk evaluation for a B&W pressurized water reactor, effects of fire protection system actuation on safety-related equipment : evaluation of generic issue 57
- Risk evaluation for a General Electric BWR, effects of fire protection system actuation on safety-related equipment : evaluation of generic issue 57
- Risk evaluation for a Westinghouse PWR, effects of fire protection system actuation on safety-related equipment : evaluation of generic issue 57
- Risk impact of BWR technical specifications requirements during shutdown
- Risk methodology for geologic disposal of radioactive waste : scenario selection procedure
- Risk methods insights gained from fire incidents
- Risk-informed assessment of degraded containment vessels
- Round robin posttest analysis of a steel containment vessel model
- Round robin pretest analyses of a steel containment vessel model and contact structure assembly subject to static internal pressurization
- Round-Robin analysis of the behavior of a 1:6-scale reinforced concrete containment model pressurized to failure : posttest evaluations
- SAFE users manual prepared for Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, under memorandum of understanding DOE 40-550-75
- SECPOP2000 : sector population, land fraction, and economic estimation program
- SECPOP90 : sector population, land fraction, and economic estimation program
- SWIFT II self-teaching curriculum : illustrative problems for the Sandia waste-isolation flow and transport model for fractured media
- Sandia National Laboratories : further improvements needed to strengthen controls over the purchase card program
- Sandia technology
- Security system signal supervision
- Seismic analysis of a prestressed concrete containment vessel model
- Seismic analysis of a reinforced concrete containment vessel model
- Severe accident testing of electrical penetration assemblies
- Shutdown decay heat removal analysis : plant case studies and special issue : summary report
- Shutdown decay heat removal analysis of a Babcock and Wilcox pressurized water reactor : case study
- Shutdown decay heat removal analysis of a General Electric BWR4/Mark I : case study
- Shutdown decay heat removal analysis of a combustion engineering 2-loop pressurized water reactor : case study
- Solar radiation energy resource atlas of the United States
- Solar thermal repowering
- Source term attenuation by water in the Mark I boiling water reactor drywell
- Spent nuclear fuel transportation package performance study issues report
- Spent nuclear fuel transportation package performance study issues report
- Status report : correlation of electrical reactor cable failure with materials degradation
- Steam explosion experiments with single drops of iron oxide melted with a CO2 laser
- Stored energy of gamma-irradiated WIPP salt
- Structural seismic fragility analysis of the Surry containment
- Structural seismic fragility analysis of the Zion containment
- Submergence and high temperature steam testing of class 1E electrical cables
- Submission for the CSNI/GREST benchmark exercise on chemical thermodynamic modeling in core-concrete interaction releases of radionuclides
- Summary of MELCOR 1.8.2 calculations for three LOCA sequences (AG, S2D, and S3D) at the Surry plant
- Summer raptors on Steens Mountain
- Surface-complexation modeling of radionuclide adsorption in subsurface environments
- TAC2D studies of Mark I containment drywell shell melt-through
- TRISO-coated particle fuel phenomenon identification and ranking tables (PIRTs) for fission product transport due to manufacturing, operations, and accidents
- Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants
- Technical basis for review of high-level waste repository modeling
- Techniques for determining probabilities of events and processes affecting the performance of geologic repositories
- The appropriate and effective use of security technologies in U.S. schools : a guide for schools and law enforcement agencies
- The impact of thermal aging on the flammability of electric cables
- The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel
- The mixing of immiscible liquid layers by gas bubbling
- The postirradiation examination of the DC melt dynamics experiments
- The probability of containment failure by direct containment heating in Zion
- The probability of containment failure by direct containment heating in Zion : [supplement]
- The response of the aerodynamic particle sizer to nonspherical particles and use in experimental determination of dynamic shape factor
- The risk management implications of NUREG-1150 methods and results
- Thermodynamic tables for nuclear waste isolation
- U.S./French joint research program regarding the behavior of polymer base materials subjected to beta radiation
- U.S.S. Iowa explosion : Sandia National Laboratories' final technical report : supplement to a report to congressional requesters
- Uncertainties associated with performance assessment of high-level radioactive waste repositories : a summary report
- Uncertainty and sensitivity analysis of early exposure results with the MACCS reactor accident consequence model
- Use of computerized microtomography to examine the relationships of sorption sites in alluvial soils to iron and pore space distributions
- Use of performance assessment in assessing compliance with the containment requirements in 40 CFR part 191
- User's manual for CONTAIN 1.1 : a computer code for severe nuclear reactor accident containment analysis
- User's manual for the NEFTRAN II computer code
- User/programmer guide for UCMD 99 component addition from magnetic tape coordinate data
- User/programmer guide for USMD86 hybrid microcircuit hole table
- VICTORIA 2.0 : a mechanistic model for radionuclide behavior in a nuclear reactor coolant system under severe accident conditions
- Validation of models of gas holdup in the CORCON code
- VICTORIA : a mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions
- Video systems for alarm assessment
- We crash, burn, and crush : a history of transportation technology at Sandia
- Weapons Evaluation Test Laboratory at Pantex : testing and data handling capabilities of Sandia National Laboratories at the Pantex Plant, Amarillo, Texas
- XSOR codes users manual
- Your gateway to California site computer support
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