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.
The Resource
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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- (EG-IN)1144456
100+ Items by the Organization Sandia National Laboratories
5 Items that are about the Organization Sandia National Laboratories
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Context of Sandia National LaboratoriesContributor of
- 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
- VICTORIA : a mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions
- Validation of models of gas holdup in the CORCON code
- 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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