U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Systems Research
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U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Systems Research
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218 Items by the Organization U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research | Division of Systems Research
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- A Computational model for critical flow through intergranular stress corrosion cracks
- A Framework for the assessment of severe accident management strategies
- A Performance indicator of the effectiveness of human-machine interfaces for nuclear power plants
- A Review of the South Texas Project probabilistic safety analysis for accident frequency estimates and containment binning
- A Risk-based review of instrument air systems at nuclear power plants
- A cause-defense approach to the understanding and analysis of common cause failures
- A demonstration experiment of steam-driven, high-pressure melt ejection : the HIPS-10S test
- A fast bottom-up algorithm for computing the cut sets of noncoherent fault trees
- A finite element analysis of a reactor pressure vessel during a severe accident
- A review of the Crystal River unit 3 probabilistic risk assessment : internal events, core damage frequency
- A review of the Three Mile Island-1 probabilistic risk assessment
- A simplified model of aerosol removal by natural processes in reactor containments
- A standard problem for HECTR-MAAP comparison : in-cavity oxidation
- A standard problem for HECTR-MAAP comparison : incomplete burning
- A summary of hydrogen-air detonation experiments
- A systematic process for developing and assessing accident management plans
- A thermodynamic model of fuel disruption in ST-1
- A unified interpretation of one-fifth to full scale thermal mixing experiments related to pressurized thermal shock
- A user's manual for the postprocessing program PSTEVNT
- Accident management information needs
- Accident management information needs for a BWR with a MARK I containment
- Advanced human-system interface design review guideline
- 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
- Air-water simulation of phenomena of corium dispersion in direct containment heating
- An Investigation of core liquid level depression in small break loss-of-coolant accidents
- An assessment of RELAP5/MOD2 applicability to loss-of-feedwater transient analysis in a Babcock and Wilcox reactor plant
- An assessment of steam-explosion-induced containment failure
- An empirical investigation of operator performance in cognitively demanding simulated emergencies
- An evaluation of the effects of local control station design configurations on human performance and nuclear power plant risk
- An evaluation of the reliability and usefulness of external-initiator PRA methodologies
- An integrated structure and scaling methodology for severe accident technical issue resolution : draft report for comment
- An international comparison of commercial nuclear power plant staffing regulations and practice : 1980-1990
- 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 long-term station blackout without automatic depressurization at Peach Bottom using MELCOR (version 1.8)
- Analysis of natural circulation during a Surry Station blackout using SCDAP/RELAP5
- Application of containment and release management strategies to PWR dry-containment plants
- Application of containment and release management to a PWR ice-condenser plant
- Assessment of candidate accident management strategies
- Assessment of databases and modeling capabilities for the CANDU 3 design
- Assessment of the XSOR codes : final report
- Assessment of the combustion model in the HECTR code
- Assessment of the potential for high-pressure melt ejection resulting from a Surry Station blackout transient
- BWR stability analysis with the BNL engineering plant analyzer
- Basic considerations in predicting error probabilities in human task performance
- Bias in peak clad temperature predictions due to uncertainties in modeling of ECC bypass and dissolved non-condensable gas phenomena
- Boron flushing during a BWR anticipated transient without scram
- CHARM, a model for aerosol behavior in time varying thermal-hydraulic conditions
- COMMIX-1C, a three-dimensional transient single-phase computer program for thermal-hydraulic analysis of single-component and multicomponent engineering systems
- CORCON-MOD3, an integrated computer model for analysis of molten core-concrete interactions : user's manual
- Calculation of absorbed doses to water pools in severe accident sequences
- Calculations to estimate the margin to failure in the TMI-2 vessel
- Class 1E digital systems studies
- Code and model extensions of the THATCH code for modular high temperature gas-cooled reactors
- Compendium of ECCS research for realistic LOCA analysis : final report
- Containment venting analysis for the Shoreham Nuclear Power Station
- 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
- Damaged fuel experiment DF-1 : results and analyses
- Data summary report for fission product release test VI-1
- Data summary report for fission product release test VI-2
- Data summary report for fission product release test VI-3
- Data summary report for fission product release test VI-4
- Data summary report for fission product release test VI-5
- Data summary report for fission product release test VI-6
- Degradation modeling with application to aging and maintenance effectiveness evaluations
- Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments
- Developing and assessing accident management plans for nuclear power plants
- Development of the NRC's human performance investigation process (HPIP)
- Diagnosis of condensation-induced waterhammer
- Effectiveness of containment sprays in containment management
- Eliciting and analyzing expert judgment : a practical guide
- Evaluation of computer-based ultrasonic inservice inspection systems
- 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
- Examination of relocated fuel debris adjacent to the lower head of the TMI-2 reactor vessel
- Experimental modeling of heat and mass transfer in a two-fluid bubbling pool with application to molten core-concrete interactions
- Experimental study on the combustion behavior of hydrogen-air mixtures with turbulent jet ignition at large scale
- 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
- FASTGRASS, a mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions : user's guide for mainframe, workstation, and personal computer applications
- FLAME facility : the effect of obstacles and transverse venting on flame acceleration and transition to detonation for hydrogen-air mixtures at large scale
- Feasibility study for improved steady-state initialization algorithms for the RELAP5 computer code
- Findings of a workshop on developing a methodology for evaluating effectiveness of nuclear power plant training
- Fission product release and fuel behavior of irradiated light water reactor fuel under severe accident conditions : the ACRR ST-1 experiment
- Flow visualization study of post critical heat flux region for inverted bubbly, slug and annular flow regimes
- Fragmentation and quench behavior of corium melt streams in water
- HECTR analyses of the Nevada Test Site (NTS) premixed combustion experiments
- High-temperature hydrogen-air-steam detonation experiments in the BNL small-scale development apparatus
- Human factors engineering guidance for the review of advanced alarm systems
- Human factors issues associated with advanced instrumentation and controls technologies in nuclear plants
- Hydrogen mixing studies (HMS) : assessment manual
- Hydrogen mixing studies (HMS) : theory and computational model
- INEL personal computer version of MACCS 1.5
- Identification and assessment of containment and release management strategies for a BWR Mark I containment
- Identification and assessment of containment and release management strategies for a BWR Mark II containment
- Identification and assessment of containment and release management strategies for a BWR Mark III containment
- Identification and evaluation of PWR in-vessel severe accident management strategies
- Implications for accident management of adding water to a degrading reactor core
- In-pile observation of fuel and clad relocation during LMFBR core-disruptive accidents
- In-vessel zircaloy oxidation/hydrogen generation behavior during severe accidents
- Industry based performance indicators for nuclear power plants
- Influence of organizational factors on performance reliability
- Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment
- Instrumentation availability for a pressurized water reactor with a large dry containment during severe accidents
- Integrated fuel-coolant interaction (IFCI 6.0) code : user's manual
- Integrated reliability and risk analysis system (IRRAS) version 2.0 user's guide
- Integrated reliability and risk analysis system (IRRAS) version 2.5
- Integrated systems analysis of the PIUS reactor
- Intentional depressurization accident management strategy for pressurized water reactors
- Iodine chemical forms in LWR severe accidents : draft report for comment
- Iodine chemical forms in LWR severe accidents : final report
- Issues and approaches for using equipment reliabilty alert levels
- Light water reactor lower head failure analysis
- Local control stations : human engineering issues and insights
- 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
- MELPROG-PWR/MOD1, a two-dimensional, mechanistic code for analysis of reactor core melt progression and vessel attack under severe accident conditions
- MORECA-2 : interactive simulator for modular high-temperature gas-cooled reactor core transients and heatup accidents with ATWS options
- Magnitude and reactivity consequences of moisture ingress into the modular high-temperature gas-cooled reactor core
- Man-machine interface issues in nuclear power plants : report on a workshop held on January 10-12, 1989
- Melt progression, oxidation, and natural convection in a severely damaged reactor core
- Metallographic and hardness examinations of TMI-2 lower pressure vessel head samples
- Methods for dependency estimation and system unavailability evaluation based on failure data statistics
- Multiloop integral system test (MIST) : MIST facility functional specification
- Multiloop integral system test (MIST) : final report
- Nuclear computerized library for assessing reactor reliability (NUCLARR)
- Nuclear computerized library for assessing reactor reliability (NUCLARR)
- Nuclear computerized library for assessing reactor reliability (NUCLARR)
- Nuclear computerized library for assessing reactor reliability (NUCLARR)
- Nuclear computerized library for assessing reactor reliability (NUCLARR)
- Organization and safety in nuclear power plants
- PARTITION, a program for defining the source term/consequence analysis interface in the NUREG-1150 probabilistic risk assessments : user's guide
- PBF severe fuel damage test 1-3 test results report
- Power burst facility (PBF) severe fuel damage test 1-4 : test results report
- Pressurized melt ejection into water pools
- Probability and consequences of misloading fuel in a PWR
- Procedures for the external event core damage frequency analyses for NUREG-1150
- Proceedings of the CSNI Specialists Meeting on Fuel-coolant Internations : held in Santa Barbara, California, USA, January 5-8, 1993
- Proceedings of the CSNI Workshop on PSA Applications and Limitations : held in Santa Fe, New Mexico, September 4-6, 1990
- Proceedings of the Digital Systems Reliability and Nuclear Safety Workshop : held at Rockville Crowne Plaza Hotel, Rockville, Maryland, September 13-14, 1993
- Quality assurance and verification of the MACCS code, version 1.5
- Quality assurance procedures for the CONTAIN severe reactor accident computer code
- Quantifying reactor safety margins : application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident
- Quantitative evaluation of surveillance test intervals including test-caused risks
- RELAP5 thermal-hydraulic analysis of the SNUPPS pressurized water reactor
- RELAP5 thermal-hydraulic analysis of the WNP1 pressurized water reactor
- RELAP5/MOD3 code manual
- Radionuclide behavior in the environment
- Reactivity accidents : a reassessment of the design-basis events
- Recriticality in a BWR following a core damage event
- Reference manual for the CONTAIN 1.1 code for containment severe accident analysis
- Relocation of metallic constituents in core debris beds
- Review and assessment of thermodynamic and transport properties for the CONTAIN code
- Review of the Brunswick Steam Electric Plant probabilisitic risk assessment
- Review of the chronic exposure pathway models in MACCS and several other well-known probabilistic risk assessment models
- Revised severe accident research program plan : FY 1990 - 1992
- Risk assessment for the intentional depressurization strategy in PWRs
- Risk impact of BWR technical specifications requirements during shutdown
- SCDAP/RELAP5/MOD2 code manual
- Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors
- Severe accident research program plan update
- Severe accident risks : an assessment for five U.S. nuclear power plants
- Solidus and liquidus temperatures of core-concrete mixtures
- Source term attenuation by water in the Mark I boiling water reactor drywell
- Specific topics in severe accident management
- Staffing decision processes and issues : case studies of seven U.S. nuclear power plants
- Steam explosions : fundamentals and energetic behavior
- Stepwise integral scaling method and its application to severe accident phenomena
- Study of operational risk-based configuration control
- Submission for the CSNI/GREST benchmark exercise on chemical thermodynamic modeling in core-concrete interaction releases of radionuclides
- Summary of a workshop on severe accident management for BWRs
- Summary of a workshop on severe accident management for PWRs
- Summary of important results and SCDAP/RELAP5 analysis for OECD LOFT experiment LP-FP-2
- Survey and assessment of conventional software verification and validation methods
- Systems analysis of the CANDU 3 reactor
- TMI-2 instrument nozzle examinations at Argonne National Laboratory : February 1991-June 1993
- TMI-2 nozzle examinations performed at the Idaho National Laboratory
- TMI-2 vessel investigation project integration report
- TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor
- TRAC-BF1/MOD1 : an advanced best-estimate computer program for BWR accident analysis
- TRAC-BF1/MOD1 models and correlations
- TRAC-PF1/MOD2 code manual
- Technical specification action statements requiring shutdown : a risk perspective with application to the RHR/SSW systems of a BWR
- The DF-4 fuel damage experiment in ACRR with a BWR control blade and channel box
- The Probability of Mark-I containment failure by melt-attack of the liner
- The Probability of liner failure in a Mark-I containment
- The cognitive environment simulation as a tool for modeling human performance and reliability
- The effects of stress on nuclear power plant operational decision making and training approaches to reduce stress effects
- The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel
- The management of ATWS by boron injection
- 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
- The use of fiber optics for remote temperature measurement in fission product release tests
- Thermal-hydraulic processes during reduced inventory operation with loss of residual heat removal
- Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B&W plant : an application of the CSAU methodology using the RELAP5/MOD3 computer code
- Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant : an application of the CSAU methodology using the BNL engineering plant analyzer
- Universal treatment of plumes and stresses for pressurized thermal shock evaluations
- User's manual for CONTAIN 1.1 : a computer code for severe nuclear reactor accident containment analysis
- 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
- Validation of risk-based performance indicators : safety system function trends
- Value-impact assessment for a candidate operating procedure upgrade program
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