U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research
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U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research
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The organization U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research represents an institution, an association, or corporate body that is associated with resources found in Missouri University of Science & Technology Library.
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- Draft regulatory guide DG-1187 : concrete radiation shields and generic shield testing for nuclear power plants
- Draft regulatory guide DG-1189 : an acceptable model and related statistical methods for the analysis of fuel densification
- Draft regulatory guide DG-1195 : availability of electric power sources
- Draft regulatory guide DG-1200 : an approach for determining the technical adequacy of probabilistic risk assessment results for risk-informed activities
- Draft regulatory guide DG-7004 : establishing quality assurance programs for packaging used in the transport of radioactive material
- EPRI/NRC-RES fire PRA methodology for nuclear power facilities : final report
- Regulatory guide 1.204 : guidelines for lightning protection of nuclear power plants
- Quantitative assessment of aquatic impacts of power plants
- Initial startup test program to demonstrate remote shutdown capability for water-cooled nuclear power plants
- Qualification of safety-related motor control centers for nuclear power plants
- Guidance for ITAAC closure under 10 CFR part 52
- Qualification of safety-related cables and field splices for nuclear power plants
- Information relevant to ensuring that radiation exposures at medical institutions will be as low as is reasonably achievable
- Regulatory guide 1.8 : qualification and training of personnel for nuclear power plants
- Regulatory guide 1.82 : water sources for long-term recirculation cooling following a loss-of-coolant accident
- Regulatory guide 4.19, (Task WM 408-4) : guidance for selecting sites for near-surface disposal of low-level radioactive waste
- Research activities
- Regulatory guide 1.187 : guidance for implementation of 10 CFR 50.59, changes, tests, and experiments
- Regulatory guide 1.187 (draft was issued as DG-1095) : guidance for implementation of 10 CFR 50.59, changes, tests, and experiments
- Qualification of continuous duty safety-related motors for nuclear power plants
- Training and qualification of security personnel at nuclear power reactor facilities
- Regulatory guide 1.186 : guidance and examples for identifying 10 CFR 50.2 design bases
- Three Mile Island accident of 1979 knowledge management digest
- Guidance for the assessment of beyond-design-basis aircraft impacts
- Regulatory guide 1.199 : anchoring components and structural supports in concrete
- Methods for measuring effective dose equivalent from external exposure
- Regulatory guide 1.186 (draft was issued as DG-109) : guidance and examples for identifying 10 CFR 50.2 design bases
- The Browns Ferry Nuclear Plant fire of 1975 and the history of NRC fire regulations
- ASME code cases not approved for use
- Administrative practices in radiation surveys and monitoring
- Establishing quality assurance programs for the manufacture and distribution of sealed sources and devices containing byproduct material
- Regulatory guide 1.185 : standard format and content for post-shutdown decommissioning activities report
- Alternate fire protection rule for light-water nuclear power plants
- Qualification for cement grouting for prestressing tendons in containment structures
- Regulatory guide 1.185 (draft was issued as DG-1071) : standard format and content for post-shutdown decommissioning activities report
- Regulatory guide 1.184 : decommissioning of nuclear power reactors
- An approach for plant-specific, risk-informed decisionmaking : technical specifications
- Regulatory guide 1.184 (draft was issued as DG-1067) : decommissioning of nuclear power reactors
- An approach for using probabilistic risk-assessment in risk-informed decisions on plant-specific changes to the licensing basis
- Protection of safeguards information
- Regulatory guide 1.183 : alternative radiological source terms for evaluating design basis accidents at nuclear power reactors
- Pressure-sensitive and tamper-indicating device seals for material control and accounting of special nuclear material
- Regulatory guide 1.183 (draft was issued as DG-1081) : alternative radiological source terms for evaluating design basis accidents at nuclear power reactors
- Regulatory guide 1.182 : assessing and managing risk before maintenance activities at nuclear power plants
- Standard format and content of license termination plans for nuclear power reactors
- Standard format and content for the health and safety sections of license renewal applications for uranium processing and fuel fabrication
- Bypassed and inoperable status indication for nuclear power plant safety systems
- Instrument sensing lines
- Regulatory guide 1.181 : content of the updated final safety analysis report in accordance with 10 CFR 50.71 (e)
- Measured and predicted gas flow rates through rough capillaries
- Preparation of an environmental report to support a rulemaking petition seeking an exemption for a radionuclide-containing product
- Standard format and content for emergency plans for fuel cycle and materials facilities
- Regulatory guide 1.180 : guidelines for evaluating electromagnetic and radio-frequency interference in safety-related instrumentation and control systems
- Materials and inspections for reactor vessel closure studs
- Sizing of large lead-acid storage batteries
- Regulatory guide 1.57 : design limits and loading combinations for metal primary reactor containment system components
- Service level I, II, and III protective coatings applied to nuclear power plants
- Regulatory guide 1.179 : standard format and content of license termination plans for nuclear power reactors
- Regulatory guide 1.178 : an approach for plant-specific risk-informed decisionmaking for inservice inspection of piping
- Regulatory guide 1.177 : an approach for plant-specific, risk-informed decisionmaking : technical specifications
- Regulatory guide 1.175 : an approach for plant-specific, risk-informed decisionmaking: inservice testing
- Regulatory guide 1.174 : an approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the licensing basis
- Regulatory guide 1.173 : developing software life cycle processes for digital computer software used in safety systems of nuclear power plants
- Regulatory guide 1.172 : software requirements specifications for digital computer software used in safety systems of nuclear power plants
- Regulatory guide 1.171 : software unit testing for digital computer software used in safety systems of nuclear power plants
- Regulatory guide 1.68 : initial test programs for water-cooled nuclear power plants
- Planned special exposure
- Regulatory guide 1.170 : software test documentation for digital computer software used in safety systems of nuclear power plants
- Concrete shields and generic shield testing for nuclear power plants
- Regulatory guide 1.168 (draft was issued as DG-1123 : verification, validation, reviews, and audits for digital computer software used in safety systems of nuclear power plants
- Regulatory guide 1.105 : (draft was DG-1045) : setpoints for safety-related instrumentation
- Regulatory analysis guidelines of the U.S. Nuclear Regulatory Commission, draft report for comment
- Physical models for design and operation of hydraulic structures and systems for nuclear power plants
- Regulatory Guide 1.209 : guidelines for environmental qualification of safety-related computer-based instrumentation and control systems in nuclear power plants
- Containment isolation provisions for fluid systems
- Guidance to operators at the controls and to senior operators in the control room of a nuclear power unit
- Regulatory Guide 1.208 : a performance-based approach to define the site-specific earthquake ground motion
- Regulatory Guide 1.207 : guidelines for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors
- Fire protection for nuclear power plants
- Regulatory Guide 1.205 : risk-informed, performance-based fire protection for existing light-water nuclear power plants
- Control of electroslag weld properties
- Guide for the preparation of applications for medical use programs
- Regulatory Guide 1.203 : transient and accident analysis methods
- Control of preheat temperature for welding of low-alloy steel
- Regulatory Guide 1.202 : standard format and content of decommissioning cost estimates for nuclear power reactors
- Regulatory guide 1.61 : damping values for seismic design of nuclear power plants
- Control of stainless steel weld cladding of low-alloy steel components
- Control of the processing and use of stainless steel
- Individual plant examination : submittal guidance, final report
- Regulatory Guide 1.201 : guidelines for categorizing structures, systems, and components in nuclear power plants according to their safety significance
- Regulatory Guide 1.200 : an approach for determining the technical adequacy of probabilistic risk assessment results for risk-informed activities
- Regulatory guide 1.198 : procedures and criteria for assessing seismic soil liquefaction at nuclear power plant sites
- Regulatory guide 1.197 : demonstrating control room envelope integrity at nuclear power reactors
- Criteria for use of computers in safety systems of nuclear power plants
- Inservice inspection code case acceptability, ASME section XI, division 1
- Regulatory guide 1.196 : control room habitability at light-water nuclear power reactors
- Cyber security programs for nuclear facilities
- General fire protection guide for plutonium processing and fuel fabrication plants
- Regulatory guide 1.195 : methods and assumptions for evaluating radiological consequences of design basis accidents at light-water nuclear power reactors
- Regulatory guide 1.54 : service level I, II, and III protective coatings applied to nuclear power plants
- Design, construction, and inspection of embankment retention systems at fuel cycle facilities
- Regulatory guide 1.194 : atmospheric relative concentrations for control room radiological habitability assessments at nuclear power plants
- Nuclear power plant simulation facilities for use in operator training, license examinations, and applicant experience requirements
- Regulatory guide 1.193 : ASME code cases not approved for use
- Design, construction, and inspection of embankment retention systems at uranium recovery facilities
- Nuclear criticality safety standards for fuels and material facilities
- Regulatory guide 1.192 : operation and maintenance code case acceptability, ASME OM code
- Manual initiation of protective actions
- Regulatory guide 1.191 : fire protection program for nuclear power plants during decommissioning and permanent shutdown
- Regulatory guide 1.190 : calculational and dosimetry methods for determining pressure vessel neutron fluence
- Regulatory guide 1.189 (draft was issued as DG-1097) : fire protection for operating nuclear power plants
- Regulatory guide 1.37 : quality assurance requirements for cleaning of fluid systems and associated components of water-cooled nuclear power plants
- Regulatory guide 1.188 : standard format and content for applications to renew nuclear power plant operating licenses
- Disposition of recommendations of the National Research Council in the report Revitalizing nuclear safety research
- Draft Regulatory Guide DG-1149 : qualification of safety-related motor control centers for nuclear power plants
- Draft Regulatory Guide DG-1198 : physical models for design and operation of hydraulic structures and systems for nuclear power plants
- Draft Regulatory Guide DG-1203 : containment performance for pressure loads
- Draft Regulatory Guide DG-1205 : bypassed and inoperable status indication for nuclear power plant safety systems
- Draft regulatory guide DG-1077 : guidelines for environmental qualification of microprocessor-based equipment important to safety in nuclear power plants
- Draft regulatory guide DG-1132 : qualification of safety-related cables and field splices for nuclear power plants
- Draft regulatory guide DG-1139 : risk-informed, performance-based fire protection for existing light-water nuclear power plants
- Draft regulatory guide DG-1142 : guidelines for environmental qualification of safety-related computer-based instrumentation and control systems in nuclear power plants
- Draft regulatory guide DG-1144 : guidelines for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors
- Draft regulatory guide DG-1146 : a performance-based approach to define the site-specific earthquake ground motion
- Draft regulatory guide DG-1175 : seismic qualification of electric and active mechanical equipment and functional qualification of active mechanical equipment for nuclear power plants
- Draft regulatory guide DG-1178 : instrument sensing lines
- Draft regulatory guide DG-1186 : measuring, evaluating, and reporting radioactive material in liquid and gaseous effluents and solid waste
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- 2D/3D Program work summary report
- A Comparative analysis of the International Standard Problem 8 calculations
- A Consensus estimation study of nuclear power plant structural loads
- A Description and assessment of RAMONA-3B MOD.0 Cycle 4 : a computer code with three-dimensional neutron kinetics for BWR system transients
- A Guide for reviewing estimates of production-cost increases that result from nuclear power plant outages
- A Methodology for allocating nuclear power plant control functions to human or automatic control
- A New implicit numerical solution scheme in the COMMIX-1A computer program
- A Numerical procedure for calculating steady/unsteady, single-phase/two-phase three-dimensional fluid flow with heat transfer
- A Parametric study of containment emergency sump performance
- A Simulator-based study of human errors in nuclear power plant control room tasks
- A Study of nonequilibrium flashing of water in a converging-diverging nozzle
- A Test of the Controllable Unit Approach (CUA) concept in a low-enrichment-uranium fuel-fabrication facility
- A Volume-weighted skew-upwind difference scheme in COMMIX
- A benchmark implementation of two dynamic methodologies for the reliability modeling of digital instrumentation and control systems
- A compendium of computer codes for light water reactor analysis
- A compilation of elevated temperature concrete material property data and information for use in assessments of nuclear power plant reinforced concrete structures
- A conditional model of peak and minimum loads and the load duration curve for electricity
- A description of the hardware and software of the Power Spectral Density Recognition (PSDREC) continuous on-line reactor surveillance system (California distribution)
- A formalized approach for the collection of HRA data from nuclear power plant simulators
- A generalized age-dependent lung model with applications to radiation standards
- A generalized procedure for generating flaw-related inputs for the FAVOR code
- A large scale validation of a methodology for assessing software reliability
- A model for the migration of the fission products along the coolant channels of a high temperature gas cooled reactor following a hypothetical accident of complete loss of cooling
- A new steam cooled reactor
- A one-dimensional neutron kinetics model for the THOR code
- A parametric study of PWR pressure vessel integrity during overcooling accidents, considering both 2-D and 3-D flaws
- A phenomena identification and ranking table (PIRT) exercise for nuclear power plant fire modeling applications
- A plan for the modification and assessment of TRAC-PF1/MOD2 for use in analyzing CANDU 3 transient thermal-hydraulic phenomena
- A quantitative impact assessment of hypothetical spent fuel reconfiguration in spent fuel storage casks and transportation packages
- A review of the effects of radiation on microstructure and properties of concretes used in nuclear power plants
- A review of the impact of copper released into marine and estuarine environments
- A review of the resolving power of reflection seismology methods to detect subsurface faults and/or changes in layer thickness
- A single-phase pump model for analysis of LMFBR heat transport systems
- A standardized methodology for the linkage of computer codes : application to RELAP5/MOD3.2
- A study of control room staffing levels for advanced reactors
- A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3
- A subsurface decision model for supporting environmental compliance
- A survey of methods for improving operator acceptance of computerized aids
- A tool for drawing with Excel
- A torsional ultrasonic technique for LWR liquid level measurement
- A user's guide for the stock-recruitment model validation program
- A varying elasticity model of electricity demand with given appliance saturation
- ARMA models for earthquake ground motions : seismic safety margins research program
- ATWS at Browns Ferry Unit One : accident sequence analysis
- Activity and fluence calculations for the startup and two-year irradiation experiments performed at the poolside facility
- Advanced two-phase flow instrumentation program quarterly progress report for ...
- Advanced two-phase flow instrumentation program quarterly progress report for April-June 1980
- Advanced two-phase flow instrumentation program quarterly progress report for October-December 1980
- Advanced two-phase flow instrumentation program, quarterly progress report for October-December 1979
- Advanced two-phase instrumentation program quarterly progress report for July-September 1977
- Aerosol release and transport program quarterly progress report
- Aging Research Information Conference--abstracts of papers held at Holiday Inn Crowne Plaza, Rockville, Maryland, March 24-27, 1992
- An Appraisal of possible combustion hazards associated with a high-temperature gas-cooled reactor
- An advanced thermohydraulic simulation code for pool-type LMFBRs : (SSC-P code)
- An advanced thermohydraulic simulation code for transients in LMFBRs (SSC-L code)
- An analysis of transient film boiling of high-pressure water in a rod bundle
- An approach for validating actinide and fission product burnup credit criticality safety analyses--isotopic composition predictions
- An assessment of thermal gradient tube results from the HI series of fission product release tests
- An econometric study of electricity demand by manufacturing industries
- An empirical model of impingement impact
- An evaluation of seismic qualification tests for nuclear power plant equipment : final report, September 1, 1976-August 31, 1978
- An evaluation of the unloading compliance procedure for J integral testing in the hot cell : final report
- An evaluation of ultrasonic phased array testing for cast austenitic stainless steel pressurizer surge line piping welds
- An examination of the size effects and data scatter observed in small-specimen cleavage fracture toughness testing
- An identification of processes and parameters of importance to estimation of uncertainties in long-term collective dose and health effects resulting from geologic disposal of high-level radioactive waste
- An implicit steady-state initialization package for the RELAP5 computer code
- An improvement in the calculation of turbulent friction in smooth concentric annuli
- An integrated system for forecasting electric energy and load for states and utility service areas
- An overview of rod-bundle termal-hydraulic analysis
- Analyses of KS test data on the heated rod bundle temperature behavior in RBMK-1500 core model under stop and recovery flow using RELAP5/MOD3.2 and RELAP5/MOD3.2.2 GAMMA
- Analysis of LOBI test BLO2 (three percent cold leg break) with RELAP5 code
- Analysis of LOFT test L5-1 using RELAP5/MOD2
- Analysis of PANDA experiments P3 and P6 using RELAP5/MOD3.2
- Analysis of REALP5/MOD3.3 prediction of 2-inch loss-of-coolant accident at Krško Nuclear Power Plant
- Analysis of a double-ended cold-leg break simulation--THTF test 3.05.5B
- Analysis of experimental data for high-burnup PWR spent fuel isotopic validation : Vandellós II reactor
- Analysis of inadvertent pressurizer spray valve opening real transient with RELAP5/MOD3.2
- Analysis of pin-by-pin effects for LWR rod ejection accident
- Analysis of semiscale test S-LH-1 using RELAP5/MOD2
- Analysis of semiscale test S-LH-2 using RELAP5/MOD2
- Analysis of surface displacements of Zircaloy fuel cladding in the HOBBIE creepdown irradiation experiments
- Analysis of temperature data from the ORR-PSF irradiation experiment : methodology and computer software
- Analysis of the LOBI experiment test BT-56 using the RELAP5/MOD3.2 code
- Analysis of the RELAP5/MOD3.2.2 beta critical flow models and assessment against critical flow data from the Marviken tests
- Analysis of the THETIS boildown experiments using RELAP5/MOD2
- Analysis of the UPTF separate effects test 11 (steam-water countercurrent flow in the broken loop hot leg) using RELAP5/MOD2
- Analysis of the VENUS PWR engineering mockup experiment : phase 1, source distribution
- Analysis of the VTI test data on the behavior of the heated rod temperatures in the partially uncovered VVER-440 core model using RELAP5/MOD3.2.2 gamma
- Analysis of the critical flow model in TRAC-BF1
- Analysis of the heat and mass transfer processes of a UO2 bubble in sodium for the fuel aerosol simulant test (FAST)
- Analysis of the performance of the Westinghouse Reactor Vessel Level Indicating System for tests at SEMISCALE
- Analysis of training and certification of operations technicians at independent spent fuel storage installations
- Analytic rebalance technique for pressure calculation in two-phase flow systems
- Analytical techniques for stress analysis of the nuclear steam-supply system : a bibliography
- Analyzing safeguards alarms and response decisions
- Annotated bibliography of safety-related events at boiling-water nuclear power plants as reported in 1977
- Annotated bibliography of safety-related events at pressurized-water nuclear power plants as reported in 1977
- Annotated bibliography on the transportation and handling of radioactive materials - 3
- Appendix to radon and radon-daughter concentrations in air in the vicinity of the Anaconda Uranium Mill
- Applicability of LEFM to the analysis of PWR vessels under LOCA-ECC thermal shock
- Application of RELAP5/MOD3.1 to ATWS analysis of control rod withdrawal from 1% power level
- Application of RELAP5/MOD3.2 to the loss-of-residual-heat-removal event under shutdown condition
- Application of RELAP5/MOD3.2 to the loss-of-residual-heat-removal event under shutdown condition
- Application of TRACE V5.0 P2 to China domestic PWR LBLOCA analysis
- Application of TRACE V5.0 P2 to natural circulation reactor safety analysis
- Application of bounding spectra to seismic design of piping based on the performance of above ground piping in power plants subjected to strong motion earthquakes
- Application of full power blackout for C. N. Almaraz with RELAP5/MOD2
- Application of model abstraction techniques to simulate transport in soils
- Application of point precipitation frequency estimates to watersheds
- Application of reduction methods to nuclear power plant structures
- Application of static and dynamic crack arrest theory to thermal shock experiment TSE-4
- Application of surface complexation modeling to selected radionuclides and aquifer sediments
- Applications of energy release rate techniques to part-through cracks in plates and cylinders
- Applying ultrasonic testing in lieu of radiography for volumetric examination of carbon steel piping
- Aqueous iodine chemistry in LWR accidents : review and assessment
- Argonne model boiler test results
- Aspects of the winter predator-prey relationship between sauger and threadfin shad in Watts Bar Reservoir, Tennessee
- Assessing the potential for biorestoration of uranium in situ recovery sites
- Assessment and application of blackout transients at Asco nuclear power plant with RELAP5/MOD2
- Assessment of BETHSY test 9.1.b using RELAP5/MOD3
- Assessment of CCFL model of RELAP5/MOD3 against simple vertical tubes and rod bundle tests
- Assessment of MSIV full closure for Santa Maria De Garoña nuclear power plant using TRAC-BF1 (G1J1)
- Assessment of NDE methods on inspection of HDPE butt fusion piping joints for lack of fusion
- Assessment of PWR steam generator modelling in RELAP5/MOD2
- Assessment of RELAP5/ MOD2 against a main feedwater turbopump trip transient in the Vandellos II nuclear power plant
- Assessment of RELAP5/MOD 2 against Marviken jet impingement test 11 level swell
- Assessment of RELAP5/MOD 2, cycle 36, against FIX-II split break experiment no. 3027
- Assessment of RELAP5/MOD2 against ECN-reflood experiments