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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- Workshop on Environmental Qualification of Electrical Equipment : held at Holiday Inn Crowne Plaza, Rockville, Maryland, November 15-16, 1993
- 2D/3D program work summary report
- A benchmark implementation of two dynamic methodologies for the reliability modeling of digital instrumentation and control systems
- A phenomena identification and ranking table (PIRT) exercise for nuclear power plant fire modeling applications
- 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 tool for drawing with Excel
- 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 Assessment of TRAC-PF1/MOD1 using Strathclyde 1/10 scale model refill tests
- An Assessment of TRAC-PF1/MOD1 using Strathclyde 1/10 scale model refill tests : 2nd report
- An analysis of semiscale Mod-2C S-FS-1 steam line break test using RELAP5/MOD2
- An assessment of the CORCON-MOD3 code
- 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 KS-1 experimental data on the behavior of the heated rod temperatures in the partially uncovered VVER core model using RELAP5/MOD3.2
- 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 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 the LOBI experiment test BT-56 using the RELAP5/MOD3.2 code
- Analysis of the RELAP5/MOD3.2.2beta 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/MOD
- 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
- 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 full power blackout for C.N. Almaraz with RELAP5/MOD2
- 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
- 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 verticle tubes and rod bundle tests
- Assessment of MSIV full closure for Santa Maria De Garoña nuclear power plant using TRAC-BF1 (G1J1)
- Assessment of PWR steam generator modelling in RELAP5/MOD2
- Assessment of RELAP5/MOD 2 against 25 dryout experiments conducted at the Royal Institute of Technology
- Assessment of RELAP5/MOD2 against ECN-reflood experiments
- Assessment of RELAP5/MOD2 against a 10% load rejection transient from 75% steady state in the Vandellós II nuclear power plant
- Assessment of RELAP5/MOD2 against a load rejection from 100% to 50% power in the Vandellos II nuclear power plant
- Assessment of RELAP5/MOD2 against a main feedwater turbopump trip transient in the Vandellos II nuclear power plant
- Assessment of RELAP5/MOD2 against a natural circulation experiment in nuclear power plant Borssele
- Assessment of RELAP5/MOD2 against a pressurizer spray valve inadverted fully opening transient and recovery by natural circulation in Jose Cabrera nuclear station
- Assessment of RELAP5/MOD2 against a turbine trip from 100% power in the Vandellos II nuclear power plant
- Assessment of RELAP5/MOD2 against natural circulation experiments performed with the REWET-III facility
- Assessment of RELAP5/MOD2 computer code against the natural circulation test data from Yong-Gwang unit 2
- Assessment of RELAP5/MOD2 computer code against the net load trip test data from Yong-Gwang, unit 2
- Assessment of RELAP5/MOD2 critical flow model using Marviken test data 15 and 24
- Assessment of RELAP5/MOD2 cycle 36.04 with LOFT large break LOCE L2-3
- Assessment of RELAP5/MOD2 using LOCE large break loss-of-coolant experiment L2-5
- Assessment of RELAP5/MOD2 using semiscale intermediate break loss-of-coolant experiment S-IB-3
- Assessment of RELAP5/MOD2 using semiscale large break loss-of-coolant experiment S-06-3
- Assessment of RELAP5/MOD2 using the test data of REWET-II reflooding experiment SGI/R
- Assessment of RELAP5/MOD2, cycle 36.02, using NEPTUN reflooding experimental data
- Assessment of RELAP5/MOD2, cycle 36.04 against FIX-II guillotine break experiment no. 5061
- Assessment of RELAP5/MOD2, cycle 36.04 against LOFT small break experiment L3-5
- Assessment of RELAP5/MOD2, cycle 36.04 against LOFT small break experiment L3-6
- Assessment of RELAP5/MOD2, cycle 36.04 using LOFT immediate break experiment L2-1
- Assessment of RELAP5/MOD2, cycle 36.04, against the Loviisa-2 stuck-open turbine by-pass valve transient on September 1, 1981
- Assessment of RELAP5/MOD3 against twenty-five post-dryout experiments performed at the Royal Institute of Technology
- Assessment of RELAP5/MOD3 version 5m5 using inadvertent safety injection incident data of Kori unit 3 plant
- Assessment of RELAP5/MOD3 with the LOFT L9-1/L3-3 experiment simulating an anticipated transient with multiple failures
- Assessment of RELAP5/MOD3 with the SNUF test simulating hot leg break LOCA in the view of mass and energy release analysis
- Assessment of RELAP5/MOD3.1 using LSTF ten-percent main steam-line-break test run SB-SL-01
- Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient
- Assessment of RELAP5/MOD3.2 against a main steam isolation valve closure at TRILLO I Nuclear Power Plant
- Assessment of RELAP5/MOD3.2 for steam condensation experiments in the presence of noncondensibles in a vertical tube of PCCS
- Assessment of RELAP5/MOD3.2 for thermohydraulic processes in heated rod bundles with tight lattice at CKTI Test Facility
- Assessment of RELAP5/MOD3.2 to the loss-of-residual-heat-removal event under shutdown condition
- Assessment of RELAP5/MOD3.2 using LOFT large break LOCA test, LP-02-6
- Assessment of RELAP5/MOD3.2 with the LSTF experiment simulating a loss of residual heat removal event during mid-loop operation
- Assessment of RELAP5/MOD3.2 with the semiscale natural circulation experiment, S-NC-8B
- Assessment of RELAP5/MOD3.2-NPA3.4 against a transient of high nuclear flux variation reactor trip, natural circulation and the start of a main pump in the VANDELLOS II nuclear power plant
- Assessment of RELAP5/MOD3.2-NPA3.4 against an inadvertent closure of all three MSIV's in VANDELLOS-II nuclear power plant
- Assessment of RELAP5/MOD3.2.2 gamma with the LOFT L9-3 experiment simulating an anticipated transient without scram
- Assessment of RELAP5/MOD3.2.2γ against flooding database in horizontal-to-inclined pipes
- Assessment of RELAP5/MOD3/V5m5 against the UPTF test no. 11 (countercurrent flow in PWR hot leg)
- Assessment of TRAC-PF1/MOD1 against a loss-of-grid transient in Ringhals 4 power plant
- Assessment of TRAC-PF1/MOD1 against an inadvertent feedwater line isolation transient in the Ringhals 4 power plant
- Assessment of TRAC-PF1/MOD1 against an inadvertent pressurizer spray total opening transient in José Cabrera power plant
- Assessment of TRAC-PF1/MOD1 against an inadvertent steam line isolation valve closure in the Ringhals 2 power plant
- Assessment of TRAC-PF1/MOD1 version 14.3 using separate effects critical flow and blowdown experiments
- Assessment of a pressurizer spray valve faulty opening transient at Asco nuclear power plant with RELAP5/MOD2
- Assessment of a reactor coolant pump trip for TRILLO NPP with RELAP5/MOD3.2
- Assessment of analysis methods for seismic shear wall capacity using JNES/NUPEC multi-axial cyclic and shaking table test data
- Assessment of full power turbine trip start-up test for C. Trillo I with RELAP5/MOD2
- Assessment of interfacial shear and wall heat transfer of RELAP5/MOD2/36.02 during reflooding
- Assessment of interphase drag correlations in the RELAP5/MOD2 and TRAC-PF1/MOD2 codes
- Assessment of noise level for eddy current inspection of steam generator tubes
- Assessment of risk significance associated with issues identified at D.C. Cook Nuclear Power Plant : main report
- Assessment of selected TRAC and RELAP5 calculations for Oconee-1 pressurized thermal shock study
- Assessment of single recirculation pump trip transient in Santa Maria de Garona Nuclear Power Plant with TRAC-BF1/MOD1, version 0.4
- Assessment of subcooled boiling model used in RELAP5/MOD2 (cycle 36.05, version E03) against experimental data
- Assessment of the "one feedwater pump trip transient" in Cofrentes Nuclear Power Plant with TRAC-BF1
- Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1
- Assessment study of RELAP5/MOD2 against IVO loop seal tests
- Assessment study of RELAP5/MOD2 cycle 36.04 based on pressurizer safety and relief valve tests
- Assessment study of RELAP5/MOD2 cycle 36.04 based on the DOEL-4 manual loss of load test of November 23, 1985
- Assessment study of RELAP5/MOD2 cycle 36.04 based on the commissioning test reactor trip at full load at the Philippsburg 2 nuclear power plant
- Assessment study of RELAP5/MOD2 cycle 36.05 based on the Doel 4 reactor trip of November 22, 1985
- Assessment study of RELAP5/MOD2 cycle 36.05 based on the Tihange-2 reactor trip of January 11, 1983
- Assessment study of RELAP5/MOD3.2 based on the Kalinin NPP unit-1 stop of feedwater supply to the steam generator no. 4
- Assessment study on the PMK-2 total loss of feedwater experiment using RELAP5 code
- Cable response to live fire (CAROLFIRE)
- Coexistence assessment of industrial wireless protocols in the nuclear facility environment
- Computational fluid dynamics best practice guidelines for dry cask applications : draft report for comment
- Confrentes NPP (BWR/6) ATWS (MSIVC) analysis with TRAC-BF1 : 1D vs. point kinetics and containment response
- Contrast of RELAP5/MOD3.2 results from different computing platforms
- Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments
- Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments
- Data base on nuclear power plant dose reduction research projects
- Data base on the behavior of high burnup fuel rods with Zr-1%Nb cladding and UO2 fuel (VVER type) under reactivity accident conditions
- Description and RELAP5 assessment of the PMK-2 CAMP-CLB experiment : 2% cold leg break without HPIS with secondary bleed
- Design practices for communications and workstations in highly integrated control rooms
- Design, instrumentation, and testing of a steel containment vessel model
- Development and validation of a transition boiling model for the RELAP5/MOD3 reflood simulation
- Development, implementation, and assessment of specific closure laws for inverted-annular film-boiling in a two-fluid model
- Developmental assessment of RELAP5/MOD3.1 with separate-effect and integral test experiments : model changes and options
- Dispersed flow film boiling : an investigation of the possibility to improve the models implemented in the NRC computer codes for the reflooding phase of the LOCA
- Draft regulatory guide DG-0002 : proposed appendix X, guidance on complying with new part 20 requirements to Regulatory Guide 10.8, revision 2, "Guide for the preparation of applications for medical use programs."
- Draft regulatory guide DG-1009 : standard format and content of technical information for applications to renew nuclear power plant operating licenses
- Draft regulatory guide DG-1012 : qualification and training of personnel for nuclear power plants
- Draft regulatory guide DG-1033 : (previously issued as Draft MS 140-5 and DG-1016) : nuclear power plant instrumentation for earthquakes
- Draft regulatory guide DG-1035 : (previously issued as Draft DG-1018) : restart of a nuclear power plant shut down by a seismic event
- Draft regulatory guide DG-1081 : alternative radiological source terms for evaluating design basis accidents at nuclear power reactors
- Draft regulatory guide DG-1093 : guidance and examples for identifying 10 CFR 50.2 design bases
- Draft regulatory guide DG-1095 : guidance for implementation of 10 CFR 50.59, changes, tests, and experiments
- Draft regulatory guide DG-3014 : (proposed revision to Regulatory guide [3.66]) : standard format and content of financial assurance mechanisms required for decommissioning under 10 CFR parts 30, 40, 70, and 72 : for comment
- Draft regulatory guide DG-4004 : (previously issued as DG-4003) : general site suitability criteria for nuclear power stations
- Draft regulatory guide DG-8004 : radiation protection programs for nuclear power plants
- Draft regulatory guide DG-8008 : planned special exposures
- Draft regulatory guide DG-8010 : criteria for monitoring and methods for summation of internal and external occupational doses
- Draft regulatory guide DG-8011 : radiation dose to the embryo/fetus
- Dry cask storage characterization project--phase 1 : CASTOR V/21 cask opening and examination
- EPRI/NRC-RES fire human reliability analysis guidelines : draft report for comment
- Effect of LWR coolant environments on the fatigue life of reactor materials : final report
- Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels
- Electric raceway fire barrier systems in U.S. nuclear power plants : draft report for comment
- Estimation of operator action time windows by RELAP/MOD3.3
- Evaluation of isotope migration - land burial : water chemistry at commercially operated low-level radioactive waste disposal sites
- Evaluation of reliability technology applicable to LWR operational safety
- Evaluation of station blackout accidents at nuclear power plants : technical findings related to unresolved safety issue A-44, final report
- Expert panel report on proactive materials degradation assessment
- Feasibility study for a risk-informed and performance-based regulatory structure for future plant licensing, V. 1 and 2
- Features, limitations and uncertainties in enclosure fire hazard analyses : preliminary review
- Field studies to confirm uncertainty estimates of ground-water recharge
- Final generic environmental impact statement on decommissioning of nuclear facilities
- Final report of NRC AP600 research conducted at Oregon State University
- Final report, assessment of potential phosphate ion-cementitious materials interactions
- Fracture analysis of vessels : Oak Ridge FAVOR, v.04.1, computer code : theory and implementation of algorithms, methods, and correlations
- Fracture analysis of vessels : Oak Ridge FAVOR, v.04.1, computer code : users guide
- GSI-191 PWR sump screen blockage chemical effects tests : thermodynamic simulations
- Geophysical investigations of the western Ohio-Indiana region
- Global seismicity and world cities
- Guidance on the treatment of uncertainties associated with PRAs in risk-informed decision making : draft report for comment
- Guidance on the treatment of uncertainties associated with PRAs in risk-informed decision making : main report
- Heavy-section steel technology program
- Heavy-section steel technology program
- High-temperature gas-cooled reactor safety studies for the Division of Accident Evaluation quarterly progress report
- High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research ..
- Human Event Repository and Analysis (HERA) system, overview
- Human factors considerations with respect to emerging technology in nuclear power plants
- Human factors review for severe accident sequence analysis
- Human interaction with reused soil : an information search : final report
- Human reliability data bank for nuclear power plant operations
- Implementation and assessment of improved models and options in TRAC-BF1
- In-tube steam condensation in the presence of air under transient conditions
- Independent assessment of TRAC-PF1 (version 7.0), RELAP5/MOD1 (cycle 14), and TRAC-BD1 (version 12.0) codes using separate-effects experiments
- Individual plant examination : submittal guidance : draft report for comment
- Information on the confinement capability of the facility disposal area at West Valley, New York
- Initial demonstration of the NRC's capability to conduct a performance assessment for a high-level waste repository
- Installation of RELAP/MOD3.2 on 80486 and Pentium based personal computers
- Instrumentation and controls in nuclear power plants : an emerging technologies update
- International HRA empirical study--phase 1 report : description of overall approach and pilot phase results from comparing HRA methods to similar performance data
- International code assessment and applications program : annual report
- International code assessment and applications program : summary of code assessment studies concerning RELAP5/MOD2, RELAP5/MOD3, and TRAC-B
- International collaborative project to evaluate fire models for nuclear power plant applications : summary of planning meeting held at University of Maryland, College Park, MD 20742, October 25-26, 1999 : proceedings
- Investigations of the VVER-1000 Coolant Transient Benchmark Phase 1 with the coupled code system RELAP5/PARCS
- Irradiation-assisted stress corrosion cracking of austenitic stainless steels and alloy 690 from Halden phase-II irradiations
- LBLOCA analysis in a Westinghouse PWR 3-loop design using RELAP5/MOD3
- LOFT input dataset reference document for RELAP5 validation studies
- Lessons learned in process control at the Halden Reactor Project
- Mechanical properties of thermally aged cast stainless steels from Shippingport Reactor components
- Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding : procedures and results of low temperature biaxial burst tests and axial tensile tests
- Modification of IPSN's SCANAIR fuel rod transient code for high burnup VVER fuel
- Modification of USNRC's FRAP-T6 fuel rod transient code for high burnup VVER fuel
- Multirod burst test program
- NRC iterative performance assessment phase 2 : development of capabilities for review of a performance assessment for a high-level waste repository
- NRC safety research in support of regulation
- NRC safety research in support of regulation
- NUCRAC, a code for the estimation of adversary-action consequences in the nuclear power fuel cycle
- Next generation nuclear plant phenomena identification and ranking tables (PIRTs)
- Non-destructive and failure evaluation of tubing from a retired steam generator
- Nonlinear analyses for embedded cracks under pressurized thermal shock : comparisons with FAVOR and Weibull stress approaches
- Nowadays tools for graphical post-processing of TRAC-BF1 results
- Nuclear reactor safety
- Nuclear research programs to ensure public health and safety
- Numerics and implementation of the UK horizontal stratification entrainment off-take model into RELAP5/MOD3
- ORNL rod bundle heat transfer test data
- Operating experience and aging-seismic assessment of electric motors
- PRA procedures guide : a guide to the performance of probabilistic risk assessments for nuclear power plants
- PRA procedures guide : a guide to the performance of probabilistic risk assessments for nuclear power plants
- PWR and BWR pressure vessel fluence calculation benchmark problems and solutions
- Passive nondestructive assay of nuclear materials
- Post-test analysis of LOBI test BT-12 using RELAP5/MOD2
- Post-test analysis of P5 experiment in PANDA facility with TRAC-BF1 code
- Post-test-analysis and nodalization studies of OECD LOFT experiment LP-02-6 with RELAP5/MOD2 CY36-02
- Post-test-analysis and nodalization studies of OECD LOFT experiment LP-LB-1 with RELAP5/MOD2 CY36-02
- Pre- and post-test analysis of LOBI MOD2 test ST-02 (BT-00) with RELAP5/MOD1 and MOD2 (loss of feed water)
- Preapplication safety evaluation report for the sodium advanced fast reactor (SAFR) liquid-metal reactor
- Pretest round robin analysis of a prestressed concrete containment vessel model
- Probabilistic modules for the RESRAD and RESRAD-BUILD computer codes : user guide
- Proceedings of the ... Nuclear Safety Research Conference
- Proceedings of the ... Water Reactor Safety Information Meeting
- Proceedings of the Aging Research Information Conference : held at Holiday Inn Crowne Plaza, Rockville, Maryland, March 24-27, 1992
- Proceedings of the DOE/NRC Nuclear Air Cleaning Conference
- Proceedings of the International Conference on Wire System Aging : held at DoubleTree Hotel, Rockville, Maryland, April 23-25, 2002
- Proceedings of the International Nuclear Power Plant Aging Sympoisum : held at Hyatt Regency, Bethesda, Maryland, August 30-31, 1988 and September 1, 1988
- Proceedings of the International Nuclear Power Plant Aging Symposium : held at Hyatt Regency, Bethesda, Maryland, August 30-31, 1988 and September 1, 1988
- Proceedings of the Meeting on Ultrasensitive Techniques for Measurement of Uranium in Biological Samples and the Nephrotoxicity of Uranium : held at General Services Administration, Washington, DC, December 4-5, 1985
- Proceedings of the Nuclear Safety Research Conference : held at Marriott Hotel at Metro Center, Washington, DC, October 22-24, 2001
- Proceedings of the OECD/CSNI Specialist Meeting on Advanced Instrumentation and Measurement Techniques : held at Fess Parker's Red Lion Resort, Santa Barbara, CA, March 17-20, 1997
- Proceedings of the Seminar on Leak-Before-Break : further developments in regulatory policies and supporting research : held at Taipei, Taiwan, May 11-12, 1989
- Proceedings of the Third International Atomic Energy Agency Specialists' Meeting on Subcritical Crack Growth : held at Moscow, USSR, May 14-17, 1990
- Proceedings of the Twenty-sixth Water Reactor Safety Information Meeting : held at Bethesda Marriott Hotel, Bethesda, Maryland, October 26-28, 1998 : proceedings
- Proceedings of the U.S. Nuclear Regulatory Commission ... Water Reactor Safety Information Meeting
- Proceedings of the U.S. Nuclear Regulatory Commission ... Water Reactor Safety Research Information Meeting
- Proceedings of the U.S. Nuclear Regulatory Commission : fourteenth Water Reactor Safety Information Meeting : held at National Bureau of Standards, Gaithersburg, Maryland, October 27-31, 1986
- Proceedings of the U.S. Nuclear Regulatory Commission : twenty-fifth Water Reactor Safety Information Meeting, held at Bethesda Marriott Hotel, Bethesda, Maryland, October 20-22, 1997
- Proceedings of the U.S. Nuclear Regulatory Commission, fifteenth Water Reactor Safety Information Meeting : held at National Bureau of Standards, Gaithersburg, Maryland, October 26-29, 1987
- Proceedings of the Workshop on Cement Stabilization of low-level radioactive waste : held at Gaithersburg Marriott Hotel, Gaithersburg, Maryland, May 31 - June 2, 1989
- Proceedings of the Workshop on Electronic Dosimetry : Gaithersburg, Maryland, USA, October 14-16, 1997
- Proceedings of the Workshop on Review of Dose Modeling Methods for Demonstration of Compliance with the Radiological Criteria for License Termination : held at NRC headquarters auditorium, Rockville, Maryland, USA, November 13-14, 1997
- Proceedings of the seminar on assessment of fracture prediction technology : piping and pressure vessels : held at Nashville, Tennessee, June 18, 1990
- Proceedings of workshop V : Flow and Transport through Unsaturated Fractured Rock--Related to High-Level Radioactive Waste Disposal : held at Radisson Suite Hotel, Tucson, Arizona, January 7-10, 1991
- Properties of radioactive wastes and waste containers
- Quantitative code assessment with fast Fourier transform based method improved by signal mirroring
- RELAP/MOD3 analysis of BETHSY test 6.9c : loss of RHRS, SG manway open
- RELAP5 assessment against PACTEL experimental data (revision 1)
- RELAP5 assessment on direct-contact condensation in horizontal oncurrent stratified flow
- RELAP5 assessment using LSTF test data SB-CL-18
- RELAP5 assessment using Semiscale SBLOCA test S-NH-1
- RELAP5/MOD.2 post test analysis and accuracy quantification of Lobi test BL-34
- RELAP5/MOD2 analysis of LOFT experiment L9-4
- RELAP5/MOD2 analysis of a postulated "cold leg SBLOCA" simultaneous to a "total black-out" event in the José Cabrera Nuclear Station
- RELAP5/MOD2 assessment, OECD-LOFT small break experiment LP-SB-03
- RELAP5/MOD2 calculation of OECD LOFT test LP-FW-01
- RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-1
- RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-2
- RELAP5/MOD3 assessment for calculation of safety and relief valve discharge piping hydrodynamic loads
- RELAP5/MOD3 assessment using the semiscale 50% feed line break test S-FS-11
- RELAP5/MOD3 subcooled boiling model assessment
- RELAP5/MOD3.2 assessment using GERDA small break test, 1605AA
- RELAP5/MOD3.2 post test analysis and accuracy quantification of Lobi test BL-44
- RELAP5/MOD3.2 post test analysis and accuracy quantification of SPES test SP-SB-03
- RELAP5/MOD3.2 post test analysis and accuracy quantification of SPES test SP-SB-04
- RELAP5/MOD3.2 post test calculation of the PKL-experiment PKLIII-B4.3
- RELAP5/MOD3.2 validation using BETHSY test 6.9a
- Reactor risk reference document : draft for comment
- Reactor safety issues resolved by the 2D/3D program
- Reassessment of the technical bases for estimating source terms : final report
- Recirculation suction large break LOCA analysis of the Santa Maria De Garoña nuclear power plant using TRAC-BF1 (G1J1)
- Redox and sorption reactions of iodine and cesium during transport through aquifer sediments
- Regulatory analysis guidelines of the U.S. Nuclear Regulatory Commission : final report
- Regulatory guide 1.118 : periodic testing of electric power and protection systems
- Regulatory guide 1.84 : design and fabrication code case acceptability : ASME Section III, Division 1
- Regulatory guide 4.7 : (draft was issued as DG-4004) : general site suitability criteria for nuclear power stations
- Regulatory guide 5.15 : (draft issued as DG-5005) : tamper-indicating seals for the protection and control of special nuclear material
- Regulatory guide 5.44 : (draft was DG-5007) : perimeter intrusion alarm systems
- Regulatory guide 6.9 (draft was issued as DG-6002) : establishing quality assurance programs for the manufacture and distribution of sealed sources and devices containing byproduct material
- Regulatory/backfit analysis for the resolution of unresolved safety issue A-44, station blackout
- Relevant results obtained in the analysis of LOBI/MOD2 natural circulation experiment A2-77A
- Reliability analysis of shear wall structures
- Report to Congress on abnormal occurrences
- Report to Congress on abnormal occurrences
- Reports distributed under the NRC light-water reactor safety research
- Reports distributed under the NRC reactor safety research foreign technical exchange program
- Research news
- Research project control system (RPCS) status summary report
- Resolution of Generic Safety Issue 188 : steam generator tube leaks or ruptures concurrent with containment bypass from main steam line or feedwater line breaches
- Resolution of recurring loss alarms
- Response to public comments resulting from the public workshop on nuclear power plant license renewal
- Result of BETHSY test 9.1.b using RELAP5/MOD2
- Results and insights on the impact of smoke on digital instrumentation and control
- Review and summary of TRAC assessment from the International Code Assessment and Application Program (ICAP)
- Review of TRAC calculations for Calvert Cliffs PTS study
- Review of findings for human performance contribution to risk in operating events
- Review of technical issues related to predicting isotopic compositions and source terms for high-burnup LWR fuel
- Risk-informed assessment of degraded containment vessels
- Role and direction of nuclear regulatory research
- SAFE users manual prepared for Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, under memorandum of understanding DOE 40-550-75
- SBLOCA outside containment at Browns Ferry, unit one
- SCALE: a modular code system for performing standardized computer analyses for licensing evaluation
- SCANS (Shipping Cask ANalysis System) : a microcomputer based analysis system for shipping cask design review
- STARBUCS : a prototypic SCALE control module for automated criticality safety analyses using burnup credit
- Safety research programs sponsored by Office of Nuclear Regulatory Research
- Seismic analysis of large-scale piping systems for the JNES-NUPEC ultimate strength piping test program
- Seismic considerations for the transition break size
- Seismic safety margins research program : phase 1 final report
- Seismicity 1886-89 in the southeastern United States : the aftershock sequence of the Charleston, South Carolina, earthquake / prepared by J.G. Armbruster, L. Seeber
- Sensitivity and uncertainty analysis of commercial reactor criticals for burnup credit
- Simulation of LOCA 6" and LOCA 2" transients in the RHR of a PWR under low power conditions using RELAP5/MOD3.2
- Simulation of the propagation of pressure waves in piping systems with RELAP5/MOD 3.2.2. : comparison of computed and measured results
- Special Committee review of the Nuclear Regulatory Commission's Severe accident risks report (NUREG-1150)
- Specification and verification of nuclear power plant training simulator response characteristics
- Standards development : status summary report : data for decisions, management by objectives
- Statistical methods for nuclear material management
- Status report : correlation of electrical reactor cable failure with materials degradation
- Status summary report, advanced reactor safety research
- Status summary report, water reactor safety research
- Steam generator tube integrity issues : pressurization rate effects, failure maps, leak rate correlation models, and leak rates in restricted areas
- Structural integrity of water reactor pressure boundary components : progress report ending 28 February 1977
- Structural integrity of water reactor pressure boundary components : progress report ending 31 May 1977
- Study of the effect of integral burnable absorbers for PWR burnup credit
- Study of transients related to AMSAC actuation, sensitivity analysis
- Study of unusual occurrence of a partial core uncovery in an SBLOCA scenario
- Symbolic Nuclear Analysis Package (SNAP) : Common Application Framework for Engineering Analysis (CAFEAN) preprocessor plug-in application programming interface
- Systems analysis programs for hands-on integrated reliability evaluations (SAPHIRE)
- TRAC-PF1 code assessment using OECD LOFT LP-FP-1 experiment
- TRAC-PF1/MOD1 calculations of LOFT experiment LP-02-6
- Temperatures and water potentials in shallow unsaturated alluvium next to a burial site for low-level radioactive waste, Amargosa Desert, Nye County, Nevada, 1987-96
- Test LOBI-BL06 : post-test analysis and RELAP5/MOD3.2.1 code performance assessment
- The Browns Ferry Nuclear Plant fire of 1975 and the history of NRC fire regulations
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<div class="citation" vocab="http://schema.org/"><i class="fa fa-external-link-square fa-fw"></i> Data from <span resource="http://link.library.in.gov/resource/Uboj5W1qGak/" typeof="Organization http://bibfra.me/vocab/lite/Organization"><span property="name http://bibfra.me/vocab/lite/label"><a href="http://link.library.in.gov/resource/Uboj5W1qGak/">U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research</a></span> - <span property="potentialAction" typeOf="OrganizeAction"><span property="agent" typeof="LibrarySystem http://library.link/vocab/LibrarySystem" resource="http://link.library.in.gov/"><span property="name http://bibfra.me/vocab/lite/label"><a property="url" href="http://link.library.in.gov/">Indiana State Library</a></span></span></span></span></div>
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<div class="citation" vocab="http://schema.org/"><i class="fa fa-external-link-square fa-fw"></i> Data from <span resource="http://link.library.in.gov/resource/Uboj5W1qGak/" typeof="Organization http://bibfra.me/vocab/lite/Organization"><span property="name http://bibfra.me/vocab/lite/label"><a href="http://link.library.in.gov/resource/Uboj5W1qGak/">U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research</a></span> - <span property="potentialAction" typeOf="OrganizeAction"><span property="agent" typeof="LibrarySystem http://library.link/vocab/LibrarySystem" resource="http://link.library.in.gov/"><span property="name http://bibfra.me/vocab/lite/label"><a property="url" href="http://link.library.in.gov/">Indiana State Library</a></span></span></span></span></div>
