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Publications Prepared by NRC Contractors

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Documentation of technical, regulatory, or administrative information about NRC programs or activities prepared by a contractor. Other contractor reports may be available in ADAMS.

Document Identifier Title
NUREG/CR-0041 Manual of Respiratory Protection Against Airborne Radioactive Material
NUREG/CR-0075 Accidental Vapor Phase Explosions on Transportation Routes Near Nuclear Power Plants: Final Report January – April 1977
NUREG/CR-0152 Development and Verification of Fire Tests for Cable Systems and System Components: Quarterly Reports 2 and 3, September 1, 1977 – February 28, 1978
NUREG/CR-0200 SCALE Ver 4.4: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation
NUREG/CR-0381 A Preliminary Report on Fire Protection Research Program Fire Barriers and Fire Retardant Coatings Tests
NUREG/CR-0468 Nuclear Power Plant Fire Protection — Fire Barriers (Subsystems Study Task 3)
NUREG/CR-0488 Nuclear Power Plant Fire Protection — Fire Detection (Subsystems Study Task 2)
NUREG/CR-0596 A Preliminary Report on Fire Protection Research Program, Fire Barriers and Suppression (September 15, 1978 Test)
NUREG/CR-0636 Nuclear Power Plant Fire Protection — Ventilation (Subsystems Study Task 1)
NUREG/CR-0654 Nuclear Power Plant Fire Protection — Fire-Hazards Analysis (Subsystems Study Task 4)
NUREG/CR-0833 Fire Protection Research Program Corner Effects Tests
NUREG/CR-1156 Environmental Assessment of Ionization Chamber Smoke Detectors Containing Am-241
NUREG/CR-1184 Evaluation of Simulator Adequacy for the Radiation Qualification of Safety-Related Equipment
NUREG/CR-1405 The NACOM Code for Analysis of Postulated Sodium Spray Fires in LMFBRs
NUREG/CR-1429 Seismic Review Table
NUREG/CR-1444 Investigation of Distorted-Geometry Simulation of Pool Dynamics in Horizontal-Vent BWR Containments
NUREG/CR-1552 Development and Verification of Fire Tests for Cable Systems and System Components: Quarterly Report 12, March – May 1980
NUREG/CR-1614 Approaches to Acceptable Risk: A Critical Guide
NUREG/CR-1682 Electrical Insulators in a Reactor Accident Environment
NUREG/CR-1798 Acceptance and Verification for Early Warning Fire Detection Systems: Interim Guide
NUREG/CR-1819 Development and Testing of a Model for Fire Potential in Nuclear Power Plants
NUREG/CR-1916 A Risk Comparison
NUREG/CR-1930 Index of Risk Exposure and Risk Acceptance Criteria
NUREG/CR-2015 Seismic Safety Margins Research Program Phase I Final Report
NUREG/CR-2040 A Study of the Implications of Applying Quantitative Risk Criteria in the Licensing of Nuclear Power Plants in the United States
NUREG/CR-2258 Fire Risk Analysis for Nuclear Power Plants
NUREG/CR-2260 Technical Basis for Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants"
NUREG/CR-2269 Probabilistic Models for the Behavior of Compartment Fires
NUREG/CR-2300 PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants
NUREG/CR-2321 Investigation of Fire Stop Test Parameters
NUREG/CR-2377 Test and Criteria for Fire Protection of Cable Penetrations
NUREG/CR-2409 Requirements for Establishing Detector Siting Criteria in Fires InvolvingElectrical Materials
NUREG/CR-2431 Burn Mode Analysis of Horizontal Cable Tray Fires
NUREG/CR-2475 Hydrogen Combustion Characteristics Related to Reactor Accidents
NUREG/CR-2486 Final Results of the Hydrogen Igniter Experimental Program
NUREG/CR-2490 Hazards to Nuclear Power Plants from Large Liquefied Natural Gas (LNG) Spills on Water
NUREG/CR-2607 Fire Protection Research Program for the U.S. Nuclear Regulatory Commission: 1975–1981
NUREG/CR-2650 Allowable Shipment Frequencies for the Transport of Toxic Gases Near Nuclear Power Plants
NUREG/CR-2658 Characteristics of Combustion Products: A Review of the Literature
NUREG/CR-2680 Seismic Safety Margins Research Program: Equipment Fragility Data Base
NUREG/CR-2726 Light Water Reactor Hydrogen Manual
NUREG/CR-2730 Hydrogen Burn Survival: Preliminary Thermal Model and Test Results
NUREG/CR-2815 Probabilistic Safety Analysis Procedures Guide
NUREG/CR-2858 PAVAN: An Atmospheric-Dispersion Program for Evaluating Design-Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations
NUREG/CR-2868 Aging Effects on Fire-Retardant Additives in Organic Materials for Nuclear Plant Applications
NUREG/CR-2927 Nuclear Power Plant Electrical Cable Damageability Experiments
NUREG/CR-3037 User's Manual for FIRIN: A Computer Code to Estimate Accidental Fire and Airborne Releases in Nuclear Fuel Cycle Facilities
NUREG/CR-3122 Potentially Damaging Failure Modes of High- and Medium-Voltage Electrical Equipment
NUREG/CR-3139 Scenarios and Analytical Methods for UF6 Releases at NRC-Licensed Fuel Cycle Facilities
NUREG/CR-3192 Investigation of Twenty-Foot Separation Distance as a Fire Protection Method as Specified in 10 CFR 50, Appendix R
NUREG/CR-3239 COMPBRN — A Computer Code for Modeling Compartment Fires
NUREG/CR-3242 >The Los Alamos National Laboratory/New Mexico State University Filter Plugging Test Facility: Description and Preliminary Test Results
NUREG/CR-3263 Status Report: Correlation of Electrical Cable Failure with Mechanical Degradation
NUREG/CR-3330 Vulnerability of Nuclear Power Plant Structures to Large External Fires
NUREG/CR-3385 Measures of Risk Importance and Their Applications
NUREG/CR-3468 Hydrogen:Air:Steam Flammability Limits and Combustion Characteristics in the FITS Vessel
NUREG/CR-3493 A Review of the Limerick Generating Station Severe Accident Risk Assessment: Review of Core-Melt Frequency
NUREG/CR-3521 Hydrogen-Burn Survival Experiments at Fully Instrumented Test Site (FITS)
NUREG/CR-3527 Material Transport Analysis for Accident-Induced Flow in Nuclear Facilities
NUREG/CR-3532 Response of Rubber Insulation Materials to Monoenergetic Electron Irradiations
NUREG/CR-3629 The Effect of Thermal and Irradiation Aging Simulation Procedures on Polymer Properties
NUREG/CR-3638 Hydrogen-Steam Jet-Flame Facility and Experiments
NUREG/CR-3656 Evaluation of Suppression Methods for Electrical Cable Fires
NUREG/CR-3719 Detonation Calculations Using a Modified Version of CSQII: Examples for Hydrogen-Air Mixtures
NUREG/CR-3735 Accident-Induced Flow and Material Transport in Nuclear Facilities: A Literature Review
NUREG/CR-3805 Engineering Characterization of Ground Motion: Task II: Summary Report
NUREG/CR-3922 Survey and Evaluation of System Interaction Events and Sources
NUREG/CR-4112 Investigation of Cable and Cable System Fire Test Parameters
NUREG/CR-4138 Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests
NUREG/CR-4229 Evaluation of Current Methodology Employed in Probabilistic Risk Assessment (PRA) of Fire Events at Nuclear Power Plants
NUREG/CR-4230 Probability-Based Evaluation of Selected Fire Protection Features in Nuclear Power Plants
NUREG/CR-4231 Evaluation of Available Data for Probabilistic Risk Assessments (PRA) of Fire Events at Nuclear Power Plants
NUREG/CR-4264 Investigation of High-Efficiency Particulate Air Filter Plugging by Combustion Aerosols
NUREG/CR-4310 Investigation of Potential Fire-Related Damage to Safety-Related Equipment in Nuclear Power Plants
NUREG/CR-4321 Full-Scale Measurements of Smoke Transport and Deposition in Ventilation System Ductwork
NUREG/CR-4330 Review of Light Water Reactor Regulatory Requirements
NUREG/CR-4432 Comparison of Dynamic Characteristics of Fukushima Nuclear Power Plant Containment Building Determined From Tests and Earthquakes
NUREG/CR-4461 Tornado Climatology of the Contiguous United States
NUREG/CR-4479 The Use of a Field Model To Assess Fire Behavior in Complex Nuclear Power Plant Enclosures: Present Capabilities and Future Prospects
NUREG/CR-4482 Recommendations To The Nuclear Regulatory Commission On Trial Guidelines For Seismic Margin Reviews Of Nuclear Power Plants — Draft Report for Comment
NUREG/CR-4513 Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems
NUREG/CR-4517 Design Features for Enhancing International Safeguards of Away-from- Reactor Dry Storage for Spent LWR Fuel
NUREG/CR-4527 An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets
NUREG/CR-4534 Analysis of Diffusion Flame Tests
NUREG/CR-4561 FIRAC User's Manual: A Computer Code to Simulate Fire Accidents in Nuclear Facilities
NUREG/CR-4566 COMPBRN III - A Computer Code for Modeling Compartment Fires
NUREG/CR-4570 Description and Testing of an Apparatus for Electrically Initiating Fires Through Simulation of a Faulty Connection
NUREG/CR-4586 Users' Guide for a Personal-Computer-Based Nuclear Power Plant Fire Data Base
NUREG/CR-4596 Screening Tests of Representative Nuclear Power Plant Components Exposed to Secondary Environments Created by Fires
NUREG/CR-4638 Transient Fire Environment Cable Damageability Test Results
NUREG/CR-4667 Environmentally Assisted Cracking in Light Water Reactors: Annual Report
NUREG/CR-4674 Precursors to Potential Severe Core Damage Accidents: 1998 A Status Report
NUREG/CR-4679 Quantitative Data on the Fire Behavior of Combustible Materials Found in Nuclear Power Plants: A Literature Review
NUREG/CR-4680 Heat and Mass Release for Some Transient Fuel Source Fires: A Test Report
NUREG/CR-4681 Enclosure Environment Characterization Testing for the Base Line Validation of Computer Fire Simulation Codes
NUREG/CR-4736 Combustion Aerosols Formed During Burning of Radioactively Contaminated Materials, Experimental Results
NUREG/CR-4775 Guide for Preparing Operating Procedures for Shipping Packages
NUREG/CR-4826 Seismic Margin Review of the Maine Yankee Atomic Power Station
NUREG/CR-4829 Shipping Container Response to Severe Highway and Railway Accident Conditions
NUREG/CR-4830 MELCOR Validation and Verification: 1986 Papers
NUREG/CR-4839 Methods for External Event Screening Quantification: Risk Methods Integration and Evaluation Program (RMIEP) Methods Development
NUREG/CR-4840 Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150
NUREG/CR-4855 Development and Application of a Computer Model for Large-Scale Flame Acceleration Experiments
NUREG/CR-4905 Detonability of H2-Air-Diluent Mixtures
NUREG/CR-5037 Fire Environment Determination in the LaSalle Nuclear Power Plant Control Rroom
NUREG/CR-5076 An Approach to the Quantification of Seismic Margins in Nuclear Power Plants: The Importance of BWR Plant Systems and Functions to Seismic Margins
NUREG/CR-5079 Experimental Results Pertaining to the Performance of Thermal Igniters
NUREG/CR-5117 Steam Generator Tube Integrity Program/Steam Generator Group Project: Final Project Summary Report
NUREG/CR-5176 Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two Representative Nuclear Power Plants
NUREG/CR-5233 A Computer Code for Fire Protection and Risk Analysis of Nuclear Plants
NUREG/CR-5260 Individual Plant Examinations for External Events: Review Plan and Evaluation Criteria
NUREG/CR-5275 FLAME Facility: The Effect of Obstacles and Transverse Venting on Flame Acceleration and Transition to Detonation for Hydrogen-Air Mixtures at Large Scale
NUREG/CR-5279 Sulfate-Attack Resistance and Gamma-Irradiation Resistance of Some Portland Cement Based Mortars
NUREG/CR-5281 Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent Fuel Pools
NUREG/CR-5347 Recommendations for Resolution of Public Comments on USI A-40, “Seismic Design Criteria”
NUREG/CR-5384 A Summary of Nuclear Power Plant Fire Safety Research at Sandia National Laboratories, 1975–1987
NUREG/CR-5385 Initial Assessment of the Mechanisms and Significance of Low-Temperature Embrittlement of Cast Stainless Steels in LWR Systems
NUREG/CR-5392 Elements of an Approach to Performance-Based Regulatory Oversight
NUREG/CR-5457 A Review of the Three Mile Island-1 Probabilistic Risk Assessment
NUREG/CR-5500 Reliability Study
NUREG/CR-5512 Residual Radioactive Contamination From Decommissioning: User's Manual DandD Version 2.1
NUREG/CR-5525 Hydrogen-Air-Diluent Detonation Study for Nuclear Reactor Safety Analyses
NUREG/CR-5546 An Investigation of the Effects of Thermal Aging on the Fire Damageability of Electric Cables
NUREG/CR-5580 Evaluation of Generic Issue 57
NUREG/CR-5585 The High Level Vibration Test Program - Final Report
NUREG/CR-5591 Heavy-Section Steel Irradiation Program: Progress Report April 1997 - March 1998
NUREG/CR-5609 Electromagnetic Compatibility Testing for Conducted Susceptibility Along Interconnecting Signal Lines
NUREG/CR-5619 The Impact of Thermal Aging on the Flammability of Electric Cables
NUREG/CR-5632 Incorporating Aging Effects into Probabilistic Risk Assessment — A Feasibility Study Utilizing Reliability Physics Models
NUREG/CR-5655 Submergence and High Temperature Steam Testing of Class lE Electrical Cables
NUREG/CR-5669 Evaluation of Exposure Limits to Toxic Gases for Nuclear Reactor Control Room Operators
NUREG/CR-5679 Design, Instrumentation and Testing of a Steel Containment Vessel Model
NUREG/CR-5694 Results of Field Studies at the Maricopa Environmental Monitoring Site, Arizona
NUREG/CR-5698 Comparing Monitoring Strategies at the Maricopa Environmental Monitoring Site, Arizona
NUREG/CR-5704 Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels
NUREG/CR-5732 Iodine Chemical Forms in LWR Severe Accidents
NUREG/CR-5733 Re-evaluation of Regulatory Guidance Provided in Regulatory Guides 1.142 and 1.143
NUREG/CR-5734 Recommendations to the NRC on Acceptable Standard Format and Content for the Fundamental Nuclear Material Control (FNMC) Plan Requiredfor Low-Enriched Uranium Enrichment Facilities
NUREG/CR-5736 Evaluation of WF-70 Weld Metal From the Midland Unit 1 Reactor Vessel
NUREG/CR-5737 Hydrogeologic Performance Assessment of the Commercial Low-Level Radioactive Waste Disposal Facility Near West Valley, New York
NUREG/CR-5738 Field Investigations for Foundations of Nuclear Power Facilities
NUREG/CR-5739 Laboratory Investigations of Soils and Rocks For Engineering Analysis and Design of Nuclear Power Facilities
NUREG/CR-5741 Technical Bases for Regulatory Guide for Soil Liquefaction
NUREG/CR-5789 Risk Evaluation for a Westinghouse PWR, Effects of Fire Protection System Actuation on Safety-Related Equipment: Evaluation of Generic Issue 57
NUREG/CR-5790 Risk Evaluation for a B&W Pressurized Water Reactor, Effects of Fire Protection System Actuation on Safety-Related Equipment: Evaluation of Generic Issue 57
NUREG/CR-5791 Risk Evaluation for a General Electric BWR, Effects of Fire Protection System Actuation on Safety-Related Equipment: Evaluation of Generic Issue 57
NUREG/CR-5912 Review of the Technical Basis and Verification of Current Analysis Methods Used to Predict Seismic Response of Spent Fuel Storage Racks
NUREG/CR-6017 Fire Modeling of the Heiss Dampf Reaktor Containment
NUREG/CR-6042 Perspectives on Reactor Safety
NUREG/CR-6078 Analysis of Crack Initiation and Growth in the High Level Vibration Test at Tadotsu
NUREG/CR-6082 Data Communications
NUREG/CR-6083 Reviewing Real-Time Performance of Nuclear Reactor Safety Systems
NUREG/CR-6090 The Programmable Logic Controller and Its Application in Nuclear Reactor Systems
NUREG/CR-6093 An Analysis of Operational Experience During Low Power and Shutdown and a Plan for Addressing Human Reliability Assessment Issues
NUREG/CR-6095 Aging, Loss-of-Coolant Accident (LOCA), and High Potential Testing of Damaged Cables
NUREG/CR-6101 Software Reliability and Safety in Nuclear Reactor Protection Systems
NUREG/CR-6115 PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions
NUREG/CR-6119 MELCOR Computer Code Manuals
NUREG/CR-6142 Tensile-Property Characterization of Thermally Aged Cast Stainless Steels
NUREG/CR-6150 SCDAP/RELAP5/MOD 3.3 Code Manual
NUREG/CR-6173 A Summary of the Fire Testing Program at the German HDR Test Facility
NUREG/CR-6212 Value of Public Health and Safety Actions and Radiation Dose Avoided
NUREG/CR-6213 High-Temperature Hydrogen-Air- Steam Detonation Experiments in the BNL Small-Scale Development Apparatus
NUREG/CR-6214 Production and Testing of the Revised VITAMIN-B6 Fine-Group and the BUGLE-93 Broad-Group Neutron/Photon Cross-Section Libraries Derived From END/B-VI.3 Nuclear Data
NUREG/CR-6220 An Assessment of Fire Vulnerability for Aged Electrical Relays
NUREG/CR-6230 Radioanalytical Technology for 10 CFR Part 61 and Other Selected Radionuclides: Literature Review
NUREG/CR-6241 Technical Guidelines for Aseismic Design of Nuclear Power Plants
NUREG/CR-6265 Multidisciplinary Framework for Human Reliability Analysis with an Application to Errors of Commission and Dependency
NUREG/CR-6268 Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding
NUREG/CR-6275 Mechanical Properties of Thermally Aged Cast Stainless Steels from Shippingport Reactor Components
NUREG/CR-6303 Method for Performing Diversity and Defense-in-Depth Analyses of Reactor Protection Systems
NUREG/CR-6314 Quality Assurance Inspections for Shipping and Storage Containers
NUREG/CR-6342 Fracture Toughness Testing With Cracked Round Bars: Feasibility Study
NUREG/CR-6345 Radiation Dose Estimates for Radiopharmaceuticals
NUREG/CR-6346 Hydrologic Evaluation Methodology for Estimating Water Movement Through the Unsaturated Zone at Commercial Low-Level Radioactive Waste Disposal Sites
NUREG/CR-6350 A Technique for Human Error Analysis (ATHEANA)
NUREG/CR-6358 Assessment of United States Industry Structural Codes and Standards for Application to Advanced Nuclear Power Reactors
NUREG/CR-6369 Drywell Debris Transport Study
NUREG/CR-6372 Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts
NUREG/CR-6384 Literature Review of Environmental Qualification of Safety-Related Electric Cables
NUREG/CR-6406 Environmental Testing of an Experimental Digital Safety Channel
NUREG/CR-6407 Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety
NUREG/CR-6410 Nuclear Fuel Cycle Facility Accident Analysis Handbook
NUREG/CR-6420 Self-Monitoring Surveillance System for Prestressing Tendons
NUREG/CR-6421 A Proposed Acceptance Process for Commercial Off-the-Shelf (COTS) Software in Reactor Applications
NUREG/CR-6424 Report on Aging of Nuclear Power Plant Reinforced Concrete Structures
NUREG/CR-6427 Assessment of the DCH Issue for Plants with Ice Condenser Containments
NUREG/CR-6428 Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds
NUREG/CR-6431 Recommended Electromagnetic Operating Envelopes for Safety-Related I&C Systems in Nuclear Power Plants
NUREG/CR-6441 Analysis of Spent Fuel Heatup Following Loss of Water in a Spent Fuel Pool
NUREG/CR-6463 Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems
NUREG/CR-6471 Characterization of Flaws in U.S. Reactor Pressure Vessels
NUREG/CR-6476 Circuit Bridging of Components by Smoke
NUREG/CR-6477 Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities
NUREG/CR-6479 Technical Basis for Environmental Qualification of Microprocessor-Based Safety-Related Equipment in Nuclear Power Plants
NUREG/CR-6500 Owners of Nuclear Power Plants
NUREG/CR-6509 The Effect of Initial Temperature on Flame Acceleration and Deflagration-to-Detonation Transition Phenomenon
NUREG/CR-6511 Steam Generator Tube Integrity Program Annual Report
NUREG/CR-6524 The Effect of Lateral Venting on Deflagration-to-Detonation Transition in Hydrogen-Air-Steam Mixtures at Various Initial Temperatures
NUREG/CR-6525 SECPOP2000: Sector Population, Land Fraction, and Economic Estimation Program
NUREG/CR-6530 Deliberate Ignition of Hydrogen-Air-Steam Mixtures in Condensing Steam Environments
NUREG/CR-6534 FRAPCON-3 Updates, Including Mixed-Oxide Fuel Properties
NUREG/CR-6543 Effects of Smoke on Functional Circuits
NUREG/CR-6544 A Methodology for Analyzing Precursors to Earthquake-Initiated and Fire-Initiated Accident Sequences
NUREG/CR-6554 Finite Element Analyses for Seismic Shear Wall International Standard Problem
NUREG/CR-6559 Large-Scale Vibration Tests of Main Steam and Feedwater Piping Systems With Conventional and Energy-Absorbing Supports
NUREG/CR-6565 Uncertainty Analyses of Infiltration and Subsurface Flow and Transport for SDMP Sites
NUREG/CR-6567 Low-Level Radioactive Waste Classification, Characterization, and Assessment: Waste Streams and Neutron-Activated Metals
NUREG/CR-6572 Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA: Procedure Guides for a Probabilistic Risk Assessment (English Version)
NUREG/CR-6577 U.S. Nuclear Power Plant Operating Cost and Experience Summaries
NUREG/CR-6583 Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels
NUREG/CR-6584 Evaluation of the Hualien Quarter Scale Model Seismic Experiment
NUREG/CR-6589 The Effects of Surface Condition on an Ultrasonic Inspection: Engineering Studies Using Validated Computer Model
NUREG/CR-6595 An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events
NUREG/CR-6597 Results and Insights on the Impact of Smoke on Digital Instrumentation and Control
NUREG/CR-6607 Guidance for Performing Probabilistic Seismic Hazard Analysis for a Nuclear Plant Site: Example Application to the Southeastern United States
NUREG/CR-6609 Comparison of Irradiation-Induced Shifts of KJc and Charpy Impact Toughness for Reactor Pressure Vessel Steels
NUREG/CR-6622 Probabilistic Liquefaction Analysis
NUREG/CR-6623 Vapor Explosions in a One-Dimensional Large Scale Geometry with Simulant Melts
NUREG/CR-6624 Recommendations for Revision of Regulatory Guide 1.78
NUREG/CR-6625 Automated Seismic Event Monitoring System
NUREG/CR-6627 The Role of Organic Complexants and Colloids in the Transport of Radionuclides by Groundwater
NUREG/CR-6628 The Effects of Aging at 343°C on the Microstructure and Mechanical Properties of Type 308 Stainless Steel Weldments
NUREG/CR-6629 Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld
NUREG/CR-6632 Solubility and Leaching of Radionuclides in Site Decommissioning Management Plan (SDMP) Slags
NUREG/CR-6633 Advanced Information Systems Design: Technical Basis and Human Factors Review Guidance
NUREG/CR-6634 Computer-Based Procedure Systems: Technical Basis and Human Factors Review Guidance
NUREG/CR-6635 Soft Controls: Technical Basis and Human Factors Review Guidance
NUREG/CR-6636 Maintainability of Digital Systems: Technical Basis and Human Factors Review Guidance
NUREG/CR-6637 Human Systems Interface and Plant Modernization Process: Technical Basis and Human Factors Review Guidance
NUREG/CR-6638 Advanced NDE for Steam Generator Tubing
NUREG/CR-6647 Adsorption and Desorption Behavior of Selected 10 CFR Part 61 Radionuclides From Ion Exchange Resin by Waters of Different Chemical Composition
NUREG/CR-6648 Environmental Assessment: San Bernadino National Wildlife Refuge Well 10
NUREG/CR-6650 PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices
NUREG/CR-6651 International Comparative Assessment Study of Pressurized Thermal Shock in Reactor Pressure Vessels
NUREG/CR-6653 Comparison of Estimated Ground-Water Recharge Using Different Temporal Scales of Field Data
NUREG/CR-6654 A Study of Air-Operated Valves in U.S. Nuclear Power Plants
NUREG/CR-6655 Sensitivity and Uncertainty Analyses Applied to Criticality Safety Validation
NUREG/CR-6656 Information on Hydrologic Conceptual Models, Parameters, Uncertainty Analysis, and Data Sources for Dose Assessments at Decommissioning Sites
NUREG/CR-6658 TRAC-M Programmer's Guide: Fortran 77 Version 5.5
NUREG/CR-6662 KENO3D Visualization Tool for KENO V.a and KENO-VI Geometry Models
NUREG/CR-6664 Pressure and Leak-Rate Tests and Models for Predicting Failure of Flawed Steam Generator Tubes
NUREG/CR-6666 Survey of Waste Solidification Process Technologies
NUREG/CR-6668 Standard Review Plan for Training and Qualifications Plans for Security Personnel at Category I Fuel Facilities
NUREG/CR-6669 Evaluation of Terminated Licenses Parts 30, 40, and 70:  The Terminated License Tracking System
NUREG/CR-6672 Reexamination of Spent Fuel Shipment Risk Estimates
NUREG/CR-6673 Hydrogen Generation in TRU Waste Transportation Packages
NUREG/CR-6675 Interaction of Zinc Vapor with Zircaloy and the Effect of Zinc Vapor on the Mechanical Properties of Zircaloy
NUREG/CR-6676 Probabilistic Dose Analysis Using Parameter Distributions Developed For RESRAD and RESRAD-BUILD Codes
NUREG/CR-6677 Evaluation Of Risk Associated With Intergranular Stress Corrosion Cracking In Boiling Water Reactor Internals
NUREG/CR-6678 Pretest Round Robin Analysis of a Prestressed Concrete Containment Vessel Model
NUREG/CR-6679 Assessment of Age-Related Degradation of Structures and Passive Components for U.S. Nuclear Power Plants
NUREG/CR-6680 Review Templates for Computer-Based Reactor Protection Systems
NUREG/CR-6681 Ampacity Derating and Cable Functionality for Raceway Fire Barriers
NUREG/CR-6682 Summary and Categorization of Public Comments on Controlling the Disposition of Solid Materials
NUREG/CR-6683 A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage
NUREG/CR-6684 Advanced Alarm Systems: Revision of Guidance and Its Technical Basis
NUREG/CR-6685 Pretest Analysis of a 1:4-Scale Prestressed Concrete Containment Vessel Model
NUREG/CR-6686 Experience With the Scale Criticality Safety Cross-Section Libraries
NUREG/CR-6687 Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys
NUREG/CR-6688 Testing, Verifying, and Validating SAPHIRE Versions 6.0 and 7.0
NUREG/CR-6689 Proposed Approach for Reviewing Changes to Risk-Important Human Actions
NUREG/CR-6690 The Effects of Interface Management Tasks on Crew Performance and Safety in Complex, Computer-Based Systems: Overview and Main Findings
NUREG/CR-6691 The Effects of Alarm Display, Processing, and Availability on Crew Performance
NUREG/CR-6692 Probabilistic Modules for the RESRAD and RESRAD-BUILD Computer Codes: User Guide
NUREG/CR-6694 POLIDENT: A Module for Generating Continuous-Energy Cross Sections From ENDF Resonance Data
NUREG/CR-6695 Hydrologic Uncertainty Assessment for Decommissioning Sites: Hypothetical Test Case Applications
NUREG/CR-6696 LAPUR 5.2 Verification and User's Manual
NUREG/CR-6697 Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes
NUREG/CR-6699 A Review of Large-Scale Fracture Experiments Relevant to Pressure Vessel Integrity Under Pressurized Thermal Shock Conditions
NUREG/CR-6700 Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel
NUREG/CR-6701 Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel
NUREG/CR-6702 Limited Burnup Credit in Criticality Safety Analysis: A Comparison of ISG-8 and Current International Practice
NUREG/CR-6703 Environmental Effects of Extending Fuel Burnup Above 60 Gwd/MTU
NUREG/CR-6704 Assessment of Environmental Qualification Practices and Condition Monitoring Techniques for Low-Voltage Electric Cables
NUREG/CR-6705 Historical Case Analysis of Uranium Plume Attenuation
NUREG/CR-6706 Capacity of Steel & Concrete Containment Vessels with Corrosion Damage
NUREG/CR-6707 Seismic Analysis of a Reinforced Concrete Containment Vessel Model
NUREG/CR-6708 Surface Complexation Modeling of Uranium (VI) Adsorption on Natural Mineral Assemblages
NUREG/CR-6710 Extending the Dynamic Flowgraph Methodology (DFM) to Model Human Performance and Team Effects
NUREG/CR-6711 Environmental Assessment of Major Revision of 10 CFR Part 71
NUREG/CR-6712 Summary and Categorization of Public Comments on the Major Revision of 10 CFR Part 71
NUREG/CR-6713 Regulatory Analysis of Major Revision of 10 CFR Part 71
NUREG/CR-6714 Hanford Tank Waste Remediation System Pretreatment Chemistry and Technology
NUREG/CR-6715 Probability-Based Evaluation of Degraded Reinforced Concrete Components in Nuclear Power Plants
NUREG/CR-6716 Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks
NUREG/CR-6717 Environmental Effects on Fatigue Crack Initiation in Piping and Pressure Vessel Steels
NUREG/CR-6718 OPUS/PlotOPUS: An ORIGEN-S Post-Processing Utility and Plotting Program for SCALE
NUREG/CR-6719 Assessment of the Relevance of Displacement Bases Design Methods/Criteria to Nuclear Plant Structures
NUREG/CR-6720 TRAC-M Validation Test Matrix
NUREG/CR-6721 Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds
NUREG/CR-6722 TRAC-M/FORTRAN 90 (Version 3.0) User's Manual
NUREG/CR-6724 TRAC-M/FORTRAN 90 (Version 3.0) Theory Manual
NUREG/CR-6725 TRAC-M/FORTRAN 90 (Version 3.0) Programmer's Manual
NUREG/CR-6728 Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-consistent Ground Motion Spectra Guidelines
NUREG/CR-6729 Field Studies for Estimating Uncertainties in Ground-Water Recharge Using Near-Continuous Peizometer Data
NUREG/CR-6730 TRAC-M/F77, Version 5.5 Developmental Assessment Manual
NUREG/CR-6732 Zinc-Zircaloy Interaction in Dry Storage Casks
NUREG/CR-6733 A Baseline Risk-Informed, Performance-Based Approach for In Situ Leach Uranium Extraction Licensees
NUREG/CR-6734 Digital Systems Software Requirements Guidelines
NUREG/CR-6735 Effects of Deregulation on Safety:  Implications Drawn From The Aviation, Rail and United Kingdom Nuclear Power Industries
NUREG/CR-6737 Bases for Predicting the Earliest Penetrations Due to SCC for Alloy 600 on the Secondary Side of PWR Steam Generators
NUREG/CR-6738 Risk Methods Insights Gained from Fire Incidents
NUREG/CR-6739 FRAPTRAN: NRC's Computer Code
NUREG/CR-6742 Phenomenon Identification and Ranking Tables (PIRTs) for Rod Ejection Accidents in Pressurized Water Reactors Containing High Burnup Fuel
NUREG/CR-6743 Phenomenon Identification and Ranking Tables (PIRTs) for Power Oscillations Without Scram in Boiling Water Reactors Containing High Burnup Fuel
NUREG/CR-6744 Phenomenon Identification and Ranking Tables (PIRTs) for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High Burnup Fuel
NUREG/CR-6745 Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination
NUREG/CR-6746 Advanced Nondestructive Evaluation for Steam Generator Tubing
NUREG/CR-6747 Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit
NUREG/CR-6748 STARBUCS: A Prototypic SCALE Control Module for Automated Criticality Safety Analyses Using Burnup Credit
NUREG/CR-6749 Integrating Digital and Conventional Human-System Interfaces: Lessons Learned from a Control Room Modernization Program
NUREG/CR-6750 Performance of MOV Stem Lubricants at Elevated Temperature
NUREG/CR-6751 The Human Performance Evaluation Process: A Resource for Reviewing the Identification and Resolution of Human Performance Problems
NUREG/CR-6752 A Comparative Analysis of Special Treatment Requirements for Systems, Structures, and Components (SSCs) of Nuclear Power Plants With Commercial Requirements of Non-Nuclear Power Plants
NUREG/CR-6753 Review of Findings for Human Performance Contribution to Risk in Operating Events
NUREG/CR-6754 Review of Industry Responses to NRC Generic Letter 97-06 on Degradation of Steam Generator Internals
NUREG/CR-6755 Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario Using the Probabilistic RESRAD-BUILD 3.0 Code
NUREG/CR-6756 Analysis of Potential for Jet-Impingement Erosion from Leaking Steam Generator Tubes During Severe Accidents
NUREG/CR-6757 Large-Scale Molecular Dynamics Simulations of Metal Sorption onto the Basal Surfaces of Clay Minerals
NUREG/CR-6758 Radionuclide-Chelating Agent Complexes in Low-Level Radioactive Decontamination Waste: Stability, Adsorbtion, and Transport Potential
NUREG/CR-6759 Parametric Study of Effect of Control Rods for PWR Burnup Credit
NUREG/CR-6760 Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit
NUREG/CR-6761 Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit
NUREG/CR-6762 Generic-Safety-Issue (GSI) 191 Technical Assessment
NUREG/CR-6763 Aging Assessment of Safety-Related Fuses Used in Low- and Medium- Voltage Applications in Nuclear Power Plants
NUREG/CR-6764 Burnup Credit PIRT Report
NUREG/CR-6765 Development of Technical Basis for Leak-Before-Break Evaluation Procedures
NUREG/CR-6766 Release of Radionuclides and Chelating Agents from Full-System Decontamination Ion-Exchange Resins
NUREG/CR-6767 Evaluation of Hydrologic Uncertainty Assessments for Decommissioning Sites Using Complex and Simplified Models
NUREG/CR-6768 Spent Nuclear Fuel Transportation Package Performance Study Issues Report
NUREG/CR-6769 Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Development of Hazard- & Risk-Consistent Seismic Spectra for Two Sites
NUREG/CR-6770 GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System & Containments to Selected Accident Sequences
NUREG/CR-6771 GSI-191: The Impact of Debris Induced Loss of ECCS Recirculation on PWR Core Damage Frequency
NUREG/CR-6772 GSI-191: Separate-Effects Characterization of Debris Transport in Water
NUREG/CR-6773 GSI-191: Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries
NUREG/CR-6774 Validation on Failure & Leak-Rate Correlations for Stress Corrosion Cracks in Steam Generator Tubes
NUREG/CR-6775 Human Performance Characterization in the Reactor Oversight Process
NUREG/CR-6776 Cable Insulation Resistance Measurements Made During Cable Fire Tests
NUREG/CR-6777 Results and Analysis of The ASTM Round Robin On Reconstitution
NUREG/CR-6778 The Effects of Composition and Heat Treatment on Hardening and Embrittlement of Reactor Pressure Vessel Steels
NUREG/CR-6780 Effects of Adsorption Constant Uncertainty on Containment Plume Migration: One- and Two-Dimensional Numerical Studies
NUREG/CR-6781 Recommendations on the Credit for Cooling Time in PWR Burnup Credit Analyses
NUREG/CR-6782 Comparison of U.S. Military and International Electromagnetic Compatibility Guidance
NUREG/CR-6783 Structural Seismic Fragility Analysis of the Surry Containment
NUREG/CR-6784 Use of Computerized Microtomography to Examine the Relationships of Sorption Sites in Alluvial Soils to Iron and Pore Space Distributions
NUREG/CR-6785 Evaluation of Eddy Current Reliability from Steam Generator Mock-Up Round-Robin
NUREG/CR-6786 ANL/CANTIA: A Computer Code for Steam Generator Integrity Assessments
NUREG/CR-6787 Mechanism and Estimation of Fatigue Crack Initiation in Austenitic Stainless Steels in LWR Environments
NUREG/CR-6788 Evaluation of Aging and Qualification Practices for Cable Splices Used in Nuclear Plants
NUREG/CR-6789 Results From Pressure and Leak-Rate Testing of Laboratory-Degraded Steam Generator Tubing
NUREG/CR-6791 Eddy Current Reliability Results from the Steam Generator Mock-up Analysis Round-Robin: Revision 1
NUREG/CR-6792 Behavior of PWR Reactor Coolant System Components, Other than Steam Generator Tubes, Under Severe Accident Conditions
NUREG/CR-6793 Numerical Simulation of the Howard Street Tunnel Fire, Baltimore, Maryland, July 2001
NUREG/CR-6794 Evaluation of Aging and Environmental Qualification Practices for Power Cables Used in Nuclear Power Plants
NUREG/CR-6795 A Comparison of Three Round Robin Studies on ISI Reliability of Wrought Stainless Steel Piping
NUREG/CR-6798 Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the Takahama-3 Reactor
NUREG/CR-6799 Analysis of Rail Car Components Exposed to a Tunnel Fire Environment
NUREG/CR-6800 Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs
NUREG/CR-6801 Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses
NUREG/CR-6802 Recommendations for Shielding Evaluations for Transport & Storage Packages
NUREG/CR-6804 Second U.S. Nuclear Regulatory Commission International Steam Generator Tube Integrity Research Program
NUREG/CR-6805 A Comprehensive Strategy of Hydrogeologic Modeling and Uncertainty Analysis for Nuclear Facilities and Sites
NUREG/CR-6806 MOV Stem Lubricant Aging Research
NUREG/CR-6807 Results of NRC-Sponsored Stellite 6 Aging & Friction Testing
NUREG/CR-6808 Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance
NUREG/CR-6809 Posttest Analysis of the NUPEC/NRC 1:4 Scale Prestressed Concrete Containment Vessel Model
NUREG/CR-6810 Overpressurization Test of a 1:4-Scale Prestressed Concrete Containment Vessel Model
NUREG/CR-6811 Strategies for Application of Isotopic Uncertainties in Burnup Credit
NUREG/CR-6812 Emerging Technologies in Instrumentation and Controls
NUREG/CR-6813 Issues and Recommendations for Advancement of PRA Technology In Risk-Informed Decision Making
NUREG/CR-6814 Final Report on Advanced Nondestructive Evaluation for Steam Generator Tubing for the Second International Steam Generator Tube Integrity Program
NUREG/CR-6815 Review of the Margins for ASME Code Fatigue Design Curve: Effects of Surface Roughness and Material Variability
NUREG/CR-6816 Review and Assessment of Codes and Procedures for HTGR Components
NUREG/CR-6817 A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code
NUREG/CR-6818 Drop Test Results for the Combustion Engineering Model No. ABB-2901 Fuel Pellet Shipping Package
NUREG/CR-6819 Common-Cause Failure Event Insights
NUREG/CR-6820 Application of Surface Complexation Modeling to Describe Uranium (VI) Adsorption and Retardation at the Uranium Mill Tailings Site at Naturita, Colorado
NUREG/CR-6821 Solubility and Leaching of Radionuclides in Site Decommissioning Management Plan (SDMP) Soil and Ponded Wastes
NUREG/CR-6822 Collaborative Study of NUPEC Seismic Field Test Data for NPP Structures
NUREG/CR-6823 Handbook of Parameter Estimation for Probabilistic Risk Assessment
NUREG/CR-6824 Materials Behavior in HTGR Environments
NUREG/CR-6825 Literature Review and Assessment of Plant and Animal Transfer Factors Used in Performance Assessment Modeling
NUREG/CR-6826 Fracture Toughness and Crack Growth Rates of Irradiated Austenitic Stainless Steels
NUREG/CR-6831 Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage
NUREG/CR-6832 Regulatory Effectiveness of Unresolved Safety Issue (USI) A-45, "Shutdown Decay Heat Removal Requirements"
NUREG/CR-6833 Formal Methods of Decision Analysis Applied to Prioritization of Research and Other Topics
NUREG/CR-6834 Circuit Analysis: Failure Mode and Likelihood Analysis
NUREG/CR-6835 Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks
NUREG/CR-6836 Comparing Ground-Water Recharge Estimates Using Advanced Monitoring Techniques and Models
NUREG/CR-6837 The Battelle Integrity of Nuclear Piping (BINP) Program Final Report
NUREG/CR-6838 Technical Basis for Regulatory Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m)
NUREG/CR-6839 Fort Saint Vrain Gas Cooled Reactor Operational Experience
NUREG/CR-6840 The Technical Basis for the NRC's Guidelines for External Risk Communication
NUREG/CR-6841 A Risk-Informed Basis for Establishing Non-Fixed Surface Contamination Limits for Spent Fuel Transportation Casks
NUREG/CR-6842 Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants
NUREG/CR-6843 Combined Estimation of Hydrogeologic Conceptual Model and Parameter Uncertainty
NUREG/CR-6844 TRISO-Coated Particle Fuel Phenomenon Identification and Ranking Tables (PIRTs) for Fission Product Transport Due to Manufacturing, Operations, and Accidents
NUREG/CR-6845 Sensitivity Analysis Applied to the Validation of the 10B Capture Reaction in Nuclear Fuel Casks
NUREG/CR-6846 Air Oxidation Kinetics for Zr-Based Alloys
NUREG/CR-6848 Preliminary Validation of a Methodology for Assessing Software Quality
NUREG/CR-6849 Analysis of In-Vessel Retention and Ex-Vessel Fuel Coolant Interaction for AP1000
NUREG/CR-6850 EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities
NUREG/CR-6851 Hydrogen Effects on Air Oxidation of Zirlo Alloy
NUREG/CR-6853 Comparison of Average Transport and Dispersion Among a Gaussian, a Two-Dimensional, and a Three-Dimensional Model
NUREG/CR-6854 Fracture Analysis of Vessels — Oak Ridge FAVOR v04.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations
NUREG/CR-6855 Fracture Analysis of Vessels — Oak Ridge FAVOR, V04.1, Computer Code: User’s Guide
NUREG/CR-6859 PRA Procedures and Uncertainty for PTS Analysis
NUREG/CR-6860 An Assessment of Visual Testing
NUREG/CR-6861 Barrier Integrity Research Program
NUREG/CR-6863 Development of Evacuation Time Estimate Studies for Nuclear Power Plants
NUREG/CR-6864 Identification and Analysis of Factors Affecting Emergency Evacuations
NUREG/CR-6865 Parametric Evaluation of Seismic Behavior of Freestanding Spent Fuel Dry Cask Storage Systems
NUREG/CR-6866 Technical Basis for Regulatory Guidance on Lightning Protection in Nuclear Power Plants
NUREG/CR-6868 Small-Scale Experiments: Effects of Chemical Reactions on Debris-Bed Head Loss — A Subtask of GSI-191
NUREG/CR-6869 A Reliability Physics Model for Aging of Cable Insulation Materials
NUREG/CR-6870 Consideration of Geochemical Issues in Groundwater Restoration at Uranium In-Situ Leach Mining Facilities
NUREG/CR-6871 Documentation and Applications of the Reactive Geochemical Transport Model RATEQ
NUREG/CR-6873 Corrosion Rate Measurements and Chemical Speciation of Corrosion Products Using Thermodynamic Modeling of Debris Components to Support GSI-191
NUREG/CR-6874 GSI-191: Experimental Studies of Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Loss with Emphasis on the Effects of Calcium Silicate Insulation
NUREG/CR-6875 Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials
NUREG/CR-6876 Risk-Informed Assessment of Degraded Buried Piping Systems in Nuclear Power Plants
NUREG/CR-6877 Characterization and Head-Loss Testing of Latent Debris from Pressurized-Water-Reactor Containment Buildings
NUREG/CR-6878 Effect of Material Heat Treatment on Fatigue Crack Initiation in Austenitic Stainless Steels in LWR Environments
NUREG/CR-6879 Steam Generator Tube Integrity Issues: Pressurization Rate Effects, Failure Maps, Leak Rate Correlation Models, and Leak Rates in Restricted Areas
NUREG/CR-6880 Argonne Model Boiler Facility: Topical Report
NUREG/CR-6881 Soil and Groundwater Sample Characterization and Agricultural Practices for Assessing Food Chain Pathways in Biosphere Models
NUREG/CR-6882 Assessment of Wireless Technologies and Their Application at Nuclear Facilities
NUREG/CR-6883 The SPAR-H Human Reliability Analysis Method
NUREG/CR-6884 Model Abstraction Techniques for Soil-Water Flow and Transport
NUREG/CR-6885 Screen Penetration Test Report
NUREG/CR-6886 Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario
NUREG/CR-6887 DORT/TORT Analysis of the Hatch Unit-1 Jet Pump Riser Brace Pad Neutron Dosimetry Measurements with Comparisons to Predictions Made with RAMA
NUREG/CR-6888 Emerging Technologies in Instrumentation and Controls: An Update
NUREG/CR-6889 Seismic Analysis of Simplified Piping Systems for the NUPEC Ultimate Strength Piping Test Program
NUREG/CR-6890 Reevaluation of Station Blackout Risk at Nuclear Power Plants
NUREG/CR-6891 Crack Growth Rates of Irradiated Austenitic Stainless Steel Weld Heat Affected Zone in BWR Environments
NUREG/CR-6892 Irradiation-Assisted Stress Corrosion Cracking Behavior of Austenitic Stainless Steels Applicable to LWR Core Internals
NUREG/CR-6893 Modeling Adsorption Processes: Issues in Uncertainty, Scaling, and Prediction
NUREG/CR-6894 Spent Fuel Transportation Package Response to the Caldecott Tunnel Fire Scenario
NUREG/CR-6895 Technical Review of On-Line Monitoring Techniques for Performance Assessment
NUREG/CR-6896 Assessment of Seismic Analysis Methodologies for Deeply Embedded Nuclear Power Plant Structures
NUREG/CR-6897 Assessment of Void Swelling in Austenitic Stainless Steel Core Internals
NUREG/CR-6898 A Combined Analytical Study to Characterize Uranium Soil and Sediment Contamination: The Case of the Naturita UMTRA Site and the Role of Grain Coatings
NUREG/CR-6900 The Effect of Elevated Temperature on Concrete Materials and Structures — A Literature Review
NUREG/CR-6901 Current State of Reliability Modeling Methodologies for Digital Systems and Their Acceptance Criteria for Nuclear Power Plant Assessments
NUREG/CR-6902 Effects of Insulation Debris on Throttle-Valve Flow Performance: A Subtask of GSI-191
NUREG/CR-6903 Human Event Repository and Analysis (HERA) System, Overview
NUREG/CR-6904 Evaluation of the Broadband Impedance Spectroscopy Prognostic/Diagnostic Technique for Electric Cables Used in Nuclear Power Plants
NUREG/CR-6905 Report of Experimental Results for the International Fire Model Benchmarking and Validation Exercise #3
NUREG/CR-6906 Containment Integrity Research at Sandia National Laboratories - An Overview
NUREG/CR-6907 Crack Growth Rates of Nickel Alloy Welds in a PWR Environment
NUREG/CR-6909 Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials
NUREG/CR-6910 Alternative Conceptual Models for Assessing Food Chain Pathways in Biosphere Models
NUREG/CR-6911 Tests of Uranium (VI) Adsorption Models in a Field Setting
NUREG/CR-6912 GSI-191 PWR Sump Screen Blockage Chemical Effects Tests: Thermodynamic Simulations
NUREG/CR-6913 Chemical Effects Head-Loss Research In Support of Generic Safety Issue 191
NUREG/CR-6914 Integrated Chemical Effects Test Project
NUREG/CR-6915 Aluminum Chemistry in a Prototypical Post-Loss-of-Coolant-Accident, Pressurized-Water-Reactor Containment Environment
NUREG/CR-6916 Hydraulic Transport of Coating Debris
NUREG/CR-6917 Experimental Measurements of Pressure Drop Across Sump Screen Debris Beds in Support of Generic Safety Issue 191
NUREG/CR-6918 VARSKIN 4: A Computer Code for Assessing Skin Dose from Skin Contamination
NUREG/CR-6919 Recommendations for Revision of Seismic Damping Values in Regulatory Guide 1.61
NUREG/CR-6920 Risk-Informed Assessment of Degraded Containment Vessels
NUREG/CR-6921 Crack Growth Rates in a PWR Environment of Nickel Alloys from the Davis-Besse and V.C. Summer Power Plants
NUREG/CR-6922 P-CARES: Probabilistic Computer Analysis for Rapid Evaluation of Structures
NUREG/CR-6923 Expert Panel Report on Proactive Materials Degradation Assessment
NUREG/CR-6924 Non-destructive and Failure Evaluation of Tubing from a Retired Steam Generator
NUREG/CR-6925 Assessment of Analysis Methods for Seismic Shear Wall Capacity Using JNES/NUPEC Multi-Axial Cyclic and Shaking Table Test Data
NUREG/CR-6926 Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for Application to Nuclear Power Plants
NUREG/CR-6927 Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors
NUREG/CR-6928 Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants
NUREG/CR-6929 Assessment of Eddy Current Testing for the Detection of Cracks in Cast Stainless Steel Reactor Piping Components
NUREG/CR-6930 Temperature Dependence of Weibull Stress Parameters: Studies Using the Euro-Material Similar to ASME A508 Class-3 Steel
NUREG/CR-6931 Carolfire Test Report
NUREG/CR-6932 Baseline Risk Index for Initiating Events (BRIIE)
NUREG/CR-6933 Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods
NUREG/CR-6934 Fatigue Crack Flaw Tolerance in Nuclear Power Plant Piping - A Basis for Improvements to ASME Code Section XI Appendix L
NUREG/CR-6935 Sensitivity Studies of Failure of Steam Generator Tubes during Main Steam Line Break and Other Secondary Side Depressurization Events
NUREG/CR-6936 Probabilities of Failure and Uncertainty Estimate Information for Passive Components – A Literature Review
NUREG/CR-6938 Final Report-Assessment of Potential Phosphate Ion-Cenmentitious Materials Interactions
NUREG/CR-6939 Coexistence Assessment of Industrial Wireless Protocols in the Nuclear Facility Environment
NUREG/CR-6940 Combined Estimation of Hydrogeologic Conceptual Model, Parameter, and Scenario Uncertainty with Application to Uranium Transport at the Hanford Site 300 Area
NUREG/CR-6941 Soil-to-Plant Concentration Ratios for Assessing Food-Chain Pathways in Biosphere Models
NUREG/CR-6942 Dynamic Reliability Modeling of Digital Instrumentation and Control Systems for Nuclear Reactor Probabilistic Risk Assessments
NUREG/CR-6943 A Study of Remote Visual Methods to Detect Cracking in Reactor Components
NUREG/CR-6944 Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs)
NUREG/CR-6945 Fabrication Flaw Density and Distribution in Repairs to Reactor Pressure Vessel and Piping Welds
NUREG/CR-6946 Field Studies to Confirm Uncertainty Estimates of Ground-Water Recharge
NUREG/CR-6947 Human Factors Considerations with Respect to Emerging Technology in Nuclear Power Plants
NUREG/CR-6948 Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities and Sites: Logic, Strategic Approach and Discussion
NUREG/CR-6949 The Employment of Empirical Data and Bayesian Methods in Human Reliability Analysis: A Feasibility Study
NUREG/CR-6951 Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit
NUREG/CR-6952 Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE)
NUREG/CR-6953 Review of NUREG-0654, Supplement 3, "Criteria for Protective Action Recommendations for Severe Accidents"
NUREG/CR-6954 Fracture Analysis of Vessels - Oak Ridge FAVOR, v04.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations
NUREG/CR-6955 Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask
NUREG/CR-6956 Nonlinear Analyses for Embedded Cracks Under Pressurized Thermal Shock: Comparisons with FAVOR and Weibull Stress Approaches
NUREG/CR-6957 Correlation Analysis of JNES Seismic Wall Pressure Data for ABWR Model Structures
NUREG/CR-6958 LAPUR 6.0 Manual
NUREG/CR-6959 Application of Surface Complexation Modeling to Selected Radionuclides and Aquifer Sediments
NUREG/CR-6960 Crack Growth Rates and Fracture Toughness of Irradiated Austenitic Stainless Steels in BWR Environments
NUREG/CR-6962 Traditional Probabilistic Risk Assessment Methods for Digital Systems
NUREG/CR-6963 An Assessment of PWR Steam Generator Condensation at the Oregon State University APEX Facility
NUREG/CR-6964 Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments
NUREG/CR-6965 Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels and Alloy 690 from Halden Phase-II Irradiations
NUREG/CR-6966 Tsunami Hazard Assessment at Nuclear Power Plant Sites in the United States of America
NUREG/CR-6967 Cladding Embrittlement During Postulated Loss-of-Coolant Accidents
NUREG/CR-6968 Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation – Calvert Cliffs, Takahama, and Three Mile Island Reactors
NUREG/CR-6969 Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation — ARIANE and REBUS Programs
NUREG/CR-6971 Spent Fuel Decay Heat Measurements Performed at the Swedish Central Interim Storage Facility
NUREG/CR-6972 Validation of SCALE 5 Decay Heat Predictions for LWR Spent Nuclear Fuel
NUREG/CR-6973 Technical Basis for Assessing Uranium Bioremediation Performance
NUREG/CR-6974 Symbolic Nuclear Analysis Package (SNAP): Common Application Framework for Engineering Analysis (CAFEAN) Preprocessor Plug-in Application Programming Interface
NUREG/CR-6975 Rod Bundle Heat Transfer Test Facility Test Plan and Design
NUREG/CR-6976 Rod Bundle Heat Transfer Test Facility Description
NUREG/CR-6977 Redox and Sorption Reactions of Iodine and Cesium During Transport Through Aquifer Sediments
NUREG/CR-6978 A Phenomena Identification and Ranking Table (PIRT) Exercise for Nuclear Power Plant Fire Modeling Applications
NUREG/CR-6979 Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data
NUREG/CR-6980 RBHT Reflood Heat Transfer Experiments Data and Analysis
NUREG/CR-6981 Assessment of Emergency Response Planning and Implementation for Large Scale Evacuations
NUREG/CR-6982 Assessment of Noise Level for Eddy Current Inspection of Steam Generator Tubes
NUREG/CR-6983 Seismic Analysis of Large-Scale Piping Systems for the JNES-NUPEC Ultimate Strength Piping Test Program
NUREG/CR-6984 Field Evaluation of Low-Frequency SAFT-UT on Cast Stainless Steel and Dissimilar Metal Weld Components
NUREG/CR-6985 A Benchmark Implementation of Two Dynamic Methodologies for the Reliability Modeling of Digital Instrumentation and Control Systems
NUREG/CR-6986 Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs
NUREG/CR-6987 Analysis of Structural Materials Exposed to a Severe Fire Environment
NUREG/CR-6988 Final Report — Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant
NUREG/CR-6989 Methodology for Estimating Fabrication Flaw Density and Distribution – Reactor Pressure Vessel Welds
NUREG/CR-6990 Development and Testing of ENDF/B-VI.8 and ENDF/B-VII.0 Coupled Neutron-Gamma Libraries for SCALE 6
NUREG/CR-6991 Design Practices for Communications and Workstations in Highly Integrated Control Rooms
NUREG/CR-6992 Instrumentation and Controls in Nuclear Power Plants: An Emerging Technologies Update
NUREG/CR-6994 Argonne Model Boiler Test Results
NUREG/CR-6995 SCDAP/RELAP5 Thermal-Hydraulic Evaluations of the Potential for Containment Bypass During Extended Station Blackout Severe Accident Sequences in a Westinghouse Four-Loop PWR
NUREG/CR-6996 Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechanism Penetrations
NUREG/CR-6997 Modeling a Digital Feedwater Control System Using Traditional Probabilistic Risk Assessment Methods
NUREG/CR-6998 Review of Information for Spent Nuclear Fuel Burnup Confirmation
NUREG/CR-6999 Technical Basis for a Proposed Expansion of Regulatory Guide 3.54 — Decay Heat Generation in an Independent Spent Fuel Storage Installation
NUREG/CR-7000 Essential Elements of an Electric Cable Condition Monitoring Program
NUREG/CR-7001 Predictive Bias and Sensitivity in NRC Fuel Performance Codes
NUREG/CR-7002 Criteria for Development of Evacuation Time Estimate Studies
NUREG/CR-7003 Background and Derivation of ANS-5.4 Standard Fission Product Release Model
NUREG/CR-7004 Technical Basis for Regulatory Guidance on Design-Basis Hurricane-Borne Missile Speeds for Nuclear Power Plants
NUREG/CR-7005 Technical Basis for Regulatory Guidance on Design-Basis Hurricane Wind Speeds for Nuclear Power Plants
NUREG/CR-7006 Guidelines for Field-Programmable Gate Arrays in Nuclear Power Plant Safety Systems Plant
NUREG/CR-7007 Diversity Strategies for Nuclear Power Plant Instrumentation and Control Systems
NUREG/CR-7010 Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE)
NUREG/CR-7011 Evaluation of Treatment of Effects of Debris in Coolant on ECCS and CSS Performance in Pressurized Water Reactors and Boiling Water Reactors
NUREG/CR-7012 Uncertainties in Predicted Isotopic Compositions for High Burnup PWR Spent Nuclear Fuel
NUREG/CR-7013 Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation—Vandellós II Reactor
NUREG/CR-7014 Processes, Properties, and Conditions Controlling In Situ Bioremediation of Uranium in Shallow, Alluvial Aquifers
NUREG/CR-7015 Analysis of JNES Seismic Tests on Degraded Piping
NUREG/CR-7016 Human Reliability Analysis-Informed Insights on Cask Drops
NUREG/CR-7017 Preliminary, Qualitative Human Reliability Analysis for Spent Fuel Handling
NUREG/CR-7018 Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments
NUREG/CR-7019 Results of the Program for the Inspection of Nickel Alloy Components
NUREG/CR-7024 Material Property Correlations: Comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO
NUREG/CR-7025 Radionuclide Release from Slag and Concrete Waste Materials, Part I: Conceptual Models of Leaching from Complex Materials and Laboratory Test Methods
NUREG/CR-7026 Application of Model Abstraction Techniques to Simulate Transport in Soils
NUREG/CR-7027 Degradation of LWR Core Internal Materials Due to Neutron Irradiation
NUREG/CR-7028 Engineered Covers for Waste Containment: Changes in Engineering Properties and Implications for Long-Term Performance Assessment
NUREG/CR-7029 Lessons Learned in Detecting, Monitoring, Modeling and Remediating Radioactive Ground-Water Contamination
NUREG/CR-7030 Atmospheric Stress Corrosion Cracking Susceptibility of Welded and Unwelded 304, 304L, and 316L Austenitic Stainless Steels Commonly Used for Dry Cask Storage Containers Exposed to Marine Environments
NUREG/CR-7031 A Compilation of Elevated Temperature Concrete Material Property Data and Information for Use in Assessments of Nuclear Power Plant Reinforced Concrete Structures
NUREG/CR-7032 Developing an Emergency Risk Communication (ERC)/Joint Information Center (JIC) Plan for a Radiological Emergency
NUREG/CR-7033 Guidance on Developing Effective Radiological Risk Communication Messages: Effective Message Mapping and Risk Communication with the Public in Nuclear Plant Emergency Planning Zones
NUREG/CR-7034 Analysis of Severe Railway Accidents Involving Long Duration Fires
NUREG/CR-7035 Analysis of Severe Roadway Accidents Involving Long Duration Fires
NUREG/CR-7037 Industry Performance of Relief Valves at U.S. Commercial Nuclear Power Plants through 2007
NUREG/CR-7038 Verification of RESRAD-OFFSITE
NUREG/CR-7039 Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8
NUREG/CR-7040 Evaluation of JNES Equipment Fragility Tests for Use in Seismic Probabilistic Risk Assessments for U.S. Nuclear Power Plants
NUREG/CR-7041 SCALE/TRITON Primer: A Primer for Light Water Reactor Lattice Physics Calculations
NUREG/CR-7042 A Large Scale Validation of a Methodology for Assessing Software Reliability
NUREG/CR-7044 Development of Quantitative Software Reliability Models for Digital Protection Systems of Nuclear Power Plants
NUREG/CR-7045 Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data
NUREG/CR-7046 Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America
NUREG/CR-7047 LAPUR 6.0 Benchmark Against Data from the GENESIS Facility
NUREG/CR-7100 Direct Current Electrical Shorting in Response to Exposure Fire
NUREG/CR-7101 Structural Materials Analyses of the Newhall Pass Tunnel Fire, 2007
NUREG/CR-7102 Kerite Analysis in Thermal Environment of FIRE (KATE-Fire): Test Results – Final Report
NUREG/CR-7103 Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Alloys
NUREG/CR-7105 Radionuclide Release from Slag and Concrete Waste Materials – Part 2: Relationship Between Laboratory Tests and Field Leaching
NUREG/CR-7106 Generation of a Broad-Group HTGR Library for Use with SCALE
NUREG/CR-7108 An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses—Isotopic Composition Predictions
NUREG/CR-7109 An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses—Criticality (keff) Predictions
NUREG/CR-7110 State-of-the-Art Reactor Consequence Analyses Project
NUREG/CR-7111 A Summary of Aging Effects and Their Management in Reactor Spent Fuel Pools, Refueling Cavities, Tori, and Safety-Related Concrete Structures
NUREG/CR-7113 An Assessment of Ultrasonic Techniques for Far-Side Examinations of Austenitic Stainless Steel Piping Welds
NUREG/CR-7114 Methodology for Low Power/Shutdown Fire PRA — Draft Report for Comment
NUREG/CR-7115 Performance of Metal and Polymeric O-Ring Seals in Beyond-Design-Basis Temperature Excursions
NUREG/CR-7116 Materials Aging Issues and Aging Management for Extended Storage and Transportation of Spent Nuclear Fuel
NUREG/CR-7117 Secure Network Design
NUREG/CR-7119 Experimental Studies of Reinforced Concrete Structures Under Multi-Directional Earthquakes and Design Implications
NUREG/CR-7120 Radionuclide Behavior in Soils and Soil-to-Plant Concentration Ratios for Assessing Food Chain Pathways
NUREG/CR-7122 An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds
NUREG/CR-7123 A Literature Review of the Effects of Smoke from a Fire on Electrical Equipment
NUREG/CR-7124 Validation of LAPUR 6.0 Code
NUREG/CR-7126 Human-Performance Issues Related to the Design and Operation of Small Modular Reactors
NUREG/CR-7127 New Source Term Model for the RESRAD-OFFSITE Code Version 3
NUREG/CR-7128 Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR-60 Reactor
NUREG/CR-7134 The Estimation of Very-Low Probability Hurricane Storm Surges for Design and Licensing of Nuclear Power Plants in Coastal Areas
NUREG/CR-7136 Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion
NUREG/CR-7137 Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment - 2009
NUREG/CR-7139 Assessment of Current Test Methods for Post-LOCA Cladding Behavior
NUREG/CR-7142 Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation
NUREG/CR-7143 Characterization of Thermal-Hydraulic and Ignition Phenomena in Prototypic, Full-Length Boiling Water Reactor Spent Fuel Pool Assemblies After a Postulated Complete Loss-of-Coolant Accident
NUREG/CR-7144 Laminar Hydraulic Analysis of a Commercial Pressurized Water Reactor Fuel Assembly
NUREG/CR-7148 Confirmatory Battery Testing: The Use of Float Current Monitoring to Determine Battery State-of-Charge
NUREG/CR-7149 Effects of Degradation on the Severe Accident Consequences for a PWR Plant with a Reinforced Concrete Containment Vessel
NUREG/CR-7150 Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE) – Final Report
NUREG/CR-7151 Development of a Fault Injection-Based Dependability Assessment Methodology for Digital I&C Systems
NUREG/CR-7154 Risk Informing Emergency Preparedness Oversight: Evaluation of Emergency Action Levels — A Pilot Study of Peach Bottom, Surry and Sequoyah
NUREG/CR-7157 Computational Benchmark for Estimated Reactivity Margin from Fission Products and Minor Actinides in BWR Burnup Credit
NUREG/CR-7158 Review and Prioritization of Technical Issues Related to Burnup Credit for BWR Fuel
NUREG/CR-7161 Synthesis of Distributions Representing Important Non-Site-Specific Parameters in Off-Site Consequence Analyses
NUREG/CR-7171 A Review of the Effects of Radiation on Microstructure and Properties of Concretes Used in Nuclear Power Plants
NUREG/CR-7172 Knowledge Base Report on Emergency Core Cooling Sump Performance in Operating Light Water Reactors
Page Last Reviewed/Updated Tuesday, April 15, 2014