| 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-1005 |
A Radioactive Waste Disposal Classification System |
| 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-1621 |
A Characterization of Faults in the Appalachian Foldbelt |
| NUREG/CR-1682 |
Electrical Insulators in a Reactor Accident Environment |
| NUREG/CR-1756 |
Technology, Safety and Costs of Decommissioning Reference Nuclear Research and Test Reactors |
| NUREG/CR-1759 |
Data Base for Radioactive Waste Management |
| 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-2384 |
Age-Specific Inhalation Radiation Dose Commitment Factors for Selected Radionuclides |
| 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-2907 |
Radioactive Effluents from Nuclear Power Plants |
| NUREG/CR-2927 |
Nuclear Power Plant Electrical Cable Damageability Experiments |
| NUREG/CR-3021 |
Regional Tectonics and Seismicity of Southwestern Iowa |
| 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-3331 |
A Methodology for Allocating Nuclear Power Plant Control Nuclear to Human or Automatic Control |
| NUREG/CR-3332 |
Radiological Assessment: A Textbook on Environmental Dose Analyses |
| 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-4062 |
Extended Storage of Low-Level Radioactive Waste: Potential Problem Areas |
| 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-4251 |
Mitigative Techniques for Ground-Water Contamination Associated With Severe Nuclear Accidents |
| 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-4370 |
Update of Part 61 – Impacts Analysis Methodology: Codes and Example Problems |
| 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-4644 |
Geochemical Studies of Commercial Low-Level Radioactive Waste Disposal Sites |
| 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-4884 |
Interpretation of Bioassay Measurements |
| 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-5081 |
Tactical Exercise Planning Handbook |
| NUREG/CR-5117 |
Steam Generator Tube Integrity Program/Steam Generator Group Project: Final Project Summary Report |
| NUREG/CR-5172 |
Tactical Training Reference Manual |
| 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-5434 |
Anchor Bolt Behavior and Strength During Earthquakes |
| NUREG/CR-5457 |
A Review of the Three Mile Island-1 Probabilistic Risk Assessment |
| NUREG/CR-5466 |
Service Life of Concrete |
| 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-5532 |
A Performance Assessment Methodology for Low-Level Waste Facilities |
| NUREG/CR-5542 |
Models for Estimation of Service Life of Concrete Barriers in Low-Level Radioactive Waste Disposal |
| NUREG/CR-5546 |
An Investigation of the Effects of Thermal Aging on the Fire Damageability of Electric Cables |
| NUREG/CR-5562 |
Dating and Earthquakes: Review of Quaternary Geochronology and Its Application to Paleoseismology |
| 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-5927 |
Evaluation of a Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities: Evaluation of Modeling Approaches |
| NUREG/CR-5941 |
Technical Basis for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related I&C Systems |
| NUREG/CR-5966 |
A Simplified Model of Aerosol Removal by Containment Sprays |
| 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-6174 |
Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station |
| NUREG/CR-6189 |
A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments |
| NUREG/CR-6210 |
Computer Codes for Evaluation of Control Room Habitability (HABIT) |
| 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-6331 |
Atmospheric Relative Concentrations in Building Wakes |
| 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-6361 |
Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages |
| 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-6377 |
Effects of Radionuclide Concentrations by Cement/Ground-Water Interactions in Support of Performance Assessment of Low-Level Radioactive Waste Disposal Facilities |
| NUREG/CR-6384 |
Literature Review of Environmental Qualification of Safety-Related Electric Cables |
| NUREG/CR-6393 |
Integrated System Validation: Methodology and Review Criteria |
| NUREG/CR-6400 |
Human Factors Engineering (HFE) Insights for Advanced Reactors Based Upon Operating Experience |
| 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-6478 |
Motor-Operated Valve (MOV) Actuator Motor and Gearbox Testing |
| 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 |
SecPop: 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-6620 |
Testing of dc-Powered Actuators for Motor-Operated Valves |
| 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-6645 |
Reevaluation of Regulatory Guidance on Modal Response Combination Methods for Seismic Response Spectrum Analysis |
| 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-6726 |
Aging Management and Performance of Stainless Steel Bellows in Nuclear Power Plants |
| 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-6741 |
Application of Microprocessor-Based Equipment in Nuclear Power Plants-Technical Basis for a Qualification Methodology |
| 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-6827 |
Assessment of Internal Oxidation (IO) as a Mechanism for Submodes of Stress Corrosion Cracking (SCC) that Occur on the Secondary Side of Steam Generators |
| 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-6857 |
RELAP5/MOD3.2.2 Gamma Assessment for Pressurized Thermal Shock Applications |
| 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 Water 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: A Computer Code for Skin Contamination and Dosimetry Assessments |
| 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-7008 |
MELCOR Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project |
| NUREG/CR-7009 |
MACCS Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project |
| 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-7021 |
A Subsurface Decision Model for Supporting Environmental Compliance |
| NUREG/CR-7022 |
FRAPCON |
| NUREG/CR-7023 |
FRAPTRAN |
| NUREG/CR-7024 |
Material Property Correlations: Comparisons between FRAPCON, FRAPTRAN, 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-7107 |
Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis |
| 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 |
A Framework for Low Power/Shutdown Fire PRA |
| 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-7131 |
Review of Probable Maximum Precipitation Procedures and Databases Used to Develop Hydrometeorological Reports |
| NUREG/CR-7132 |
Application of Radar-Rainfall Estimates to Probable Maximum Precipitation in the Carolinas |
| NUREG/CR-7133 |
Synthesis of Extreme Storm Rainfall and Probable Maximum Precipitation in the Southeastern U.S. Pilot Region |
| 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-7135 |
Compensatory and Alternative Regulatory MEasures for Nuclear Power Plant FIRE Protection (CARMEN-FIRE) |
| 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-7141 |
The U.S. Nuclear Regulatory Commission's Cyber Security Regulatory Framework for Nuclear Power Reactors |
| 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-7145 |
Nuclear Power Plant Security Assessment Guide |
| 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-7152 |
Rod Bundle Heat Transfer Facility – Steady-State Steam Cooling Experiments |
| NUREG/CR-7153 |
Expanded Materials Degradation Assessment (EMDA) |
| NUREG/CR-7154 |
Risk Informing Emergency Preparedness Oversight: Evaluation of Emergency Action Levels — A Pilot Study of Peach Bottom, Surry and Sequoyah |
| NUREG/CR-7155 |
State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Long-Term Station Blackout of the Peach Bottom Atomic Power Station |
| NUREG/CR-7156 |
Fitness for Duty in the Nuclear Power Industry: An Update of Technical Issues on Drugs of Abuse Testing and Fatigue Management |
| 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-7159 |
Reliability of Ultrasonic In-Service Inspection of Welds in Reactor Internals of Boiling Water Reactors |
| NUREG/CR-7160 |
Emergency Preparedness Significance Quantification Process: Proof of Concept |
| NUREG/CR-7161 |
Synthesis of Distributions Representing Important Non-Site-Specific Parameters in Off-Site Consequence Analyses |
| NUREG/CR-7162 |
Analysis of Experimental Data for High Burnup BWR Spent Fuel Isotopic Validation – SVEA-96 and GE14 Assembly Designs |
| NUREG/CR-7163 |
A Formalized Approach for the Collection of HRA Data from Nuclear Power Plant Simulators |
| NUREG/CR-7164 |
Cross Section Generation Guidelines for TRACE–PARCS |
| NUREG/CR-7165 |
The Technical Basis Supporting ASME Code, Section XI, Appendix VIII: Performance Demonstration for Ultrasonic Examination Systems |
| NUREG/CR-7166 |
Radiological Toolbox User's Guide |
| NUREG/CR-7167 |
Assessing the Potential for Biorestoration of Uranium In Situ Recovery Sites |
| NUREG/CR-7168 |
Regulatory Approaches for Addressing Reprocessing Facility Risks: An Assessment |
| NUREG/CR-7169 |
Sensors and Monitoring to Assess Grout and Vault Behavior for Performance Assessment |
| NUREG/CR-7170 |
Assessment of Stress Corrosion Cracking Susceptibility for Austenitic Stainless Steels Exposed to Atmospheric Chloride and Non-Chloride Salts |
| 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 |
| NUREG/CR-7174 |
Transfer Factors for Contaminant Uptake by Fruit and Nut Trees |
| NUREG/CR-7175 |
Susceptibility of Nuclear Stations to External Faults |
| NUREG/CR-7176 |
Safety and Regulatory Issues of the Thorium Fuel Cycle |
| NUREG/CR-7177 |
Compendium of Analyses to Investigate Select Level 1 Probabilistic Risk Assessment End-State Definition and Success Criteria Modeling Issues |
| NUREG/CR-7178 |
Uranium Sequestration During Biostimulated Reduction and In Response to the Return of Oxic Conditions In Shallow Aquifers |
| NUREG/CR-7179 |
BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 1: Model Development and Events
Leading to Instability |
| NUREG/CR-7180 |
BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 2: Sensitivity Studies for Events Leading to Instability |
| NUREG/CR-7181 |
BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 3: Events Leading to Emergency Depressurization |
| NUREG/CR-7182 |
BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 4: Sensitivity Studies for Events Leading to Emergency Depressurization |
| NUREG/CR-7183 |
Best Practices for Behavioral Observation Programs at Operating Power Reactors and Power Reactor Construction Sites |
| NUREG/CR-7184 |
Crack Growth Rate and Fracture Toughness Tests on Irradiated Cast Stainless Steels |
| NUREG/CR-7185 |
Effect of Thermal Aging and Neutron Irradiation on Crack Growth Rate and Fracture Toughness of Cast Stainless Steels and Austenitic Stainless Steel Welds |
| NUREG/CR-7186 |
Experimental Measurement of Suppression Pool Void Distribution During Blowdown in Support of Generic Issue 193 |
| NUREG/CR-7187 |
Managing PWSCC in Butt Welds by Mitigation and Inspection |
| NUREG/CR-7188 |
Testing to Evaluate Extended Battery Operation in Nuclear Power Plants |
| NUREG/CR-7189 |
User's Guide for RESRAD-OFFSITE |
| NUREG/CR-7190 |
Workload, Situation Awareness, and Teamwork |
| NUREG/CR-7191 |
Thermal Analysis of Horizontal Storage Casks for Extended Storage Applications |
| NUREG/CR-7192 |
Rod Bundle Heat Transfer Facility Steam Cooling with Droplet Injection Experiments Data Report |
| NUREG/CR-7193 |
Evaluations of NRC Seismic-Structural Regulations and Regulatory Guidance, and Simulation-Evaluation Tools for Applicability to Small Modular Reactors (SMRs) |
| NUREG/CR-7194 |
Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems |
| NUREG/CR-7195 |
Risk-Informed and Performance-Based Oversight of Radiological Emergency Response Programs |
| NUREG/CR-7196 |
Large Scale Earthquake Simulation of a Hybrid Lead Rubber Isolation System Designed with Consideration of Nuclear Seismicity |
| NUREG/CR-7197 |
Heat Release Rates of Electrical Enclosure Fires (HELEN-FIRE) |
| NUREG/CR-7198 |
Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications |
| NUREG/CR-7199 |
Radionuclide Release from Slag and Concrete Waste Materials – Part 3: Testing Protocol |
| NUREG/CR-7200 |
Influence of Coupling Erosion and Hydrology on the Long-Term Performance of Engineered Surface Barriers |
| NUREG/CR-7201 |
Characterizing Explosive Effects on Underground Structures |
| NUREG/CR-7202 |
NRC Reviewer Aid for Evaluating the Human-Performance Aspects Related to the Design and Operation of Small Modular Reactors |
| NUREG/CR-7203 |
A Quantitative Impact Assessment of Hypothetical Spent Fuel Reconfiguration in Spent Fuel Storage Casks and Transportation Packages |
| NUREG/CR-7204 |
Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping |
| NUREG/CR-7205 |
Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks |
| NUREG/CR-7206 |
Spent Fuel Transportation Package Response to the MacArthur Maze Fire Scenario |
| NUREG/CR-7207 |
Spent Fuel Transportation Package Response to the Newhall Pass Tunnel Fire Scenario |
| NUREG/CR-7208 |
Study on Post Tensioning Methods |
| NUREG/CR-7209 |
A Compendium of Spent Fuel Transportation Package Response Analyses to Severe Fire Accident Scenarios |
| NUREG/CR-7210 |
Development and Validation of Models for Predicting Leakage from Degraded Tube-to-Tubesheet Joints During Severe Accidents |
| NUREG/CR-7211 |
Application of a Hydrological Uncertainty Methodology to Nuclear Reactor Site Evaluations |
| NUREG/CR-7212 |
Technical Manual and User's Guide for MILDOS-AREA Version 4 |
| NUREG/CR-7213 |
MILDOS-AREA Computation Verification Version 4 |
| NUREG/CR-7214 |
Toward a More Risk-Informed and Performance-Based Framework for the Regulation of the Seismic Safety of Nuclear Power Plants |
| NUREG/CR-7215 |
Spent Fuel Pool Project Phase 1: Pre-Ignition and Ignition Testing of a Single Commercial 17x17 Pressurized Water Reactor Spent Fuel Assembly under Complete Loss of Coolant Accident Conditions |
| NUREG/CR-7216 |
Spent Fuel Pool Project Phase II: Pre-Ignition and Ignition Testing of a 1x4 Commercial 17x17 Pressurized Water Reactor Spent Fuel Assemblies under Complete Loss of Coolant Accident Conditions |
| NUREG/CR-7217 |
Application of Automated Analysis Software to Eddy Current Inspection Data from Steam Generator Tube Bundle Mock-up |
| NUREG/CR-7218 |
Rod Bundle Heat Transfer Facility Two-Phase Mixture Level Swell and Uncovery Test Experiments Data Report |
| NUREG/CR-7219 |
Cladding Behavior during Postulated Loss-of-Coolant Accidents |
| NUREG/CR-7220 |
SNAP/RADTRAD 4.0: Description of Models and Methods |
| NUREG/CR-7221 |
Integrating Model Abstraction into Subsurface Monitoring Strategies |
| NUREG/CR-7222 |
Tsunami Hazard Assessment Based on Wave Generation, Propagation, and Inundation Modeling for the U.S. East Coast |
| NUREG/CR-7223 |
Tsunami Hazard Assessment: Best Modeling Practices and State-of-the-Art Technology |
| NUREG/CR-7224 |
Axial Moderator Density Distributions, Control Blade Usage, and Axial Burnup Distributions for Extended BWR Burnup Credit |
| NUREG/CR-7225 |
Stability of Circumferential Flaws in Once-Through Steam Generator Tubes Under Thermal Loading During LOCA, MSLB and FWLB |
| NUREG/CR-7226 |
Primary Water Stress Corrosion Cracking of High-Chromium, Nickel-Base Welds Near Dissimilar Metal Weld Interfaces |
| NUREG/CR-7227 |
US Commercial Spent Nuclear Fuel Assembly Characteristics: 1968-2013 |
| NUREG/CR-7228 |
Open Secondary Testing of Window-Type Current Transformers |
| NUREG/CR-7229 |
Testing to Evaluate Battery and Battery Charger Short-Circuit Current Contributions to a Fault on the DC Distribution System |
| NUREG/CR-7230 |
Seismic Design Standards and Calculational Methods in the United States and Japan |
| NUREG/CR-7231 |
Modeling of Radionuclide Transport in Freshwater Systems Associated with Nuclear Power Plants |
| NUREG/CR-7232 |
Review of Spent Fuel Reprocessing and Associated Accident Phenomena |
| NUREG/CR-7233 |
Developing a Bayesian Belief Network Model for Quantifying the Probability of Software Failure of a Protection System |
| NUREG/CR-7234 |
Development of A Statistical Testing Approach for Quantifying Safety-Related Digital System on Demand Failure Probability |
| NUREG/CR-7235 |
Results of Blind Testing for the Program to Assess the Reliability of Emerging Nondestructive Techniques |
| NUREG/CR-7236 |
Results of Open Testing for the Program to Assess the Reliability of Emerging Nondestructive Techniques |
| NUREG/CR-7237 |
Correlation of Seismic Performance in Similar SSCs (Structures, Systems, and Components) |
| NUREG/CR-7238 |
Guidance Document: Conducting Paleoliquefaction Studies for Earthquake Source Characterization |
| NUREG/CR-7239 |
Review of Exemptions and General Licenses for Fissile Material in 10 CFR 71 |
| NUREG/CR-7240 |
Impact of Operating Parameters on Extended BWR Burnup Credit |
| NUREG/CR-7243 |
PIMAL: Phantom with Moving Arms and Legs – Version 4.1.0 |
| NUREG/CR-7244 |
Response of Nuclear Power Plant Instrumentation Cables Exposed to Fire Conditions |
| NUREG/CR-7245 |
State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Sequoyah Integrated Deterministic and Uncertainty Analyses |
| NUREG/CR-7246 |
Reliability Assessment of Remote Visual Examination |
| NUREG/CR-7248 |
Capabilities and Practices of Offsite Response Organizations for Protective Actions in the Intermediate Phase of a Radiological Emergency Response |
| NUREG/CR-7249 |
Overview of Nuclear Data Uncertainty in Scale and Application to Light Water Reactor Uncertainty Analysis |
| NUREG/CR-7250 |
Thermal-Hydraulic Experiments Using A Dry Cask Simulator |
| NUREG/CR-7251 |
Margins for Uncertainty in the Predicted Spent Fuel Isotopic Inventories for BWR Burnup Credit |
| NUREG/CR-7252 |
Validation of keff Calculations for Extended BWR Burnup Credit |
| NUREG/CR-7253 |
Technical Considerations for Seismic Isolation of Nuclear Facilities |
| NUREG/CR-7254 |
Seismic Isolation of Nuclear Power Plants Using Sliding Bearings |
| NUREG/CR-7255 |
Seismic Isolation of Nuclear Power Plants using Elastomeric Bearings |
| NUREG/CR-7256 |
Effects of Environmental Conditions on Manual Actions for Flood Protection and Mitigation |
| NUREG/CR-7257 |
Paleoliquefaction Studies in Moderate Seismicity Regions with a History of Large Events |
| NUREG/CR-7258 |
Technical Manual and User’s Guide for MILDOS, Version 4.1 |
| NUREG/CR-7259 |
MILDOS Version 4.1 Computational Verification Report |
| NUREG/CR-7260 |
CFD Validation of Vertical Dry Cask Storage System |
| NUREG/CR-7262 |
State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Short-Term Station Blackout of the Surry Power Station |
| NUREG/CR-7263 |
NDE Reliability Issues for the Examination of CASS Components |
| NUREG/CR-7265 |
Phenomena Identification and Ranking Technique (PIRT) Exercise for Ranking Low-Power Shutdown Plant Operating States and Outage Types |
| NUREG/CR-7266 |
MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel |
| NUREG/CR-7267 |
Default Parameter Values and Distribution in RESRAD-ONSITE V7.2, RESRAD-BUILD V3.5, and RESRAD-OFFSITE V4.0 Computer Codes |
| NUREG/CR-7268 |
User's Manual for RESRAD-OFFSITE Code Version 4 |
| NUREG/CR-7269 |
Enhancing Guidance for Evacuation Time Estimate Studies |
| NUREG/CR-7270 |
Technical Bases for Consequence Analyses Using MACCS (MELCOR Accident Consequence Code System) |
| NUREG/CR-7271 |
Application of Point Precipitation Frequency Estimates to Watersheds |
| NUREG/CR-7272 |
NRC ATWS-I Stability Tests with Downskew Axial Power Profile: KATHY Test Series STS123 |
| NUREG/CR-7273 |
Developing a Technical Basis for Embedded Digital Devices and Emerging Technologies |
| NUREG/CR-7274 |
Validation of a Computational Fluid Dynamics Method Using Horizontal Dry Cask Simulator Data |
| NUREG/CR-7275 |
Jet Impingement in High-Energy Piping Systems |
| NUREG/CR-7276 |
Primary Water Stress Corrosion Cracking of High-Chromium Nickel-Base Welds – 2018 |
| NUREG/CR-7277 |
Primary Water Stress Corrosion Cracking of High-Chromium Nickel-Base Welds at or Near Interfaces – 2020 |
| NUREG/CR-7278 |
Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications |
| NUREG/CR-7279 |
Research to Develop Flood Barrier Testing Strategies for Nuclear Power Plants |
| NUREG/CR-7280 |
Review of Radiation-Induced Concrete Degradation and Potential Implications for Structures Exposed to High Long-Term Radiation Levels in Nuclear Power Plants |
| NUREG/CR-7281 |
Radiation Evaluation Methodology for Concrete Structures |
| NUREG/CR-7282 |
Review of Accident Tolerant Fuel Concepts with Implications to Severe Accident Progression and Radiological Releases |
| NUREG/CR-7283 |
Phenomena Identification Ranking Tables for Accident Tolerant Fuel Designs Applicable to Severe Accident Conditions |
| NUREG/CR-7284 |
SCALE 6.2 Lattice Physics Performance Assessment |
| NUREG/CR-7285 |
Nonradiological Health Consequences from Evacuation and Relocation |
| NUREG/CR-7286 |
Reactor Pressure Vessel Fluence Evaluation Methodology for Extended Beltline Locations |
| NUREG/CR-7287 |
Numerical Modeling of Local Intense Precipitation Processes |
| NUREG/CR-7288 |
Evaluation of In-Service Radon Barriers over Uranium Mill Tailings Disposal Facilities |
| NUREG/CR-7289 |
Nuclear Data Assessment for Advanced Reactors |
| NUREG/CR-7290 |
Convection-Permitting Modeling for Intense Precipitation Processes |
| NUREG/CR-7291 |
Assessments on Eddy Current Detection of Cracking Near Volumetric Indications in Steam Generator Tubes |
| NUREG/CR-7293 |
The Price-Anderson Act: 2021 Report to Congress, Public Liability Insurance and Indemnity Requirements for an Evolving Commercial Nuclear Industry Office |
| NUREG/CR-7294 |
Exploring Advanced Computational Tools and Techniques with Artificial Intelligence and Machine Learning in Operating Nuclear Plants |
| NUREG/CR-7295 |
Human Factors in Nondestructive Examination |
| NUREG/CR-7296 |
Multi-Mechanism Flood Hazard Assessment: Critical Review of Current Practice and Approaches and Example Use Studies |
| NUREG/CR-7297 |
Basis for Technical Guidance To Evaluate Evapotranspiration Covers |
| NUREG/CR-7299 |
Fuel Qualification for Molten Salt Reactors |
| NUREG/CR-7300 |
Radiation Accident Dose and Simulated Loss-of-Coolant Accident Test of Low Voltage Cables |
| NUREG/CR-7301 |
Ultrasonic Modeling and Simulation for Nuclear Nondestructive Evaluation |
| NUREG/CR-7302 |
Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses |
| NUREG/CR-7303 |
Validating Actinides and Fission Products for Burnup Credit Criticality Safety Analyses - Nuclide Compositions Prediction with Extended Validation Basis |
| NUREG/CR-7304 |
Evaluating Flaw Detectability Under Limited-Coverage Conditions |
| NUREG/CR-7305 |
Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, "Fuel Qualification for Advanced Reactors" |
| NUREG/CR-7306 |
Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water Reactor Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages |
| NUREG/CR-7307 |
Phenomena Identification and Ranking Tables on High Burnup Fuel Fragmentation, Relocation, Dispersal, and Its Consequences for Design-Basis Accidents in Pressurized- and Boiling-Water Reactors |
| NUREG/CR-7308 |
Sensitivity/Uncertainty Methods for Nuclear Criticality Safety Validation |
| NUREG/CR-7309 |
Validation Studies for High Burnup and Extended Enrichment Fuels in Burnup Credit Criticality Safety Analyses |