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Spotlight
Publications Prepared by NRC Contractors
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-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-4605 | Training Manual on Statistical Methods for Nuclear Material Management |
| 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-5908 | Advanced Human-system Interface Design Review Guideline |
| 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-6208 | An Empirical Investigation of Operator Performance in Cognitively Demanding Simulated Emergencies |
| 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-7247 | Research to Develop Guidance on Extreme Precipitation Frequency Estimates in Orographic Regions |
| 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-7264 | Managing the Effects of Degraded Digital Instrumentation and Control Conditions on Operator Performance: Human Factors Engineering Review Guidance Development |
| 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-7292 | Structured Hazard Assessment Committee Process for Flooding (SHAC-F) for Local Intense Precipitation, Riverine, and Coastal Flooding |
| 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 |
| NUREG/CR-7310 | Spent Fuel Storage and Transportation of Accident Tolerant Concepts: Cr-Coated Zirconium Alloy Cladding, FeCrAl Cladding, High Burnup and High Enrichment Fuel |
| NUREG/CR-7311 | Determination of Bias and Bias Uncertainty for Criticality Safety Computational Methods |
| NUREG/CR-7312 | Irradiation Effects on Reinforced Concrete Structures – Experimental and Analytical Study on Irradiated Concrete – Steel Bonding, Modeling and Simulation of Structural Response |
| NUREG/CR-7313 | Operating Experience Data Analysis for Digital Instrumentation and Control System Reliability and Risk Assessment |
Page Last Reviewed/Updated Tuesday, September 08, 2026
Page Last Reviewed/Updated Tuesday, September 08, 2026