| Document Identifier |
Title |
| NUREG/CR-0152 |
Development and Verification of Fire Tests for Cable Systems and System Components |
| 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 |
| NUREG/CR-0488 |
Nuclear Power Plant Fire Protection
- Fire Detection |
| NUREG/CR-0596 |
A Preliminary Report on Fire Protection Research Program, Fire Barriers and Suppression |
| NUREG/CR-0636 |
Nuclear Power Plant Fire Protection - Ventilation |
| NUREG/CR-0654 |
Nuclear Power Plant Fire Protection - Fire-Hazards Analysis |
| NUREG/CR-0833 |
Fire Protection Research Program
Corner Effects Tests |
| 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-1552 |
Development and Verification of Fire Tests for Cable Systems and System Components |
| NUREG/CR-1614 |
Approaches to Acceptable Risk: A Critical Guide |
| NUREG/CR-1682 |
Electrical Insulators in a Reactor Accident Environment |
| NUREG/CR-1798 |
Acceptance and Verification For Early Warning Fire Detection Systems |
| 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-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-2269 |
Probabilistic Models for the
Behavior of Compartment Fires |
| NUREG/CR-2300 |
A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants |
| NUREG/CR-2321 |
Investigation of Fire Stop Test Parameters Final Report |
| NUREG/CR-2377 |
Test and Criteria for Fire Protection Of Cable Penetrations |
| NUREG/CR-2409 |
Requirements for Establishing Detector
Siting Criteria in Fires Involving
Electrical Materials |
| 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-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-2868 |
Aging Effects on Fire-Retardant Additives in Organic Materials for Nuclear Plant Applications |
| NUREG/CR-2927 |
Nuclear Power Plant Electrical Cable Damageability Experiments |
| NUREG/CR-3037 |
A Computer Code to Estimate Accidental Fire and Airborne Releases in Nuclear Fuel Cycle Facilities Radioactive |
| 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-3242 |
The Los Alamos National Laboratory/New Mexico State University Filter Plugging Test Facility Description and Preliminary Test Results |
| NUREG/CR-3263 |
Status Report: Correlation of Electrical Cable Failure with Mechanical Degradation |
| NUREG/CR-3330 |
Vulnerability of Nuclear Power Plant Structures to Large External Fires |
| NUREG/CR-3385 |
Measures of Risk Importance And Their Applications |
| NUREG/CR-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-4112 |
Investigation of Cable and Cable System Fire Test Parameters |
| NUREG/CR-4138 |
Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests |
| NUREG/CR-4229 |
Evaluation of Current Methodology Employed in Probabilistic Risk Assessment (PRA) of Fire Events at Nuclear Power Plants |
| NUREG/CR-4230 |
Probability-Based Evaluation of Selected Fire Protection Features in Nuclear Power Plants |
| NUREG/CR-4231 |
Evaluation of Available Data, for, Probabilistic Risk Assessments (PRA) of Fire Events at Nuclear Power Plants |
| NUREG/CR-4264 |
Investigation of High-efficiency Particulate Air Filter Plugging by Combustion Aerosols |
| NUREG/CR-4310 |
Investigation of Potential Fire-Related Damage to Safety-Related Equipment in Nuclear Power Plants |
| NUREG/CR-4321 |
Full-Scale Measurements of Smoke Transport and Deposition in Ventilation System Ductwork |
| NUREG/CR-4330 |
Review of Light Water Reactor Regulatory Requirements Identification of Regulatory Requirements That May Have Marginal Importance To Risk |
| 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-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 |
User Guide for a Personal-Computer-Based Nuclear Power Plan Fire Data Base |
| NUREG/CR-4596 |
Screening Tests of Representative Nuclear Power Plant Components Exposed to Secondary Environments Created by Fires |
| NUREG/CR-4638 |
Transient Fire Environment Cable Damageability Test Results |
| NUREG/CR-4667 |
Environmentally Assisted Cracking in Light Water Reactors |
| 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-4829 |
Shipping Container Response to Severe Highway and Railway Accident Conditions |
| NUREG/CR-4830 |
MELCOR Validation and Verification: 1986 Papers |
| NUREG/CR-4839 |
Methods for External Event Screening Quantification: Risk Methods Integration and Evaluation Program (RMIEP) Methods Development |
| NUREG/CR-4840 |
Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150 |
| NUREG/CR-4855 |
Development and Application of a Computer Model for Large-Scale Flame Acceleration Experiments |
| NUREG/CR-4905 |
Detonability of H2-Air-Diluent Mixtures |
| NUREG/CR-5037 |
Fire Environment Determination in the LaSalle Nuclear Power Plant Control Rroom |
| NUREG/CR-5079 |
Experimental Results Pertaining to the Performance of Thermal Igniters |
| NUREG/CR-5233 |
A Computer Code for Fire Protection and Risk Analysis of Nuclear Plants |
| 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-5281 |
Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent Fuel Pools |
| NUREG/CR-5384 |
A Summary of Nuclear Power Plant Fire Safety Research at Sandia National Laboratories, 1975-1987 |
| NUREG/CR-5385 |
Initial Assessment of the Mechanisms and Significance of Low-Temperature Embrittlement of Cast Stainless Steels in LWR Systems |
| NUREG/CR-5392 |
Elements of an Approach to Performance-Based Regulatory Oversight |
| NUREG/CR-5457 |
A Review of the Three Mile Island-1 Probabilistic Risk Assessment |
| NUREG/CR-5500 |
Reliability Study |
| NUREG/CR-5525 |
Hydrogen-Air-Diluent Detonation Study for Nuclear Reactor Safety Analyses |
| NUREG/CR-5546 |
An Investigation of the Effects of Thermal Aging on, the Fire Damageability
of Electric Cables |
| NUREG/CR-5580 |
Evaluation of Generic Issue 57 |
| NUREG/CR-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-5669 |
Evaluation of Exposure Limits
to Toxic Gases for Nuclear
Reactor Control Room Operators |
| 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-5734 |
Recommendations to the NRC on Acceptable Standard Format
and Content for the Fundamental Nuclear Material Control (FNMC) Plan Required
for Low-Enriched Uranium Enrichment Facilities |
| 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-6017 |
Fire Modeling of the Heiss Dampf Reaktor Containment |
| NUREG/CR-6042 |
Perspectives on Reactor Safety |
| 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-6095 |
Aging, Loss-of-Coolant Accident (LOCA), and High Potential Testing of Damaged Cables |
| NUREG/CR-6142 |
Tensile-Property Characterization of Thermally Aged Cast Stainless Steels |
| NUREG/CR-6173 |
A Summary of the Fire Testing Program at the German HDR Test Facility |
| NUREG/CR-6213 |
High-Temperature Hydrogen-Air- Steam Detonation Experiments in the BNL Small-Scale Development Apparatus |
| 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-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-6314 |
Quality Assurance Inspections for Shipping and Storage Containers |
| NUREG/CR-6345 |
Radiation Dose Estimates for Radiopharmaceuticals |
| NUREG/CR-6358 |
Assessment of United States Industry Structural Codes and Standards for Application to Advanced Nuclear Power Reactors |
| 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-6421 |
A Proposed Acceptance Process for Commercial Off-the-Shelf (COTS) Software in Reactor Applications |
| NUREG/CR-6428 |
Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds |
| NUREG/CR-6444 |
Methodology for Analyzing Precursors to Earthquake-Initiated and Fire-Initiated Accident Sequences |
| NUREG/CR-6476 |
Circuit Bridging of Components by Smoke |
| NUREG/CR-6477 |
Revised Analyses of Decommissioning Reference Non-Fuel-Cycle
Facilities |
| NUREG/CR-6479 |
Technical Basis for Environmental Qualification of Microprocessor-Based Safety-Related Equipment in Nuclear Power Plants |
| NUREG/CR-6500 |
Owners of Nuclear Power Plants |
| NUREG/CR-6509 |
The Effect of Initial Temperature on Flame Acceleration and Deflagration-to-Detonation Transition Phenomenon |
| NUREG/CR-6524 |
The Effect of Lateral Venting on Deflagration-to-Detonation Transition in Hydrogen-Air-Steam Mixtures at Various Initial Temperatures |
| NUREG/CR-6525 |
SECPOP2000: Sector Population, Land Fraction, and Economic
Estimation Program |
| NUREG/CR-6530 |
Deliberate Ignition of Hydrogen-Air-Steam Mixtures in Condensing Steam Environments |
| NUREG/CR-6543 |
Effects of Smoke on Functional Circuits |
| 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 |
| NUREG/CR-6583 |
Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels |
| NUREG/CR-6595 |
An Approach for Estimating the Frequencies of Various Containment
Failure Modes and Bypass Events |
| NUREG/CR-6607 |
Guidance for Performing Probabilistic Seismic Hazard Analysis
for a Nuclear Plant Site: Example Application to the Southeastern United
States |
| NUREG/CR-6632 |
Solubility and Leaching of Radionuclides in Site Decommissioning
Management Plan (SDMP) Slags |
| NUREG/CR-6656 |
Information on Hydrologic Conceptual Models, Parameters,
Uncertainty Analysis, and Data Sources for Dose Assessments at Decommissioning
Sites |
| NUREG/CR-6679 |
Assessment of Age-Related Degradation of Structures
and Passive Components for U.S. Nuclear Power Plants |
| NUREG/CR-6682 |
Summary and Categorization of Public Comments on Controlling
the Disposition of Solid Materials |
| 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-6695 |
Hydrologic Uncertainty Assessment for Decommissioning Sites:
Hypothetical Test Case Applications |
| NUREG/CR-6717 |
Environmental Effects on Fatigue Crack Initiation in Piping and Pressure Vessel Steels |
| 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-6738 |
Risk Methods Insights Gained From Fire Incidents |
| NUREG/CR-6749 |
Integrating Digital and Conventional Human-System Interfaces:
Lessons Learned from a Control Room Modernization Program |
| NUREG/CR-6751 |
The Human Performance Evaluation Process: A Resource for
Reviewing the Identification and Resolution of Human Performance Problems |
| NUREG/CR-6753 |
Review of Findings for Human Error Contribution to Risk
in Operating Events |
| NUREG/CR-6758 |
Radionuclide-Chelating Agent Complexes in Low-Level Radioactive
Decontamination Waste; Stability, Adsorbtion and Transport Potential |
| NUREG/CR-6755 |
Technical Basis for Calculating Radiation Doses for the
Building Occupancy Scenario Using the Probabilistic RESRAD-BUILD 3.0 Code |
| NUREG/CR-6761 |
Parametric Study of the Effect of Burnable Poison Rods
for PWR Burnup Credit |
| 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-6775 |
Human Performance Characterization in the Reactor Oversight Process |
| NUREG/CR-6776 |
Cable Insulation Resistance Measurements Made During Cable
Fire Tests |
| NUREG/CR-6782 |
Comparison of U.S. Military and International Electromagnetic
Compatibility Guidance |
| NUREG/CR-6787 |
Mechanism and Estimation of Fatigue Crack Initiation in Austenitic Stainless Steels in LWR Environments |
| NUREG/CR-6793 |
Numerical Simulation of the Howard Street Tunnel Fire,
Baltimore, Maryland, July 2001 |
| NUREG/CR-6799 |
Analysis of Rail Car Components Exposed to a Tunnel Fire Environment |
| NUREG/CR-6805 |
A Comprehensive Strategy of Hydrogeologic Modeling and
Uncertainty Analysis for Nuclear Facilities and Sites |
| 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-6813 |
Issues and Recommendations for Advancement of PRA Technology
In Risk-Informed Decision Making |
| 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-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-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-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-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-6848 |
Preliminary Validation of a Methodology for Assessing Software Quality |
| 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-6860 |
An Assessment of Visual Testing |
| NUREG/CR-6861 |
Barrier Integrity Research Program: Final Report |
| 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-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-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-6888 |
Emerging Technologies in Instrumentation and Controls: An Update |
| 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 |
| NUREG/CR-6903 |
Human Event Repository and Analysis (HERA) System, Overview |
| NUREG/CR-6904 |
Evaluation of the Broadband
Impedance Spectroscopy
Prognostic/Diagnostic
Technique for Electric Cables
Used in Nuclear Power Plants |
| NUREG/CR-6905 |
Report of Experimental Results for the International Fire Model Benchmarking and Validation Exercise #3 |
| NUREG/CR-6906 |
Containment Integrity Research at Sandia National Laboratories - An Overview |
| NUREG/CR-6907 |
Crack Growth Rates of Nickel Alloy Welds in a PWR Environment |
| NUREG/CR-6909 |
Effect of LWR Coolant
Environments on the
Fatigue Life of
Reactor Materials |
| NUREG/CR-6910 |
Alternative Conceptual Models for Assessing Food Chain Pathways in Biosphere Models |
| NUREG/CR-6911 |
Tests of Uranium (VI) Adsorption Models in a Field Setting |
| NUREG/CR-6912 |
GSI-191 PWR Sump
Screen Blockage Chemical
Effects Tests: Thermodynamic
Simulations |
| NUREG/CR-6913 |
Chemical Effects Head-Loss Research In Support of Generic Safety Issue 191 |
| NUREG/CR-6914 |
Integrated Chemical Effects Test Project |
| NUREG/CR-6915 |
Aluminum Chemistry in a Prototypical
Post-Loss-of-Coolant-Accident,
Pressurized-Water-Reactor Containment
Environment |
| NUREG/CR-6916 |
Hydraulic Transport of Coating Debris |
| NUREG/CR-6917 |
Experimental Measurements of Pressure Drop Across Sump Screen Debris Beds in Support of Generic Safety Issue 191 |
| NUREG/CR-6918 |
VARSKIN 3: A Computer Code for Assessing Skin Dose from Skin Contamination |
| NUREG/CR-6919 |
Recommendations for Revision of Seismic Damping Values in Regulatory Guide 1.61 |
| NUREG/CR-6920 |
Risk-Informed Assessment of Degraded Containment Vessels |
| NUREG/CR-6921 |
Crack Growth Rates in a PWR Environment of Nickel Alloys from the Davis-Besse and V.C. Summer Power Plants |
| NUREG/CR-6922 |
P-CARES: Probabilistic Computer Analysis for Rapid Evaluation of Structures |
| NUREG/CR-6923 |
Expert Panel Report on Proactive Materials Degradation Assessment |
| NUREG/CR-6924 |
Non-destructive and Failure Evaluation of Tubing from a Retired Steam Generator |
| NUREG/CR-6925 |
Assessment of Analysis Methods for Seismic Shear Wall Capacity Using JNES/NUPEC Multi-Axial Cyclic and Shaking Table Test Data |
| NUREG/CR-6926 |
Evaluation of the Seismic Design
Criteria in ASCE/SEI
Standard 43-05 for Application to
Nuclear Power Plants |
| NUREG/CR-6927 |
Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors |
| NUREG/CR-6928 |
Industry-Average Performance
for Components and Initiating
Events at U.S. Commercial
Nuclear Power Plants |
| NUREG/CR-6929 |
Assessment of Eddy Current Testing for the Detection of Cracks in Cast Stainless Steel
Reactor Piping Components |
| NUREG/CR-6930 |
Temperature Dependence of Weibull Stress Parameters: Studies Using the Euro-Material Similar to ASME A508 Class-3 Steel |
| NUREG/CR-6931 |
Carolfire Test Report |
| NUREG/CR-6932 |
Baseline Risk Index for Initiating Events (BRIIE) |
| NUREG/CR-6933 |
Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods |
| NUREG/CR-6934 |
Fatigue Crack Flaw Tolerance in Nuclear Power Plant Piping
- A Basis for Improvements to ASME Code Section XI Appendix L |
| NUREG/CR-6935 |
Sensitivity Studies of Failure of Steam Generator Tubes during Main Steam Line Break and Other Secondary Side Depressurization Events |
| NUREG/CR-6936 |
Probabilities of Failure and Uncertainty Estimate Information for Passive Components – A Literature Review |
| NUREG/CR-6938 |
Final Report-Assessment of Potential Phosphate Ion-Cenmentitious Materials Interactions |
| NUREG/CR-6939 |
Coexistence Assessment of Industrial Wireless Protocols in the Nuclear Facility Environment |
| NUREG/CR-6940 |
Combined Estimation of Hydrogeologic Conceptual Model, Parameter, and Scenario Uncertainty with Application to Uranium Transport at the Hanford Site 300 Area |
| NUREG/CR-6941 |
Soil-to-Plant Concentration Ratios for Assessing Food-Chain Pathways in Biosphere Models |
| NUREG/CR-6942 |
Dynamic Reliability Modeling of
Digital Instrumentation and
Control Systems for Nuclear Reactor
Probabilistic Risk Assessments |
| NUREG/CR-6943 |
A Study of Remote Visual Methods to Detect Cracking in Reactor Components |
| NUREG/CR-6944 |
Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) |
| NUREG/CR-6945 |
Fabrication Flaw Density and Distribution in Repairs to Reactor Pressure Vessel and Piping Welds |
| NUREG/CR-6946 |
Field Studies to Confirm Uncertainty Estimates of Ground-Water Recharge |
| NUREG/CR-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-6953 |
Review of NUREG-0654, Supplement 3, "Criteria for Protective Action Recommendations for Severe Accidents" |
| 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-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-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-6973 |
Technical Basis for Assessing Uranium Bioremediation Performance |
