Home
>
NRC Library
>
Document Collections
>
NUREG-Series Publications
> Publications Prepared by NRC Contractors
Publications Prepared by NRC Contractors
`Search NUREG-Series Pubs
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-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-0381 | A Preliminary Report on Fire Protection Research Program Fire Barriers and Fire Retardant Coatings Tests |
| NUREG/CR-0468 | Nuclear Power Plant Fire Protection — Fire Barriers (Subsystems Study Task 3) |
| NUREG/CR-0488 | Nuclear Power Plant Fire Protection — Fire Detection (Subsystems Study Task 2) |
| NUREG/CR-0596 | A Preliminary Report on Fire Protection Research Program, Fire Barriers and Suppression (September 15, 1978 Test) |
| NUREG/CR-0636 | Nuclear Power Plant Fire Protection — Ventilation (Subsystems Study Task 1) |
| NUREG/CR-0654 | Nuclear Power Plant Fire Protection — Fire-Hazards Analysis (Subsystems Study Task 4) |
| NUREG/CR-0833 | Fire Protection Research Program Corner Effects Tests |
| NUREG/CR-1156 | Environmental Assessment of Ionization Chamber Smoke Detectors Containing Am-241 |
| NUREG/CR-1184 | Evaluation of Simulator Adequacy for the Radiation Qualification of Safety-Related Equipment |
| NUREG/CR-1405 | The NACOM Code for Analysis of Postulated Sodium Spray Fires in LMFBRs |
| NUREG/CR-1429 | Seismic Review Table |
| NUREG/CR-1552 | Development and Verification of Fire Tests for Cable Systems and System Components: Quarterly Report 12, March – May 1980 |
| NUREG/CR-1614 | Approaches to Acceptable Risk: A Critical Guide |
| NUREG/CR-1682 | Electrical Insulators in a Reactor Accident Environment |
| NUREG/CR-1798 | Acceptance and Verification for Early Warning Fire Detection Systems: Interim Guide |
| NUREG/CR-1819 | Development and Testing of a Model for Fire Potential in Nuclear Power Plants |
| NUREG/CR-1916 | A Risk Comparison |
| NUREG/CR-1930 | Index of Risk Exposure and Risk Acceptance Criteria |
| NUREG/CR-2040 | A Study of the Implications of Applying Quantitative Risk Criteria in the Licensing of Nuclear Power Plants in the United States |
| NUREG/CR-2258 | Fire Risk Analysis for Nuclear Power Plants |
| NUREG/CR-2260 | Technical Basis for Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants" |
| NUREG/CR-2269 | Probabilistic Models for the Behavior of Compartment Fires |
| NUREG/CR-2300 | PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants |
| NUREG/CR-2321 | Investigation of Fire Stop Test Parameters |
| NUREG/CR-2377 | Test and Criteria for Fire Protection of Cable Penetrations |
| NUREG/CR-2409 | Requirements for Establishing Detector
Siting Criteria in Fires Involving Electrical 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-2726 | Light Water Reactor Hydrogen Manual |
| NUREG/CR-2730 | Hydrogen Burn Survival: Preliminary Thermal Model and Test Results |
| NUREG/CR-2815 | Probabilistic Safety Analysis Procedures Guide |
| NUREG/CR-2858 | PAVAN: An Atmospheric-Dispersion Program for Evaluating Design-Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations |
| NUREG/CR-2868 | Aging Effects on Fire-Retardant Additives in Organic Materials for Nuclear Plant Applications |
| NUREG/CR-2927 | Nuclear Power Plant Electrical Cable Damageability Experiments |
| NUREG/CR-3037 | User's Manual for FIRIN: A Computer Code to Estimate Accidental Fire and Airborne Releases in Nuclear Fuel Cycle Facilities |
| NUREG/CR-3122 | Potentially Damaging Failure Modes of High- and Medium-Voltage Electrical Equipment |
| NUREG/CR-3139 | Scenarios and Analytical Methods for UF6 Releases at NRC-Licensed Fuel Cycle Facilities |
| NUREG/CR-3192 | Investigation of Twenty-Foot Separation Distance as a Fire Protection Method as Specified in 10 CFR 50, Appendix R |
| NUREG/CR-3239 | COMPBRN — A Computer Code for Modeling Compartment Fires |
| NUREG/CR-3242 | The Los Alamos National Laboratory/New Mexico State University Filter Plugging Test Facility: Description and Preliminary Test Results |
| NUREG/CR-3263 | Status Report: Correlation of Electrical Cable Failure with Mechanical Degradation |
| NUREG/CR-3330 | Vulnerability of Nuclear Power Plant Structures to Large External Fires |
| NUREG/CR-3385 | Measures of Risk Importance and Their Applications |
| NUREG/CR-3468 | Hydrogen:Air:Steam Flammability Limits and Combustion Characteristics in the FITS Vessel |
| NUREG/CR-3493 | A Review of the Limerick Generating Station Severe Accident Risk Assessment: Review of Core-Melt Frequency |
| NUREG/CR-3521 | Hydrogen-Burn Survival Experiments at Fully Instrumented Test Site (FITS) |
| NUREG/CR-3527 | Material Transport Analysis for Accident-Induced Flow in Nuclear Facilities |
| NUREG/CR-3532 | Response of Rubber Insulation Materials to Monoenergetic Electron Irradiations |
| NUREG/CR-3629 | The Effect of Thermal and Irradiation Aging Simulation Procedures on Polymer Properties |
| NUREG/CR-3638 | Hydrogen-Steam Jet-Flame Facility and Experiments |
| NUREG/CR-3656 | Evaluation of Suppression Methods for Electrical Cable Fires |
| NUREG/CR-3719 | Detonation Calculations Using a Modified Version of CSQII: Examples for Hydrogen-Air Mixtures |
| NUREG/CR-3735 | Accident-Induced Flow and Material Transport in Nuclear Facilities: A Literature Review |
| NUREG/CR-3805 | Engineering Characterization of Ground Motion: Task II: Summary Report |
| NUREG/CR-3922 | Survey and Evaluation of System Interaction Events and Sources |
| NUREG/CR-4112 | Investigation of Cable and Cable System Fire Test Parameters |
| NUREG/CR-4138 | Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests |
| NUREG/CR-4229 | Evaluation of Current Methodology Employed in Probabilistic Risk Assessment (PRA) of Fire Events at Nuclear Power Plants |
| NUREG/CR-4230 | Probability-Based Evaluation of Selected Fire Protection Features in Nuclear Power Plants |
| NUREG/CR-4231 | Evaluation of Available Data for Probabilistic Risk Assessments (PRA) of Fire Events at Nuclear Power Plants |
| NUREG/CR-4264 | Investigation of High-Efficiency Particulate Air Filter Plugging by Combustion Aerosols |
| NUREG/CR-4310 | Investigation of Potential Fire-Related Damage to Safety-Related Equipment in Nuclear Power Plants |
| NUREG/CR-4321 | Full-Scale Measurements of Smoke Transport and Deposition in Ventilation System Ductwork |
| NUREG/CR-4330 | Review of Light Water Reactor Regulatory Requirements |
| NUREG/CR-4432 | Comparison of Dynamic Characteristics of Fukushima Nuclear Power Plant Containment Building Determined From Tests and Earthquakes |
| NUREG/CR-4461 | Tornado Climatology of the Contiguous United States |
| NUREG/CR-4479 | The Use of a Field Model To Assess Fire Behavior in Complex Nuclear Power Plant Enclosures: Present Capabilities and Future Prospects |
| NUREG/CR-4513 | Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems |
| NUREG/CR-4517 | Design Features for Enhancing International Safeguards of Away-from- Reactor Dry Storage for Spent LWR Fuel |
| NUREG/CR-4527 | An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets |
| NUREG/CR-4534 | Analysis of Diffusion Flame Tests |
| NUREG/CR-4561 | FIRAC User's Manual: A Computer Code to Simulate Fire Accidents in Nuclear Facilities |
| NUREG/CR-4566 | COMPBRN III - A Computer Code for Modeling Compartment Fires |
| NUREG/CR-4570 | Description and Testing of an Apparatus for Electrically Initiating Fires Through Simulation of a Faulty Connection |
| NUREG/CR-4586 | Users' Guide for a Personal-Computer-Based Nuclear Power Plant Fire Data Base |
| NUREG/CR-4596 | Screening Tests of Representative Nuclear Power Plant Components Exposed to Secondary Environments Created by Fires |
| NUREG/CR-4638 | Transient Fire Environment Cable Damageability Test Results |
| NUREG/CR-4667 | Environmentally Assisted Cracking in Light Water Reactors: Annual Report |
| NUREG/CR-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-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-5117 | Steam Generator Tube Integrity Program/Steam Generator Group Project: Final Project Summary Report |
| 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-5347 | Recommendations for Resolution of Public Comments on USI A-40, “Seismic Design Criteria” |
| NUREG/CR-5384 | A Summary of Nuclear Power Plant Fire Safety Research at Sandia National Laboratories, 1975–1987 |
| NUREG/CR-5385 | Initial Assessment of the Mechanisms and Significance of Low-Temperature Embrittlement of Cast Stainless Steels in LWR Systems |
| NUREG/CR-5392 | Elements of an Approach to Performance-Based Regulatory Oversight |
| NUREG/CR-5457 | A Review of the Three Mile Island-1 Probabilistic Risk Assessment |
| NUREG/CR-5500 | Reliability Study |
| NUREG/CR-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-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-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-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-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-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-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-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-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-6314 | Quality Assurance Inspections for Shipping and Storage Containers |
| 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-6372 | Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts |
| NUREG/CR-6384 | Literature Review of Environmental Qualification of Safety-Related Electric Cables |
| NUREG/CR-6406 | Environmental Testing of an Experimental Digital Safety Channel |
| NUREG/CR-6407 | Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety |
| NUREG/CR-6410 | Nuclear Fuel Cycle Facility Accident Analysis Handbook |
| NUREG/CR-6420 | Self-Monitoring Surveillance System for Prestressing Tendons |
| NUREG/CR-6421 | A Proposed Acceptance Process for Commercial Off-the-Shelf (COTS) Software in Reactor Applications |
| NUREG/CR-6424 | Report on Aging of Nuclear Power Plant Reinforced Concrete Structures |
| NUREG/CR-6428 | Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds |
| NUREG/CR-6463 | Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems |
| 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-6544 | A Methodology for Analyzing Precursors to Earthquake-Initiated and Fire-Initiated Accident Sequences |
| 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-6583 | Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels |
| 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-6627 | The Role of Organic Complexants and Colloids in the Transport of Radionuclides by Groundwater |
| 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-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-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-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-6755 | Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario Using the Probabilistic RESRAD-BUILD 3.0 Code |
| NUREG/CR-6758 | Radionuclide-Chelating Agent Complexes in Low-Level Radioactive Decontamination Waste: Stability, Adsorbtion, and Transport Potential |
| 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-6791 | Eddy Current Reliability Results from the Steam Generator Mock-up Analysis Round-Robin: Revision 1 |
| 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-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-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: A Subtask of GSI-191 |
| NUREG/CR-6903 | Human Event Repository and Analysis (HERA) System, Overview |
| NUREG/CR-6904 | Evaluation of the Broadband Impedance Spectroscopy Prognostic/Diagnostic Technique for Electric Cables Used in Nuclear Power Plants |
| NUREG/CR-6905 | Report of Experimental Results for the International Fire Model Benchmarking and Validation Exercise #3 |
| NUREG/CR-6906 | Containment Integrity Research at Sandia National Laboratories - An Overview |
| NUREG/CR-6907 | Crack Growth Rates of Nickel Alloy Welds in a PWR Environment |
| NUREG/CR-6909 | Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials |
| NUREG/CR-6910 | Alternative Conceptual Models for Assessing Food Chain Pathways in Biosphere Models |
| NUREG/CR-6911 | Tests of Uranium (VI) Adsorption Models in a Field Setting |
| NUREG/CR-6912 | GSI-191 PWR Sump Screen Blockage Chemical Effects Tests: Thermodynamic Simulations |
| NUREG/CR-6913 | Chemical Effects Head-Loss Research In Support of Generic Safety Issue 191 |
| NUREG/CR-6914 | Integrated Chemical Effects Test Project |
| NUREG/CR-6915 | Aluminum Chemistry in a Prototypical Post-Loss-of-Coolant-Accident, Pressurized-Water-Reactor Containment Environment |
| NUREG/CR-6916 | Hydraulic Transport of Coating Debris |
| NUREG/CR-6917 | Experimental Measurements of Pressure Drop Across Sump Screen Debris Beds in Support of Generic Safety Issue 191 |
| NUREG/CR-6918 | VARSKIN 4: A Computer Code for Assessing Skin Dose from Skin Contamination |
| NUREG/CR-6919 | Recommendations for Revision of Seismic Damping Values in Regulatory Guide 1.61 |
| NUREG/CR-6920 | Risk-Informed Assessment of Degraded Containment Vessels |
| NUREG/CR-6921 | Crack Growth Rates in a PWR Environment of Nickel Alloys from the Davis-Besse and V.C. Summer Power Plants |
| NUREG/CR-6922 | P-CARES: Probabilistic Computer Analysis for Rapid Evaluation of Structures |
| NUREG/CR-6923 | Expert Panel Report on Proactive Materials Degradation Assessment |
| NUREG/CR-6924 | Non-destructive and Failure Evaluation of Tubing from a Retired Steam Generator |
| NUREG/CR-6925 | Assessment of Analysis Methods for Seismic Shear Wall Capacity Using JNES/NUPEC Multi-Axial Cyclic and Shaking Table Test Data |
| NUREG/CR-6926 | Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for Application to Nuclear Power Plants |
| NUREG/CR-6927 | Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent Factors |
| NUREG/CR-6928 | Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants |
| NUREG/CR-6929 | Assessment of Eddy Current Testing for the Detection of Cracks in Cast Stainless Steel Reactor Piping Components |
| NUREG/CR-6930 | Temperature Dependence of Weibull Stress Parameters: Studies Using the Euro-Material Similar to ASME A508 Class-3 Steel |
| NUREG/CR-6931 | Carolfire Test Report |
| NUREG/CR-6932 | Baseline Risk Index for Initiating Events (BRIIE) |
| NUREG/CR-6933 | Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-Frequency Ultrasonic Methods |
| NUREG/CR-6934 | Fatigue Crack Flaw Tolerance in Nuclear Power Plant Piping - A Basis for Improvements to ASME Code Section XI Appendix L |
| NUREG/CR-6935 | Sensitivity Studies of Failure of Steam Generator Tubes during Main Steam Line Break and Other Secondary Side Depressurization Events |
| NUREG/CR-6936 | Probabilities of Failure and Uncertainty Estimate Information for Passive Components – A Literature Review |
| NUREG/CR-6938 | Final Report-Assessment of Potential Phosphate Ion-Cenmentitious Materials Interactions |
| NUREG/CR-6939 | Coexistence Assessment of Industrial Wireless Protocols in the Nuclear Facility Environment |
| NUREG/CR-6940 | Combined Estimation of Hydrogeologic Conceptual Model, Parameter, and Scenario Uncertainty with Application to Uranium Transport at the Hanford Site 300 Area |
| NUREG/CR-6941 | Soil-to-Plant Concentration Ratios for Assessing Food-Chain Pathways in Biosphere Models |
| NUREG/CR-6942 | Dynamic Reliability Modeling of Digital Instrumentation and Control Systems for Nuclear Reactor Probabilistic Risk Assessments |
| NUREG/CR-6943 | A Study of Remote Visual Methods to Detect Cracking in Reactor Components |
| NUREG/CR-6944 | Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) |
| NUREG/CR-6945 | Fabrication Flaw Density and Distribution in Repairs to Reactor Pressure Vessel and Piping Welds |
| NUREG/CR-6946 | Field Studies to Confirm Uncertainty Estimates of Ground-Water Recharge |
| NUREG/CR-6947 | Human Factors Considerations with Respect to Emerging Technology in Nuclear Power Plants |
| NUREG/CR-6948 | Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities and Sites: Logic, Strategic Approach and Discussion |
| NUREG/CR-6949 | The Employment of Empirical Data and Bayesian Methods in Human Reliability Analysis: A Feasibility Study |
| NUREG/CR-6951 | Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit |
| NUREG/CR-6952 | Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) |
| NUREG/CR-6953 | Review of NUREG-0654, Supplement 3, "Criteria for Protective Action Recommendations for Severe Accidents" |
| NUREG/CR-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-6962 | Traditional Probabilistic Risk Assessment Methods for Digital Systems |
| 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-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-6981 | Assessment of Emergency Response Planning and Implementation for Large Scale Evacuations |
| NUREG/CR-6982 | Assessment of Noise Level for Eddy Current Inspection of Steam Generator Tubes |
| NUREG/CR-6983 | Seismic Analysis of Large-Scale Piping Systems for the JNES-NUPEC Ultimate Strength Piping Test Program |
| NUREG/CR-6984 | Field Evaluation of Low-Frequency SAFT-UT on Cast Stainless Steel and Dissimilar Metal Weld Components |
| NUREG/CR-6985 | A Benchmark Implementation of Two Dynamic Methodologies for the Reliability Modeling of Digital Instrumentation and Control Systems |
| NUREG/CR-6986 | Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs |
| NUREG/CR-6987 | Analysis of Structural Materials Exposed to a Severe Fire Environment |
| NUREG/CR-6988 | Final Report — Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant |
| NUREG/CR-6989 | Methodology for Estimating Fabrication Flaw Density and Distribution – Reactor Pressure Vessel Welds |
| NUREG/CR-6990 | Development and Testing of ENDF/B-VI.8 and ENDF/B-VII.0 Coupled Neutron-Gamma Libraries for SCALE 6 |
| NUREG/CR-6991 | Design Practices for Communications and Workstations in Highly Integrated Control Rooms |
| NUREG/CR-6992 | Instrumentation and Controls in Nuclear Power Plants: An Emerging Technologies Update |
| NUREG/CR-6994 | Argonne Model Boiler Test Results |
| NUREG/CR-6995 | SCDAP/RELAP5 Thermal-Hydraulic Evaluations of the Potential for Containment Bypass During Extended Station Blackout Severe Accident Sequences in a Westinghouse Four-Loop PWR |
| NUREG/CR-6996 | Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechanism Penetrations |
| NUREG/CR-6997 | Modeling a Digital Feedwater Control System Using Traditional Probabilistic Risk Assessment Methods |
| NUREG/CR-6998 | Review of Information for Spent Nuclear Fuel Burnup Confirmation |
| NUREG/CR-6999 | Technical Basis for a Proposed Expansion of Regulatory Guide 3.54 — Decay Heat Generation in an Independent Spent Fuel Storage Installation |
| NUREG/CR-7000 | Essential Elements of an Electric Cable Condition Monitoring Program |
| NUREG/CR-7001 | Predictive Bias and Sensitivity in NRC Fuel Performance Codes |
| NUREG/CR-7002 | Criteria for Development of Evacuation Time Estimate Studies |
| NUREG/CR-7003 | Background and Derivation of ANS-5.4 Standard Fission Product Release Model |
| NUREG/CR-7004 | Technical Basis for Regulatory Guidance on Design-Basis Hurricane-Borne Missile Speeds for Nuclear Power Plants |
| NUREG/CR-7005 | Technical Basis for Regulatory Guidance on Design-Basis Hurricane Wind Speeds for Nuclear Power Plants |
| NUREG/CR-7006 | Guidelines for Field-Programmable Gate Arrays in Nuclear Power Plant Safety Systems Plant |
| NUREG/CR-7007 | Diversity Strategies for Nuclear Power Plant Instrumentation and Control Systems |
| NUREG/CR-7010 | Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE) |
| NUREG/CR-7011 | Evaluation of Treatment of Effects of Debris in Coolant on ECCS and CSS Performance in Pressurized Water Reactors and Boiling Water Reactors |
| NUREG/CR-7012 | Uncertainties in Predicted Isotopic Compositions for High Burnup PWR Spent Nuclear Fuel |
| NUREG/CR-7013 | Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation—Vandellós II Reactor |
| NUREG/CR-7014 | Processes, Properties, and Conditions Controlling In Situ Bioremediation of Uranium in Shallow, Alluvial Aquifers |
| NUREG/CR-7015 | Analysis of JNES Seismic Tests on Degraded Piping |
| NUREG/CR-7016 | Human Reliability Analysis-Informed Insights on Cask Drops |
| NUREG/CR-7017 | Preliminary, Qualitative Human Reliability Analysis for Spent Fuel Handling |
| NUREG/CR-7018 | Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments |
| NUREG/CR-7019 | Results of the Program for the Inspection of Nickel Alloy Components |
| NUREG/CR-7022 | FRAPCON-3.4 |
| NUREG/CR-7023 | FRAPTRAN 1.4 |
| NUREG/CR-7024 | Material Property Correlations: Comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO |
| NUREG/CR-7025 | Radionuclide Release from Slag and Concrete Waste Materials, Part I: Conceptual Models of Leaching from Complex Materials and Laboratory Test Methods |
| NUREG/CR-7026 | Application of Model Abstraction Techniques to Simulate Transport in Soils |
| NUREG/CR-7027 | Degradation of LWR Core Internal Materials Due to Neutron Irradiation |
| NUREG/CR-7028 | Engineered Covers for Waste Containment: Changes in Engineering Properties and Implications for Long-Term Performance Assessment |
| NUREG/CR-7029 | Lessons Learned in Detecting, Monitoring, Modeling and Remediating Radioactive Ground-Water Contamination |
| NUREG/CR-7030 | Atmospheric Stress Corrosion Cracking Susceptibility of Welded and Unwelded 304, 304L, and 316L Austenitic Stainless Steels Commonly Used for Dry Cask Storage Containers Exposed to Marine Environments |
| NUREG/CR-7031 | A Compilation of Elevated Temperature Concrete Material Property Data and Information for Use in Assessments of Nuclear Power Plant Reinforced Concrete Structures |
| NUREG/CR-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-7042 | A Large Scale Validation of a Methodology for Assessing Software Reliability |
| NUREG/CR-7046 | Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America |
| 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-7105 | Radionuclide Release from Slag and Concrete Waste Materials – Part 2: Relationship Between Laboratory Tests and Field Leaching |
| NUREG/CR-7110 | State-of-the-Art Reactor Consequence Analyses Project |
| NUREG/CR-7114 | Methodology for Low Power/Shutdown Fire PRA — Draft Report for Comment |
| NUREG/CR-7115 | Performance of Metal and Polymeric O-Ring Seals in Beyond-Design-Basis Temperature Excursions |
| NUREG/CR-7120 | Radionuclide Behavior in Soils and Soil-to-Plant Concentration Ratios for Assessing Food Chain Pathways |
| NUREG/CR-7123 | A Literature Review of the Effects of Smoke from a Fire on Electrical Equipment |
Page Last Reviewed/Updated Tuesday, July 17, 2012
