HomeNRC LibraryDocument CollectionsNUREG-Series PublicationsPublications Prepared by NRC Contractors

# Phenomena Identification and Ranking Technique (PIRT) Exercise for Ranking Low-Power Shutdown Plant Operating States and Outage Types (NUREG/CR-7265)

Page Last Reviewed/Updated Thursday, March 25, 2021

Section Navigation

#### Navigation

          - NDE Reliability Issues for the Examination of CASS Components (NUREG/CR-7263, PNNL-28840)
          - CFD Validation of Vertical Dry Cask Storage System (NUREG/CR-7260)
          - Paleoliquefaction Studies in Moderate Seismicity Regions with a History of Large Events (NUREG/CR-72
          - Seismic Isolation of Nuclear Power Plants using Elastomeric Bearings (NUREG/CR-7255)
          - Seismic Isolation of Nuclear Power Plants Using Sliding Bearings (NUREG/CR-7254)
          - Technical Considerations for Seismic Isolation of Nuclear Facilities (NUREG/CR-7253)
          - 2018
          - Margins for Uncertainty in the Predicted Spent Fuel Isotopic Inventories for BWR Burnup Credit (NURE
          - Thermal-Hydraulic Experiments Using A Dry Cask Simulator (NUREG/CR-7250)
          - Overview of Nuclear Data Uncertainty in Scale and Application to Light Water Reactor Uncertainty Ana
          - Capabilities and Practices of Offsite Response Organizations for Protective Actions in the Intermedi
          - Reliability Assessment of Remote Visual Examination (NUREG/CR-7246, PNNL-27003)
          - State-of-the-Art Reactor Consequence Analyses (SOARCA) Project: Sequoyah Integrated Deterministic an
          - PIMAL: Phantom with Moving Arms and Legs – Version 4.1.0 (NUREG/CR-7243)
          - Toward a More Risk-Informed  and Performance-Based Framework for the Regulation of the Seismic Safet
          - MILDOS-AREA Computation Verification Version 4 (NUREG/CR-7213, ANL/EVS-15/10)
          - Technical Manual and User's Guide for MILDOS-AREA Version 4 (NUREG/CR-7212, ANL/EVS-15/9)
          - BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 3:  Events L
          - BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 1: Model Dev
          - Uranium Sequestration During Biostimulated Reduction and In Response to the Return of OxicCondition
          - Compendium of Analyses to Investigate Select Level 1 Probabilistic Risk Assessment End-State Definit
          - Safety and Regulatory Issues of the Thorium Fuel Cycle (NUREG/CR-7176)
          - Susceptibility of Nuclear Stations to External Faults (NUREG/CR-7175)
          - Transfer Factors for Contaminant Uptake by Fruit and Nut Trees (NUREG/CR-7174)
          - Knowledge Base Report on Emergency Core Cooling Sump Performance in Operating Light Water Reactors (
          - A Review of the Effects of Radiation on Microstructure and Properties of Concretes Used in Nuclear P
          - Assessment of Stress Corrosion Cracking Susceptibility for Austenitic Stainless Steels Exposed to At
          - Sensors and Monitoring to Assess Grout and Vault Behavior for Performance Assessment (NUREG/CR-7169)
          - Regulatory Approaches for Addressing Reprocessing Facility Risks: An Assessment (NUREG/CR-7168)
          - Assessing the Potential for Biorestoration of Uranium In Situ Recovery Sites (NUREG/CR-7167)
          - 2013
          - The Technical Basis Supporting ASME Code, Section XI, Appendix VIII: Performance Demonstration for U
          - Cross Section Generation Guidelines for TRACE–PARCS (NUREG/CR-7164)
          - A Formalized Approach for the Collection of HRA Data from Nuclear Power Plant Simulators (NUREG/CR-7
          - Analysis of Experimental Data for High Burnup BWR Spent Fuel Isotopic Validation – SVEA-96 and GE14 
          - Synthesis of Distributions Representing Important Non-Site-Specific Parameters in Off-Site Consequen
          - Emergency Preparedness Significance Quantification Process: Proof of Concept (NUREG/CR-7160)
          - Reliability of Ultrasonic In-Service Inspection of Welds in Reactor Internals of Boiling Water React
          - Review and Prioritization of Technical Issues Related to Burnup Credit for BWR Fuel (NUREG/CR-7158, 
          - Computational Benchmark for Estimated Reactivity Margin from Fission Products and Minor Actinides in
          - Fitness for Duty in the Nuclear Power Industry: An Update of Technical Issues on Drugs of Abuse Test
          - State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Long-
          - Risk Informing Emergency Preparedness Oversight: Evaluation of Emergency Action Levels — A Pilot Stu
          - BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 4:Sensitivi
          - Best Practices for Behavioral Observation Programs at Operating Power Reactors and Power Reactor Con
          - Crack Growth Rate and Fracture Toughness Tests on Irradiated Cast Stainless Steels (NUREG/CR-7184)
          - Effect of Thermal Aging and Neutron Irradiation on Crack Growth Rate and Fracture Toughness of Cast 
          - Application of a Hydrological Uncertainty Methodology to Nuclear Reactor Site Evaluations (NUREG/CR-
          - A Compendium of Spent Fuel Transportation Package Response Analyses to Severe Fire Accident Scenario
          - Study on Post Tensioning Methods (NUREG/CR-7208)
          - Spent Fuel Transportation Package Response to the Newhall Pass Tunnel Fire Scenario, Final Report (N
          - Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calcu
          - Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping
          - A Quantitative Impact Assessment of Hypothetical Spent Fuel Reconfiguration in Spent Fuel Storage Ca
          - NRC Reviewer Aid for Evaluating the Human-Performance Aspects Related to the Design and Operation of
          - Characterizing Explosive Effects on Underground Structures (NUREG/CR-7201)
          - Influence of Coupling Erosion and Hydrology on the Long-Term Performance of Engineered Surface Barri
          - Radionuclide Release from Slag and Concrete Waste Materials – Part 3: Testing Protocol (NUREG/CR-719
          - Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications  (NUREG/CR-7198, Revi
          - Heat Release Rates of Electrical Enclosure Fires (HELEN-FIRE),  Final Report (NUREG/CR-7197)
          - Risk-Informed and Performance-Based Oversight of Radiological Emergency Response Programs  (NUREG/CR
          - Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transpor
          - Evaluations of NRC Seismic-Structural Regulations and Regulatory Guidance, and Simulation-Evaluation
          - Rod Bundle Heat Transfer Facility Steam Cooling with Droplet Injection Experiments Data Report (NURE
          - Thermal Analysis of Horizontal Storage Casks for Extended Storage Applications (NUREG/CR-7191)
          - Workload, Situation Awareness, and Teamwork (NUREG/CR-7190)
          - User’s Guide for RESRAD-OFFSITE (NUREG/CR-7189)
          - Testing to Evaluate Extended Battery Operation in Nuclear Power Plants  (NUREG/CR-7188)
          - Managing PWSCC in Butt Welds by Mitigation and Inspection(NUREG/CR-7187)
          - Experimental Measurement of Suppression Pool Void Distribution During Blowdown in Support of Generic
          - Large Scale Earthquake Simulation of a Hybrid Lead Rubber Isolation System Designed with Considerati
          - Developing an Emergency Risk Communication (ERC)/Joint Information Center (JIC) Plan for a Radiologi
          - A Compilation of Elevated Temperature Concrete Material Property Data and Information for Use in Ass
          - Atmospheric Stress Corrosion Cracking Susceptibility of Welded and Unwelded 304, 304L, and 316L Aust
          - Cable Response to Live Fire (CAROLFIRE) (NUREG/CR-6931)
          - Assessment of Crack Detection in Heavy-Walled Cast Stainless Steel Piping Welds Using Advanced Low-F
          - MACCS Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Projec
          - MELCOR Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Proje
          - Diversity Strategies for Nuclear Power Plant Instrumentation and  Control Systems (NUREG/CR-7007, OR
          - Review Guidelines for Field-Programmable Gate Arrays in Nuclear Power Plant Safety Systems (NUREG/CR
          - Technical Basis for Regulatory Guidance on Design-Basis Hurricane Wind Speeds for Nuclear Power Plan
          - Background and Derivation of ANS-5.4 Standard Fission Product Release Model (NUREG/CR-7003)
          - NUREG/CR-7002
          - Predictive Bias and Sensitivity in NRC Fuel Performance Codes (NUREG/CR-7001)
          - Essential Elements of an Electric Cable Condition  Monitoring Program (NUREG/CR-7000, BNL-NUREG-9031
          - Technical Basis for a Proposed Expansion of Regulatory Guide 3.54 — Decay Heat Generation in an Inde
          - Review of Information for Spent Nuclear Fuel Burnup Confirmation (NUREG/CR-6998)
          - Modeling a Digital Feedwater Control System Using Traditional Probabilistic Risk Assessment Methods 
          - Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechani
          - SCDAP/RELAP5 Thermal-Hydraulic Evaluations of the Potential for Containment Bypass During Extended S
          - Argonne Model Boiler Test Results (NUREG/CR-6994)
          - Instrumentation and Controls in Nuclear Power Plants: An Emerging Technologies Update (NUREG/CR-6992
          - Design Practices for Communications and Workstations in Highly Integrated Control Rooms (NUREG/CR-69
          - Development and Testing of ENDF/B-VI.8 and ENDF/B-VII.0 Coupled Neutron-Gamma Libraries for SCALE 6 
          - Methodology for Estimating Fabrication Flaw Density and Distribution – Reactor Pressure Vessel Welds
          - Final Report — Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant  (NUREG/CR-6988)
          - Analysis of Structural Materials Exposed to a Severe Fire Environment  (NUREG/CR-6987)
          - Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs (NUREG/C
          - A Benchmark Implementation of Two Dynamic Methodologies for the Reliability Modeling of Digital Inst
          - Field Evaluation of Low-Frequency SAFT-UT on Cast Stainless Steel and Dissimilar Metal Weld  Compone
          - Seismic Analysis of Large-Scale Piping Systems for the JNES-NUPEC Ultimate Strength Piping Test Prog
          - Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE) (NUREG/CR-7
          - Evaluation of Treatment of Effects of Debris in Coolant on ECCS and CSS Performance in Pressurized W
          - Uncertainties in Predicted Isotopic Compositions for High Burnup PWR Spent Nuclear Fuel (NUREG/CR-70
          - Lessons Learned in Detecting, Monitoring, Modeling and Remediating Radioactive Ground-WaterContamin
          - Engineered Covers for Waste Containment: Changes in Engineering Properties and Implications for Long
          - Degradation of LWR Core Internal Materials Due to Neutron Irradiation (NUREG/CR-7027)
          - Application of Model Abstraction Techniques to Simulate Transport in Soils (NUREG/CR-7026)
          - Material Property Correlations: Comparisons between FRAPCON, FRAPTRAN, and  MATPRO (NUREG/CR-7024)
          - FRAPTRAN-1.4 and FRAPTRAN-1.5 (NUREG/CR-7023)
          - Assessment of Noise Level for Eddy Current Inspection of Steam Generator Tubes (NUREG/CR-6982)
          - FRAPCON-3.4 and FRAPCON-3.5 (NUREG/CR-7022)
          - A Subsurface Decision Model for Supporting Environmental Compliance (NUREG/CR-7021)
          - Results of the Program for the Inspection of Nickel Alloy Components (NUREG/CR-7019, Revision 1)
          - Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments (N
          - 2010
          - 2010
          - Analysis of JNES Seismic Tests on Degraded Piping (NUREG/CR-7015)
          - Processes, Properties, and Conditions Controlling In Situ Bioremediation of Uranium in Shallow, Allu
          - Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation—Vandellós II Reacto
          - Assessment of Emergency Response Planning and Implementation for Large Scale Evacuations (NUREG/CR-6
          - Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data (NUREG/CR-6979)
          - Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit (NUREG/CR-695
          - The Employment of Empirical Data and Bayesian Methods in Human Reliability Analysis: A Feasibility S
          - Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities and Sites: Logic, Strategic 
          - Human Factors Considerations with Respect to Emerging Technology in Nuclear Power Plants (NUREG/CR-6
          - Field Studies to Confirm Uncertainty Estimates of Ground-Water Recharge (NUREG/CR-6946)
          - Fabrication Flaw Density and Distribution in Repairs to Reactor Pressure Vessel and Piping Welds (NU
          - Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) (NUREG/CR-6944)
          - A Study of Remote Visual Methods to Detect Cracking in Reactor Components (NUREG/CR-6943)
          - Dynamic Reliability Modeling of Digital Instrumentation and Control Systems for Nuclear Reactor Prob
          - Soil-to-Plant Concentration Ratios for Assessing Food-Chain Pathways in Biosphere Models (NUREG/CR-6
          - Combined Estimation of Hydrogeologic Conceptual Model, Parameter, and Scenario Uncertainty with Appl
          - Coexistence Assessment of Industrial Wireless Protocols in the Nuclear Facility Environment (NUREG/C
          - Final Report-Assessment of Potential Phosphate Ion-Cenmentitious Materials Interactions (NUREG/CR-69
          - Probabilities of Failure and Uncertainty Estimate Information for Passive Components – A Literature 
          - Sensitivity Studies of Failure of Steam Generator Tubes during Main Steam Line Break and Other Secon
          - Fatigue Crack Flaw Tolerance in Nuclear Power Plant  Piping - A Basis for Improvements to ASME Code 
          - A Phenomena Identification and Ranking Table (PIRT) Exercise for Nuclear Power Plant Fire Modeling A
          - Redox and Sorption Reactions of Iodine and Cesium During Transport Through Aquifer Sediments (NUREG/
          - Rod Bundle Heat Transfer Test Facility Description (NUREG/CR-6976)
          - Rod Bundle Heat Transfer Test Facility Test Plan and Design (NUREG/CR-6975)
          - Symbolic Nuclear Analysis Package (SNAP): Common  Application Framework for Engineering Analysis (CA
          - Technical Basis for Assessing Uranium Bioremediation Performance (NUREG/CR-6973)
          - Validation of SCALE 5 Decay Heat Predictions for LWR Spent Nuclear Fuel (NUREG/CR-6972)
          - Spent Fuel Decay Heat Measurements Performed at the Swedish Central Interim Storage Facility (NUREG/
          - Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation – Calvert Cliffs, T
          - Cladding Embrittlement During Postulated  Loss-of-Coolant Accidents (NUREG/CR-6967)
          - Tsunami Hazard Assessment at Nuclear Power Plant Sites in the United States of America - Final Repor
          - RBHT Reflood Heat Transfer Experiments Data and Analysis (NUREG/CR-6980)
          - Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Componen
          - Traditional Probabilistic Risk Assessment Methods for Digital Systems (NUREG/CR-6962)
          - Crack Growth Rates and Fracture Toughness of Irradiated Austenitic Stainless Steels in BWR Environme
          - Application of Surface Complexation Modeling to Selected Radionuclides and Aquifer Sediments  (NUREG
          - LAPUR 6.0 Manual (NUREG/CR-6958)
          - Correlation Analysis of JNES Seismic Wall Pressure Data for ABWR Model Structures (NUREG/CR-6957)
          - Nonlinear Analyses for Embedded Cracks Under Pressurized Thermal Shock:  Comparisons with FAVOR and 
          - Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask (NUREG/CR-6955)
          - Fracture Analysis of Vessels - Oak Ridge FAVOR, v04.1, Computer Code: Theory and Implementation of A
          - Review of NUREG-0654, Supplement 3, "Criteria for Protective Action Recommendations for Severe Accid
          - Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) (NUREG/CR-6952)
          - An Assessment of PWR Steam Generator Condensation at the Oregon State University APEX Facility (NURE
          - 2001
          - Bases for Predicting the Earliest Penetrations Due to SCC for Alloy 600 on the Secondary Side of PWR
          - Drywell Debris Transport Study: Experimental Work (NUREG/CR-6369, Volume 2)
          - Drywell Debris Transport Study (NUREG/CR-6369, Volume 1)
          - Assessment of United States Industry Structural Codes and Standards for Application to Advanced Nucl
          - A Technique for Human ErrorAnalysis (ATHEANA) (NUREG/CR-6350)
          - Hydrologic Evaluation Methodology for Estimating Water Movement Through the Unsaturated Zone at Comm
          - Radiation Dose Estimates for Radiopharmaceuticals (NUREG/CR-6345)
          - Fracture Toughness Testing With Cracked Round Bars: Feasibility Study (NUREG/CR-6342)
          - Atmospheric Relative Concentrations in Building Wakes (NUREG/CR-6331, PNNL-10521, Revision 1)
          - Quality Assurance Inspections for Shipping and  Storage Containers (NUREG/CR-6314, INEL-95/0061)
          - Method for Performing Diversity and Defense-in-Depth Analyses of Reactor Protection Systems (NUREG/C
          - Mechanical Properties of Thermally Aged Cast  Stainless Steels from Shippingport Reactor Components 
          - Common-Cause Failure Database and Analysis  System: Event Data Collection, Classification, and Codin
          - Multidisciplinary Framework for Human Reliability Analysis with an Application to Errors of Commissi
          - Technical Guidelines for Aseismic Design of Nuclear Power Plants (NUREG/CR-6241, BNL-NUREG-52422)
          - Radioanalytical Technology for 10 CFR Part 61  and Other Selected Radionuclides: Literature Review (
          - An Assessment of Fire Vulnerability for Aged Electrical Relays (NUREG/CR-6220,  SAND94-0769)
          - Production and Testing of the Revised VITAMIN-B6 Fine-Group and the BUGLE-93 Broad-Group Neutron/Pho
          - High-Temperature Hydrogen-Air- Steam Detonation  Experiments in the BNL Small-Scale Development Appa
          - Value of Public Health and Safety Actions and Radiation Dose Avoided (NUREG/CR-6212, BNL-NUREG-52413
          - A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments (NUREG/CR-6189, S
          - Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station – Effects 
          - A Summary of the Fire Testing Program at the  German HDR Test Facility (NUREG/CR-6173, SAND94-1795)
          - SCDAP/RELAP5/MOD 3.3 Code Manual (NUREG/CR-6150)
          - Drywell Debris Transport Study: Computational Work – Final Report (NUREG/CR-6369, Volume 3)
          - Drywell Debris Transport Study (NUREG/CR-6369)
          - The Effect of Initial Temperature on Flame Acceleration and Deflagration-to-Detonation Transition Ph
          - Owners of Nuclear Power Plants (NUREG/CR-6500, ORNL/TM-13297/R2, Revision 2)
          - Technical Basis for Environmental Qualification of Microprocessor-Based Safety-Related Equipment in 
          - Motor-Operated Valve (MOV) Actuator Motor and Gearbox Testing (NUREG/CR-6478, INEL-96/0219)
          - Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities (NUREG/CR-6477, PNNL-11209)
          - Circuit Bridging of Components by Smoke (NUREG/CR-6476, SAND96-2633)
          - Characterization of Flaws in U.S. Reactor Pressure Vessels (NUREG/CR-6471)
          - Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems – Final Report
          - Analysis of Spent Fuel Heatup Following Loss of Water in a Spent Fuel Pool, Final Report (NUREG/CR-6
          - Recommended Electromagnetic Operating Envelopes for Safety-Related I&C Systems in Nuclear Power Plan
          - Effects of Thermal Aging on Fracture Toughness  and Charpy-Impact Strength of Stainless Steel Pipe W
          - Assessment of the DCH Issue for Plants with Ice Condenser Containments (NUREG/CR-6427)
          - Report on Aging of Nuclear Power Plant Reinforced Concrete Structures (NUREG/CR-6424)
          - A Proposed Acceptance Process for Commercial  Off-the-Shelf (COTS) Software in Reactor Applications 
          - Self-Monitoring SurveillanceSystem for Prestressing Tendons: Phase I Small Business Innovation Rese
          - Nuclear Fuel Cycle Facility Accident Analysis   Handbook (NUREG/CR-6410)
          - Classification of Transportation Packaging and  Dry Spent Fuel Storage System Components According t
          - Environmental Testing of an Experimental Digital Safety Channel (NUREG/CR-6406)
          - Human Factors Engineering (HFE) Insights for Advanced Reactors Based Upon Operating Experience (NURE
          - Literature Review of Environmental Qualification of Safety-Related Electric Cables (NUREG/CR-6384)
          - Effects of Radionuclide Concentrations by Cement/Ground-Water Interactions in Support of Performance
          - Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Expert
          - Tensile-Property Characterization of Thermally Aged Cast Stainless Steels (NUREG/CR-6142, ANL-93/35)
          - MELCOR Computer Code Manuals (NUREG/CR-6119)
          - Evaluation of Exposure Limits to Toxic Gases for Nuclear Reactor Control Room Operations (NUREG/CR-5
          - Submergence and High Temperature Steam Testing of Class lE Electrical Cables (NUREG/CR-5655)
          - Incorporating Aging Effects into Probabilistic Risk Assessment — A Feasibility Study Utilizing Relia
          - The Impact of Thermal Aging on the Flammability of Electric Cables (NUREG/CR-5619,  SAND90-2121)
          - Electromagnetic Compatibility Testing for Conducted Susceptibility Along Interconnecting Signal Line
          - Heavy-Section Steel Irradiation Program: Progress Report April 1997 - March 1998 (NUREG/CR-5591, Vol
          - The High Level Vibration Test Program - Final Report (NUREG/CR-5585, BNL-NUREG-52240)
          - Evaluation of Generic Issue 57 (NUREG/CR-5580)
          - Dating and Earthquakes: Review of Quaternary Geochronology and Its Application to Paleoseismology (N
          - An Investigation of the Effects of Thermal Aging on the Fire Damageability of Electric Cables (NUREG
          - Models for Estimation of Service Life of Concrete Barriers in Low-Level Radioactive Waste Disposal (
          - A Performance Assessment Methodology for Low-Level Waste Facilities (NUREG/CR-5532, SAND90-0375)
          - Hydrogen-Air-Diluent Detonation Study for Nuclear Reactor Safety Analyses (NUREG/CR-5525, SAND89-239
          - Residual Radioactive Contamination From Decommissioning:  User's Manual DandD Version 2.1 (NUREG/CR-
          - Reliability Study (NUREG/CR-5500)
          - Service Life of Concrete (NUREG/CR-5466, NISTIR 89-4086)
          - A Review of the Three Mile Island-1 Probabilistic Risk Assessment (NUREG/CR-5457, EGG-2572)
          - Anchor Bolt Behavior and Strength During Earthquakes Prepared (NUREG/CR-5434)
          - Elements of an Approach to Performance-Based Regulatory Oversight (NUREG/CR-5392, SCIE-NRC-373-98)
          - Initial Assessment of the Mechanisms and Significance of Low-Temperature Embrittlement of Cast Stain
          - A Summary of Nuclear Power Plant Fire Safety  Research at Sandia National Laboratories, 1975–1987 (N
          - Recommendations for Resolution of Public Comments on USI A-40, “Seismic Design Criteria” (NUREG/CR-5
          - Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent Fuel Pools (NUREG/CR-5
          - Design, Instrumentation and Testing of a Steel Containment Vessel Model (NUREG/CR-5679)
          - Results of Field Studies at the Maricopa Environmental Monitoring Site, Arizona (NUREG/CR-5694)
          - Comparing Monitoring Strategies at the Maricopa Environmental Monitoring Site, Arizona (NUREG/CR-569
          - Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels (NUREG/C
          - PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions (NUREG/CR-6115)
          - Software Reliability and Safety in Nuclear Reactor Protection Systems (NUREG/CR–6101, UCRL–ID–114839
          - Aging, Loss-of-Coolant Accident (LOCA), and High  Potential Testing of Damaged Cables (NUREG/CR-6095
          - An Analysis of Operational Experience During Low Power and Shutdown and a Plan for Addressing Human 
          - The Programmable Logic Controller and Its  Application in Nuclear Reactor Systems (NUREG/CR-6090, UC
          - Reviewing Real-Time Performance of Nuclear  Reactor Safety Systems (NUREG/CR-6083, UCRL-ID-114565)
          - Data Communications (NUREG/CR-6082, UCRL-ID-114567)
          - Analysis of Crack Initiation and Growth in the High Level Vibration Test at Tadotsu (NUREG/CR-6078, 
          - Perspectives on Reactor Safety (NUREG/CR-6042, SAND93-0971, Revision 2)
          - Fire Modeling of the Heiss Dampf Reaktor  Containment (NUREG/CR-6017,  SAND93-0528)
          - A Simplified Model of Aerosol Removal by Containment Sprays (NUREG/CR-5966, SAND92-2689)
          - Evaluation of a Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Faciliti
          - Technical  Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and  Risk- Co
          - Risk Evaluation for a General Electric BWR,  Effects of Fire Protection System Actuation on Safety-R
          - Risk Evaluation for a B&W Pressurized Water  Reactor, Effects of Fire Protection System Actuation on
          - Risk Evaluation for a Westinghouse PWR, Effects  of Fire Protection System Actuation on Safety-Relat
          - Technical Bases for Regulatory Guide for Soil Liquefaction (NUREG/CR-5741)
          - Laboratory Investigations of Soils and Rocks For Engineering Analysis and Design of Nuclear Power Fa
          - Field Investigations for Foundations of Nuclear Power Facilities (NUREG/CR-5738)
          - Hydrogeologic Performance Assessment of the Commercial Low-Level Radioactive   Waste Disposal Facili
          - Evaluation of WF-70 Weld Metal From the Midland Unit 1 Reactor Vessel (NUREG/CR-5736)
          - Recommendations to the NRC on Acceptable Standard Format and Content for the Fundamental Nuclear Mat
          - Re-evaluation of Regulatory Guidance Provided in Regulatory Guides 1.142 and 1.143 (NUREG/CR-5733)
          - Iodine Chemical Forms in LWR Severe Accidents (NUREG/CR-5732)
          - Technical Basis for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related I&
          - Sulfate-Attack Resistance and Gamma-Irradiation Resistance of Some Portland Cement Based Mortars (NU
          - Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Bur
          - Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and I
          - A Review of Large-Scale Fracture Experiments Relevant to Pressure Vessel Integrity Under Pressurized
          - Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes (NUREG/CR-6697)
          - LAPUR 5.2 Verification and User's Manual (NUREG/CR-6696)
          - Hydrologic Uncertainty Assessment for Decommissioning Sites: Hypothetical Test Case Applications (NU
          - POLIDENT: A Module for Generating Continuous-Energy Cross Sections From ENDF Resonance Data (NUREG/C
          - Probabilistic Modules for the RESRAD and RESRAD-BUILD Computer Codes: User Guide (NUREG/CR-6692)
          - The Effects of Alarm Display, Processing, and Availability on Crew Performance (NUREG/CR-6691)
          - The Effects of Interface Management Tasks on Crew Performance and Safety in Complex, Computer-Based 
          - Proposed Approach for Reviewing Changes to Risk-Important Human Actions (NUREG/CR-6689)
          - Testing, Verifying, and Validating SAPHIRE Versions 6.0 and 7.0 (NUREG/CR-6688)
          - Irradiation-Assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys (NUREG/CR-
          - Experience With the Scale Criticality Safety Cross-Section Libraries (NUREG/CR-6686)
          - Pretest Analysis of a 1:4-Scale Prestressed Concrete Containment Vessel Model (NUREG/CR-6685)
          - Advanced Alarm Systems: Revision of Guidance and Its Technical Basis (NUREG/CR-6684)
          - A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup C
          - Summary and Categorization of Public Comments on Controlling the Disposition of Solid Materials (NUR
          - Ampacity Derating and Cable Functionality for Raceway Fire Barriers (NUREG/CR-6681)
          - Review Templates for Computer-Based Reactor Protection Systems (NUREG/CR-6680)
          - Assessment of Age-Related Degradation of Structures and Passive Components for U.S. Nuclear Power Pl
          - Pretest Round Robin Analysis of a Prestressed Concrete Containment Vessel Model (NUREG/CR-6678)
          - Evaluation Of Risk Associated With Intergranular Stress Corrosion Cracking In Boiling Water Reactor 
          - Limited Burnup Credit in Criticality Safety Analysis: A Comparison of ISG-8 and Current Internationa
          - Probabilistic Dose Analysis Using Parameter Distributions Developed For RESRAD and RESRAD-BUILD Code
          - Environmental Effects of Extending Fuel Burnup Above 60 Gwd/MTU (NUREG/CR-6703)
          - Effects of Deregulation on Safety:  Implications Drawn From The Aviation, Rail and United Kingdom Nu
          - Digital Systems Software Requirements Guidelines (NUREG/CR-6734)
          - A Baseline Risk-Informed, Performance-Based Approach for In Situ Leach Uranium Extraction Licensees 
          - Zinc-Zircaloy Interaction in Dry Storage Casks (NUREG/CR-6732)
          - TRAC-M/F77, Version 5.5 Developmental Assessment Manual (NUREG/CR-6730)
          - Field Studies for Estimating Uncertainties in Ground-Water Recharge Using Near-Continuous Peizometer
          - TRAC-M/FORTRAN 90 (Version 3.0) Programmer's Manual (NUREG/CR-6725)
          - TRAC-M/FORTRAN 90 Version 3.0 Theory Manual (NUREG/CR-6724)
          - TRAC-M/FORTRAN 90 (Version 3.0) User's Manual (NUREG/CR-6722)
          - Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion
          - TRAC-M Validation Test Matrix (NUREG/CR-6720)
          - Assessment of the Relevance of Displacement Bases Design Methods/Criteria to Nuclear Plant Structure
          - OPUS/PlotOPUS: An ORIGEN-S Post-Processing Utility and Plotting Program for SCALE (NUREG/CR-6718)
          - Environmental Effects on Fatigue Crack Initiation in Piping and Pressure Vessel Steels (NUREG/CR-671
          - Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Cask
          - Probability-Based Evaluation of Degraded Reinforced Concrete Components in Nuclear Power Plants (NUR
          - Hanford Tank Waste Remediation System Pretreatment Chemistry and Technology (NUREG/CR-6714)
          - Regulatory Analysis of Major Revision of 10 CFR Part 71, Final Rule (NUREG/CR-6713)
          - Summary and Categorization of Public Comments on the Major Revision of 10 CFR Part 71 (NUREG/CR-6712
          - Environmental Assessment of Major Revision of 10 CFR Part 71, Final Rule (NUREG/CR-6711)
          - Extending the Dynamic Flowgraph Methodology (DFM) to Model Human Performance and Team Effects (NUREG
          - Surface Complexation Modeling of Uranium (VI) Adsorption on Natural Mineral Assemblages (NUREG/CR-67
          - Seismic Analysis of a Reinforced Concrete Containment Vessel Model (NUREG/CR-6707)
          - Capacity of Steel & Concrete Containment Vessels with Corrosion Damage (NUREG/CR-6706)
          - Historical Case Analysis of Uranium Plume Attenuation (NUREG/CR-6705)
          - Assessment of Environmental Qualification Practices and Condition Monitoring Techniques for Low-Volt
          - Interaction of Zinc Vapor with Zircaloy and the Effect of Zinc Vapor on the Mechanical Properties of
          - Hydrogen Generation in TRU Waste Transportation Packages (NUREG/CR-6673)
          - Reexamination of Spent Fuel Shipment Risk Estimates – Main Report (NUREG/CR-6672)
          - Comparison of Irradiation-Induced Shifts of KJc and Charpy Impact Toughness for Reactor Pressure Ves
          - Guidance for Performing Probabilistic Seismic Hazard Analysis for a Nuclear Plant Site: Example Appl
          - Results and Insights on the Impact of Smoke on Digital Instrumentation and Control (NUREG/CR-6597)
          - An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events (N
          - The Effects of Surface Condition on an Ultrasonic Inspection: Engineering Studies Using Validated Co
          - Evaluation of the Hualien Quarter Scale Model Seismic Experiment (NUREG/CR-6584)
          - U.S. Nuclear Power Plant Operating Cost and Experience Summaries  (NUREG/CR-6577, Supplement 2)
          - Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA: Procedure Guides for a Probabilistic Risk Assess
          - Low-Level Radioactive Waste Classification, Characterization, and Assessment: Waste Streams and Neut
          - Uncertainty Analyses of Infiltration and Subsurface Flow and Transport for SDMP Sites (NUREG/CR-6565
          - Large-Scale Vibration  Tests of Main Steam and Feedwater Piping Systems With Conventional and  Energ
          - Finite Element Analyses for Seismic Shear Wall International Standard Problem (NUREG/CR-6554, BNL/NU
          - A Methodology for Analyzing Precursors to Earthquake-Initiated and Fire-Initiated Accident Sequences
          - Effects of Smoke on Functional Circuits (NUREG/CR-6543, SAND97-2544)
          - FRAPCON-3 Updates, Including Mixed-Oxide Fuel Properties  (NUREG/CR-6534, Volume 4)
          - Deliberate Ignition of Hydrogen-Air-Steam Mixtures in Condensing Steam Environments (NUREG/CR-6530, 
          - SecPop: Sector Population, Land Fraction, and Economic Estimation Program
          - The Effect of Lateral Venting on Deflagration-to-Detonation Transition in Hydrogen-Air-Steam Mixture
          - Steam Generator Tube Integrity Program  Report (NUREG/CR-6511)
          - Testing of dc-Powered Actuators for Motor-Operated Valves (NUREG/CR-6620, INEEL/EXT-99-00083)
          - Probabilistic Liquefaction Analysis (NUREG/CR-6622)
          - Vapor Explosions in a One-Dimensional Large Scale Geometry with Simulant Melts (NUREG/CR-6623)
          - Recommendations for Revision of Regulatory Guide 1.78 (NUREG/CR-6624)
          - Evaluation of Terminated Licenses Parts 30, 40, and 70:  The Terminated License Tracking System (NUR
          - Standard Review Plan for Training and Qualifications Plans for Security Personnel at Category I Fuel
          - Survey of Waste Solidification Process Technologies (NUREG/CR-6666)
          - Pressure and Leak-Rate Tests and Models for Predicting Failure of Flawed Steam Generator Tubes (NURE
          - KENO3D Visualization Tool for KENO V.a and KENO-VI Geometry Models (NUREG/CR-6662)
          - TRAC-M Programmer's Guide: Fortran 77 Version 5.5 (NUREG/CR-6658)
          - Information on Hydrologic Conceptual Models, Parameters, Uncertainty Analysis, and Data Sources for 
          - Sensitivity and Uncertainty Analyses Applied to Criticality Safety Validation (NUREG/CR-6655)
          - A Study of Air-Operated Valves in U.S. Nuclear Power Plants (NUREG/CR-6654)
          - Comparison of Estimated Ground-Water Recharge Using Different Temporal Scales of Field Data (NUREG/C
          - International Comparative Assessment Study of Pressurized Thermal Shock in Reactor Pressure Vessels 
          - PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices (NUREG/CR-
          - Adsorption and Desorption Behavior of Selected 10 CFR Part 61 Radionuclides From Ion Exchange Resin 
          - Reevaluation of Regulatory Guidance on Modal Response Combination Methods for Seismic Response Spect
          - Advanced NDE for Steam Generator Tubing (NUREG/CR-6638)
          - Human Systems Interface and Plant Modernization Process: Technical Basis and Human Factors Review Gu
          - Maintainability of Digital Systems: Technical Basis and Human Factors Review Guidance (NUREG/CR-6636
          - Soft Controls: Technical Basis and Human Factors Review Guidance (NUREG/CR-6635)
          - Computer-Based Procedure Systems: Technical Basis and Human Factors Review Guidance (NUREG/CR-6634)
          - Advanced Information Systems Design: Technical Basis and Human Factors Review Guidance (NUREG/CR-663
          - Solubility and Leaching of Radionuclides in Site Decommissioning Management Plan (SDMP) Slags (NUREG
          - Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annea
          - The Effects of Aging at 343°C on the Microstructure and Mechanical Properties of Type 308 Stainless 
          - The Role of Organic Complexants and Colloids in the Transport of Radionuclides by Groundwater (NUREG
          - Automated Seismic Event Monitoring System (NUREG/CR-6625, Addendum 1)
          - Environmental Assessment: San Bernadino National Wildlife Refuge Well 10 (NUREG/CR-6648)
          - FLAME Facility: The Effect of Obstacles and Transverse Venting on Flame Acceleration and Transition 
          - Individual Plant Examinations for External Events: Review Plan and Evaluation Criteria (NUREG/CR-526
          - A Computer Code for Fire Protection and Risk Analysis of Nuclear Plants (NUREG/CR-5233)
          - Fire Protection Research Program Corner Effects Tests (NUREG/CR-0833, SAND79-0966)
          - SCALE/TRITON Primer: A Primer for Light Water Reactor Lattice Physics Calculations (NUREG/CR-0741, O
          - Nuclear Power Plant Fire Protection - Fire-Hazards Analysis (Subsystems Study Task 4) (NUREG/CR-0654
          - Nuclear Power Plant Fire Protection - Ventilation (Subsystems Study Task 1) (NUREG/CR-0636)
          - A Preliminary Report on Fire Protection Research Program, Fire Barriers and Suppression (September 1
          - Nuclear  Power Plant Fire Protection —  Fire Detection (Subsystems Study Task 2) (NUREG/CR-0488, SAN
          - Nuclear  Power Plant Fire Protection —  Fire Barriers (Subsystems Study Task 3) (NUREG/CR-0468,  SAN
          - SCALE Ver 4.4: A Modular Code System for Performing Standardized Computer Analyses for Licensing Eva
          - Development and Verification of Fire Tests for Cable Systems and System Components: Quarterly Report
          - Accidental Vapor Phase Explosions on Transportation Routes Near Nuclear Power Plants: Final Report J
          - Manual of Respiratory Protection Against Airborne Radioactive Material (NUREG/CR-0041, Revision 1)
          - Environmental Assessment of Ionization Chamber Smoke Detectors Containing Am-241 (NUREG/CR-1156)
          - Probabilistic Models for the Behavior of Compartment Fires (NUREG/CR-2269)
          - Technical Basis for Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Co
          - Fire Risk Analysis for Nuclear Power Plants (NUREG/CR-2258)
          - A Study of the Implications of Applying Quantitative Risk Criteria in the Licensing of Nuclear Power
          - 2015
          - Index of Risk Exposure and Risk Acceptance Criteria (NUREG/CR-1930, BNL-NUREG-51339)
          - A Risk Comparison (NUREG/CR-1916, BNL-NUREG-51338)
          - Development and Testing of a Model for Fire Potential in Nuclear Power Plants  (NUREG/CR-1819)
          - Acceptance and Verification For Early Warning Fire Detection Systems (NUREG/CR-1798)
          - Data Base for Radioactive Waste Management (NUREG/CR-1759)
          - Technology, Safety and Costs of Decommissioning Reference Nuclear Research and Test Reactors (NUREG/
          - Electrical Insulators in a Reactor Accident Environment (NUREG/CR-1682, SAND80-1957)
          - A Characterization of Faults in the Appalachian Foldbelt  (NUREG/CR-1621)
          - Approaches to Acceptable Risk: A Critical Guide (NUREG/CR-1614)
          - Investigation of Distorted-Geometry Simulation of Pool Dynamics in Horizontal-Vent BWR Containments
          - Seismic Review Table (NUREG/CR-1429)
          - The NACOM Code for Analysis of Postulated Sodium Spray Fires in LMFBRs (NUREG/CR-1405, BNL-NUREG-511
          - Evaluation of Simulator Adequacy for the Radiation Qualification of Safety-Related Equipment (NUREG/
          - A Radioactive Waste Disposal Classification System (NUREG/CR-1005)
          - PRA  Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Powe
          - Investigation of Fire Stop Test Parameters Final Report  (NUREG/CR-2321)
          - Transient Fire Environment Cable Damageability Test Results (NUREG/CR-4638,  SAND86-0839)
          - Screening Tests of Representative Nuclear Power Plant Components Exposed to Secondary Environments C
          - Users' Guide for a Personal-Computer-Based Nuclear Power Plant Fire Data Base (NUREG/CR-4586, SAND86
          - Description and Testing of an Apparatus for Electrically Initiating Fires Through Simulation of a Fa
          - COMPBRN III: A Computer Code for Modeling Compartment Fires (NUREG/CR-4566, ORNL/TM-10005)
          - FIRAC User's Manual: A Computer Code to Simulate Fire Accidents in Nuclear Facilities (NUREG/CR-4561
          - Analysis of Diffusion Flame Tests (NUREG/CR-4534, SAND86-0419)
          - An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets  (
          - Design Features for Enhancing International  Safeguards of Away-from-Reactor Dry Storage for Spent L
          - Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems (NUREG
          - Recommendations To The Nuclear Regulatory Commission On Trial Guidelines For Seismic Margin Reviews 
          - The Use of a Field Model To Assess Fire Behavior in Complex Nuclear Power Plant Enclosures: Present 
          - Tornado Climatology of the Contiguous United States (NUREG/CR-4461, Revision 2; PNNL-15112, Revision
          - Comparison of Dynamic Characteristics of Fukushima Nuclear Power Plant Containment Building Determin
          - Update of Part 61 – Impacts Analysis Methodology: Codes and Example Problems (NUREG/CR-4370, Volume 
          - Review of Light Water Reactor Regulatory Requirements   (NUREG/CR-4330, PNL-5809)
          - Full-Scale Measurements of Smoke Transport and Deposition in Ventilation System   Ductwork (NUREG/CR
          - Investigation of Potential Fire-Related Damage to Safety-Related Equipment in Nuclear Power Plants (
          - Investigation of High-Efficiency Particulate Air Filter Plugging by Combustion Aerosols (NUREG/CR-42
          - Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two Representative Nuclear Power P
          - Tactical Training Reference Manual (NUREG/CR-5172, BMI-2166)
          - Steam Generator Tube Integrity Program/Steam Generator Group Project: Final Project Summary Report (
          - Tactical Exercise Planning Handbook (NUREG/CR-5081, BMI-2162)
          - Experimental Results Pertaining to the Performance of Thermal Igniters (NUREG/CR-5079, SAND87-3139)
          - An Approach to the Quantification of Seismic Margins in Nuclear Power Plants: The Importance of BWR 
          - Fire Environment Determination in the LaSalle Nuclear Power Plant Control Room (NUREG/CR-5037, BNL-N
          - Detonability of H2-Air-Diluent Mixtures  (NUREG/CR-4905,  SAND85-1263)
          - Interpretation of Bioassay Measurements (NUREG/CR-4884, BNL-NUREG-52063)
          - Development and Application of a Computer Model for Large-Scale Flame Acceleration Experiments (NURE
          - Procedures for the External Event Core Damage Frequency Analyses for NUREG-1150 (NUREG/CR-4840,  SAN
          - Methods for External Event Screening Quantification: Risk Methods Integration and Evaluation Program
          - MELCOR Validation and Verification: 1986 Papers (NUREG/CR-4830, Revision 3)
          - Mitigative Techniques for Ground-Water Contamination Associated With Severe Nuclear Accidents (NUREG
          - Shipping Container Response to Severe Highway and Railway Accident Conditions  (NUREG/CR-4829,  UCID
          - Seismic Margin Review  of the Maine Yankee Atomic Power Station (NUREG/CR-4826, UCID-20948, Vols. 1,
          - Guide for Preparing Operating Procedures for Shipping Packages (NUREG/CR-4775)
          - Combustion Aerosols Formed During Burning of Radioactively Contaminated Materials: Experimental Resu
          - Enclosure Environment Characterization Testing for the Base Line Validation of Computer Fire Simulat
          - Heat and Mass Release for Some Transient Fuel Source Fires: A Test Report (NUREG/CR-4680, SAND86-031
          - Quantitative Data on the Fire Behavior of Combustible Materials Found in Nuclear Power Plants: A Lit
          - Precursors to Potential Severe Core Damage Accidents: 1998  A Status Report  (NUREG/CR-4674, Volume
          - Environmentally Assisted Cracking in Light Water Reactors: Annual/SemiannualReports (NUREG/CR-4667)
          - User's Manual for FIRIN: A Computer Code to Estimate Accidental Fire and Airborne Releases in Nuclea
          - Nuclear Power Plant Electrical Cable Damageability Experiments (NUREG/CR-2927, SAND82-0236)
          - Radioactive Effluents from Nuclear Power Plants (NUREG/CR-2907)
          - Aging Effects on Fire-Retardant Additives in Organic Materials for Nuclear Plant Applications (NUREG
          - PAVAN: An Atmospheric-Dispersion Program for Evaluating Design-Basis Accidental Releases of Radioact
          - Probabilistic Safety Analysis Procedures Guide (NUREG/CR-2815, BNL-NUREG-51559)
          - Hydrogen Burn Survival: Preliminary Thermal Model and Test Results (NUREG/CR-2730, SAND82-1150)
          - Light Water Reactor Hydrogen Manual (NUREG/CR-2726, SAND82-1137)
          - Seismic Safety  Margins Research Program: Equipment Fragility Data Base (NUREG/CR-2680, UCRL-53038, 
          - Characteristics of Combustion Products: A Review of the Literature (NUREG/CR-2658,  PNL-4174)
          - Allowable Shipment Frequencies for the Transport of Toxic Gases Near Nuclear Power Plants  (NUREG/CR
          - Fire Protection Research Program for the U. S. Nuclear Regulatory Commission 1975-1981 (NUREG/CR-260
          - Hazards to Nuclear Power Plants from Large Liquefied Natural Gas (LNG) Spills on   Water (NUREG/CR-2
          - Final Results of the Hydrogen Igniter Experimental Program (NUREG/CR-2486)
          - Hydrogen Combustion Characteristics Related to Reactor  Accidents  (NUREG/CR-2475, BNL-NUREG-51492)
          - Burn Mode Analysis of Horizontal Cable Tray Fires (NUREG/CR-2431, SAND81-0079)
          - Requirements for Establishing Detector Siting Criteria in Fires Involving Electrical Materials (NURE
          - Age-Specific Inhalation Radiation Dose Commitment Factors for Selected Radionuclides (NUREG/CR-2384,
          - Test and Criteria for Fire Protection of Cable Penetrations (NUREG/CR-2377, SAND81-7160)
          - Potentially Damaging Failure Modes of High- and Medium-Voltage Electrical Equipment (NUREG/CR-3122, 
          - Scenarios and Analytical Methods for UF6 Releases at NRC-Licensed Fuel Cycle Facilities  NUREG/CR-31
          - Investigation of Twenty-Foot Separation Distance as a Fire Protection Method as Specified in 10 CFR 
          - COMPBRN — A Computer Code for Modeling Compartment Fires  (NUREG/CR-3239, UCLA-ENG-8257)
          - Evaluation of Available Data for Probabilistic Risk Assessments (PRA) of Fire Events at Nuclear Powe
          - Probability-Based Evaluation of Selected Fire Protection Features in Nuclear   Power Plants (NUREG/C
          - Evaluation of Current Methodology Employed in Probabilistic Risk Assessment (PRA) of Fire Events at 
          - Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests (NUREG/CR-4138, SAND85-0135)
          - Investigation of Cable and Cable System Fire Test Parameters (NUREG/CR-4112, US 75-1)
          - Extended Storage of Low-Level Radioactive Waste: Potential Problem Areas (NUREG/CR-4062, BNL-NUREG-5
          - Survey and Evaluation of System Interaction Events and Sources (NUREG/CR-3922, ORNL/NOAC-224)
          - Engineering Characterization of Ground Motion: Task II: Summary Report (NUREG/CR-3805, Volume 5)
          - Accident-Induced Flow and Material Transport in Nuclear Facilities: A Literature   Review (NUREG/CR-
          - Evaluation of Suppression Methods for Electrical Cable Fires (NUREG/CR-3656,  SAND83-2664)
          - Hydrogen-Steam Jet-Flame Facility and Experiments (NUREG/CR-3638,  SAND84-0060)
          - The Effect of Thermal and Irradiation Aging Simulation Procedures on Polymer Properties (NUREG/CR-36
          - Response of Rubber Insulation Materials to Monoenergetic Electron Irradiations (NUREG/CR-3532,  SAND
          - Material Transport Analysis for Accident-Induced Flow in Nuclear Facilities (NUREG/CR-3527,  LA-9913
          - Hydrogen-Burn Survival Experiments at Fully Instrumented Test Site (FITS) (NUREG/CR-3521,  SAND83-17
          - A Review of the Limerick Generating Station Severe  Accident Risk Assessment: Review of Core-Melt Fr
          - Hydrogen:Air:Steam Flammability Limits and Combustion Characteristics in the FITS Vessel (NUREG/CR-3
          - Measures of Risk Importance and Their Applications (NUREG/CR-3385,  BMI-2103)
          - Radiological Assessment: A Textbook on Environmental Dose Analyses (NUREG/CR-3332, ORNL-5968)
          - Vulnerability of Nuclear Power Plant Structures to Large External Fires (NUREG/CR-3330, SAND83-1178)
          - Status Report: Correlation of Electrical Cable Failure with Mechanical Degradation (NUREG/CR-3263, S
          - The Los Alamos National Laboratory/New Mexico State University Filter Plugging Test Facility: Descri
          - Detonation Calculations Using a Modified Version of CSQII: Examples for   Hydrogen-Air Mixtures (NUR
          - Literature Review and Assessment of Plant and Animal Transfer Factors Used in Performance Assessment
          - Materials Behavior in HTGR Environments (NUREG/CR-6824, ANL-02/37)
          - 2003
          - Collaborative Study of NUPEC Seismic Field Test Data for NPP Structures (NUREG/CR-6822, BNL-NUREG-71
          - Application of Surface Complexation Modeling to Describe Uranium (VI) Adsorption and Retardation at 
          - Common-Cause Failure Event Insights (NUREG/CR-6819, INEEL/EXT-99-00613)
          - Drop Test Results for the Combustion Engineering Model No. ABB-2901 Fuel Pellet Shipping Package (NU
          - A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code (NUREG/CR-6817, Rev. 1
          - Review and Assessment of Codes and Procedures for HTGR Components (NUREG/CR-6816, ANL-02/36)
          - Review of the Margins for ASME Code Fatigue Design Curve - Effects of Surface Roughness and Material
          - Final Report on Advanced Nondestructive Evaluation for Steam Generator Tubing for the Second Interna
          - Issues and Recommendations for Advancement of PRA Technology In Risk-Informed Decision Making (NUREG
          - Emerging Technologies in Instrumentation and Controls (NUREG/CR-6812)
          - Strategies for Application of Isotopic Uncertainties in Burnup Credit (NUREG/CR-6811)
          - Overpressurization Test of a 1:4-Scale Prestressed Concrete Containment Vessel Model (NUREG/CR-6810,
          - Posttest Analysis of the NUPEC/NRC 1:4 Scale Prestressed Concrete Containment Vessel Model (NUREG/CR
          - Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Per
          - Results of NRC-Sponsored Stellite 6 Aging & Friction Testing (NUREG/CR-6807)
          - MOV Stem Lubricant Aging Research (NUREG/CR-6806)
          - A Comprehensive Strategy of Hydrogeologic Modeling and Uncertainty Analysis for Nuclear Facilities a
          - Second U.S. Nuclear Regulatory Commission International Steam Generator Tube Integrity Research Prog
          - Recommendations for Shielding Evaluations for Transport & Storage Packages (NUREG/CR-6802)
          - Fracture Toughness and Crack Growth Rates of Irradiated Austenitic Stainless Steels (NUREG/CR-6826, 
          - Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses (NUREG/CR-6801)
          - Assessment of Internal Oxidation (IO) as a Mechanism for Submodes of Stress Corrosion Cracking (SCC)
          - Regulatory Effectiveness of Unresolved Safety Issue (USI) A-45, "Shutdown Decay Heat Removal Require
          - Comparison of Average Transport and Dispersion Among a Gaussian, a Two-Dimensional, and a Three-Dime
          - Hydrogen Effects on Air Oxidation of Zirlo Alloy (NUREG/CR-6851, ANL-04/14)
          - EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Final Report, (NUREG/CR-6850, EPRI 1
          - Analysis of In-Vessel Retention and Ex-Vessel Fuel Coolant Interaction for AP1000 (NUREG/CR-6849)
          - Preliminary Validation of a Methodology for Assessing Software Quality (NUREG/CR-6848)
          - Air Oxidation Kinetics for Zr-Based Alloys (NUREG/CR-6846)
          - Sensitivity Analysis Applied to the Validation of the 10 B Capture Reaction in Nuclear Fuel Casks (N
          - TRISO-Coated Particle Fuel Phenomenon Identification and Ranking Tables (PIRTs) for Fission Product 
          - Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants (NUREG/CR-
          - A Risk-Informed Basis for Establishing Non-Fixed Surface Contamination Limits for Spent Fuel Transpo
          - The Technical Basis for the NRC's Guidelines for External Risk Communication (NUREG/CR-6840)
          - 2003
          - Technical Basis for Regulatory Guidance for Assessing Exemption Requests from the Nuclear Power Plan
          - The Battelle Integrity of Nuclear Piping (BINP) Program Final Report (NUREG/CR-6837)
          - Comparing Ground-Water Recharge Estimates Using Advanced Monitoring Techniques and Models (NUREG/CR-
          - Effects of Fuel Failure on Criticality Safety and Radiation Dose for Spent Fuel Casks (NUREG/CR-6835
          - 2002
          - Formal Methods of Decision Analysis Applied to Prioritization of Research and Other Topics (NUREG/CR
          - Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage (NUREG/CR-6831)
          - Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs (NUREG/CR-680
          - 2003
          - Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the Takahama-3 Reactor (NUREG/CR-6798)
          - Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit (NUREG/CR-6761, ORNL/TM
          - Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit (NUREG/CR-6760)
          - Parametric Study of Effect of Control Rods for PWR Burnup Credit (NUREG/CR-6759)
          - Radionuclide-Chelating Agent Complexes in  Low-Level Radioactive Decontamination Waste: Stability, A
          - Large-Scale Molecular Dynamics Simulations of Metal Sorption onto the Basal Surfaces of Clay Mineral
          - Analysis of Potential for Jet-Impingement Erosion from Leaking Steam Generator Tubes During Severe A
          - Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario Using the Probab
          - Review of Industry Responses to NRC Generic Letter 97-06 on Degradation of Steam Generator Internals
          - Review of Findings for Human Performance Contribution to Risk in Operating Events (NUREG/CR-6753,  I
          - A Comparative Analysis of Special Treatment Requirements for Systems, Structures, and Components (SS
          - The Human Performance Evaluation Process: A Resource for Reviewing the Identification and Resolution
          - Performance of MOV Stem Lubricants at Elevated Temperature (NUREG/CR-6750)
          - Integrating Digital and Conventional Human-System Interfaces: Lessons Learned from a Control Room Mo
          - STARBUCS: A Prototypic SCALE Control Module for Automated Criticality Safety Analyses Using Burnup C
          - Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinide
          - Advanced Nondestructive Evaluation for Steam Generator Tubing (NUREG/CR-6746)
          - Dry Cask Storage Characterization Project - Phase 1: CASTOR V/21 Cask Opening and Examination (NUREG
          - Phenomenon Identification and Ranking Tables (PIRTs) for Loss-of-Coolant Accidents in Pressurized an
          - Phenomenon Identification and Ranking Tables (PIRTs) for Rod Ejection Accidents in Pressurized Water
          - Application of Microprocessor-Based Equipment in Nuclear Power Plants-Technical Basis for a Qualific
          - FRAPTRAN: NRC's Computer Code (NUREG/CR-6739)
          - Generic-Safety-Issue (GSI) 191 Technical Assessment (NUREG/CR-6762)
          - Aging Assessment of Safety-Related Fuses Used in Low- and Medium- Voltage Applications in Nuclear Po
          - Burnup Credit PIRT Report (NUREG/CR-6764)
          - Development of Technical Basis for Leak-Before-Break Evaluation Procedures (NUREG/CR-6765)
          - A Comparison of Three Round Robin Studies on ISI Reliability of Wrought Stainless Steel Piping (NURE
          - Evaluation of Aging and Environmental Qualification Practices for Power Cables Used in Nuclear Power
          - 2001
          - Behavior of PWR Reactor Coolant System Components, Other than Steam Generator Tubes, Under Severe Ac
          - Eddy Current Reliability Results from the Steam Generator Mock-up Analysis Round-Robin (NUREG/CR-679
          - Results From Pressure and Leak-Rate Testing of Laboratory-Degraded Steam Generator Tubing (NUREG/CR-
          - Evaluation of Aging and Qualification Practices for  Cable Splices Used in Nuclear Plants (NUREG/CR-
          - Mechanism and Estimation of Fatigue Crack Initiation in Austenitic Stainless Steels in LWR Environme
          - ANL/CANTIA:  A Computer Code for Steam Generator Integrity Assessments (NUREG/CR-6786)
          - Evaluation of Eddy Current Reliability from Steam Generator Mock-Up Round-Robin (NUREG/CR-6785)
          - Use of Computerized Microtomography to Examine the Relationships of Sorption Sites in Alluvial Soils
          - Structural Seismic Fragility Analysis of the Surry Containment (NUREG/CR-6783)
          - Comparison of U.S. Military and International Electromagnetic Compatibility Guidance (NUREG/CR-6782,
          - Fracture Analysis of Vessels — Oak Ridge FAVOR v04.1, Computer Code: Theory and Implementation of Al
          - Recommendations on the Credit for Cooling Time in PWR Burnup Credit Analyses (NUREG/CR-6781)
          - The Effects of Composition and Heat Treatment on Hardening and Embrittlement of Reactor Pressure Ves
          - Results and Analysis of The ASTM Round Robin On Reconstitution (NUREG/CR-6777)
          - 2002
          - Human Performance Characterization in the  Reactor Oversight Process (NUREG/CR-6775, INEEL/EXT-01-01
          - Validation on Failure & Leak-Rate Correlations for Stress Corrosion Cracks in Steam Generator Tubes 
          - GSI-191: Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries  (N
          - GSI-191:  Separate-Effects Characterization of Debris Transport in Water (NUREG/CR-6772)
          - GSI-191: The Impact of Debris Induced Loss of ECCS Recirculation on PWR Core Damage Frequency (NUREG
          - GSI-191: Thermal-Hydraulic Response of PWR Reactor Coolant System & Containments to Selected Acciden
          - Technical Basis for Revision of Regulatory Guidance on Design Ground Motions:   Development of Hazar
          - 2001
          - Evaluation of Hydrologic Uncertainty Assessments for Decommissioning Sites Using Complex and Simplif
          - Release of Radionuclides and Chelating Agents from Full-System Decontamination Ion-Exchange Resins (
          - Effects of Adsorption Constant Uncertainty on Containment Plume Migration: One- and Two-Dimensional 
          - 2004
          - Temperature Dependence of Weibull Stress Parameters: Studies Using the Euro-Material Similar to ASME
          - Assessment of Eddy Current Testing for the Detection of Cracks in Cast Stainless Steel Reactor Pipin
          - Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power P
          - Primer on Durability of Nuclear Power Plant Reinforced Concrete Structures - A Review of Pertinent F
          - Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for Application to Nuclear Powe
          - Assessment of Analysis Methods for Seismic Shear Wall Capacity Using JNES/NUPEC Multi-Axial Cyclic a
          - Non-destructive and Failure Evaluation of Tubing from a Retired Steam Generator (NUREG/CR-6924)
          - Expert Panel Report on Proactive Materials Degradation Assessment  (NUREG/CR-6923)
          - P-CARES: Probabilistic Computer Analysis for Rapid Evaluation of Structures (NUREG/CR-6922)
          - Crack Growth Rates in a PWR Environment of Nickel Alloys from the Davis-Besse and V.C. Summer Power 
          - Risk-Informed Assessment of Degraded Containment Vessels (NUREG/CR-6920)
          - Recommendations for Revision of Seismic Damping Values in Regulatory Guide 1.61 (NUREG/CR-6919)
          - VARSKIN: A Computer Code for Skin Contamination and Dosimetry Assessments (NUREG/CR-6918, Revision 4
          - Experimental Measurements of Pressure Drop Across Sump Screen Debris Beds in Support of Generic Safe
          - Hydraulic Transport of Coating Debris (NUREG/CR-6916)
          - Aluminum Chemistry in a Prototypical Post-Loss-of-Coolant-Accident, Pressurized-Water-Reactor Contai
          - Integrated Chemical Effects Test Project (NUREG/CR-6914)
          - Chemical Effects Head-Loss Research In Support of Generic Safety Issue 191 (NUREG/CR-6913)
          - GSI-191 PWR Sump Screen Blockage Chemical Effects Tests: Thermodynamic Simulations (NUREG/CR-6912)
          - Screen Penetration Test Report (NUREG/CR-6885,  LA-UR-04-5416)
          - Model Abstraction Techniques for Soil-Water Flow and Transport (NUREG/CR-6884)
          - The SPAR-H Human Reliability Analysis Method (NUREG/CR-6883, INL/EXT-05-00509)
          - Assessment of Wireless Technologies and Their Application at Nuclear Facilities (NUREG/CR-6882,  ORN
          - Soil and Groundwater Sample Characterization and Agricultural Practices for Assessing Food Chain Pat
          - Argonne Model Boiler Facility: Topical Report (NUREG/CR-6880, ANL-04/29)
          - Steam Generator Tube Integrity Issues:  Pressurization Rate Effects, Failure Maps, Leak Rate Correla
          - Effect of Material Heat Treatment on Fatigue Crack Initiation in Austenitic Stainless Steels in LWR 
          - Characterization and Head-Loss Testing of Latent Debris from Pressurized-Water-Reactor Containment B
          - Risk-Informed Assessment of Degraded Buried Piping  Systems in Nuclear Power Plants (NUREG/CR-6876, 
          - Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials (NUREG/CR-6875, ANL-04/08)
          - GSI-191: Experimental Studies of  Loss-of-Coolant-Accident-Generated Debris Accumulation and Head Lo
          - Corrosion Rate Measurements and Chemical Speciation of Corrosion Products Using Thermodynamic Modeli
          - Documentation and Applications of the Reactive  Geochemical Transport Model RATEQ (NUREG/CR-6871, Dr
          - Consideration of Geochemical Issues in Groundwater Restoration at Uranium In-Situ Leach Mining Facil
          - A Reliability Physics Model for Aging of Cable Insulation Materials (NUREG/CR-6869, BNL-NUREG-73676-
          - Small-Scale Experiments: Effects of Chemical Reactions on Debris-Bed Head Loss — A Subtask of GSI-19
          - Technical Basis for Regulatory Guidance on Lightning Protection in Nuclear Power Plants (NUREG/CR-68
          - Parametric Evaluation of Seismic Behavior of Freestanding Spent Fuel Dry Cask Storage Systems (NUREG
          - 2004
          - 2004
          - Barrier Integrity Research Program (NUREG/CR-6861, ANL-04/26)
          - An Assessment of Visual Testing (NUREG/CR-6860)
          - Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario (NUREG/CR-6886)
          - DORT/TORT Analysis of the Hatch Unit-1 Jet Pump Riser Brace Pad Neutron Dosimetry Measurements with 
          - Seismic Analysis of Simplified Piping Systems for the NUPEC Ultimate Strength Piping Test Program (N
          - Tests of Uranium (VI) Adsorption Models in a Field Setting (NUREG/CR-6911)
          - Alternative Conceptual Models for Assessing Food Chain Pathways in Biosphere Models (NUREG/CR-6910)
          - Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (NUREG/CR-6909)
          - Crack Growth Rates of Nickel Alloy Welds in a PWR Environment (NUREG/CR-6907)
          - Containment Integrity Research at Sandia National Laboratories - An Overview (NUREG/CR-6906)
          - Report of Experimental Results for the International Fire Model Benchmarking and Validation Exercise
          - Evaluation of the Broadband Impedance Spectroscopy Prognostic/Diagnostic Technique for Electric Cabl
          - Human Event Repository and Analysis (HERA) System (NUREG/CR-6903)
          - Effects of Insulation Debris on Throttle-Valve Flow Performance: A Subtask of GSI-191 (NUREG/CR-6902
          - RELAP5/MOD3.2.2 Gamma Assessment for Pressurized Thermal Shock Applications (NUREG/CR-6857)
          - Current State of Reliability Modeling Methodologies for Digital Systems and Their Acceptance Criteri
          - A Combined Analytical Study to Characterize Uranium Soil and Sediment Contamination: The Case of the
          - Assessment of Void Swelling in Austenitic Stainless Steel Core Internals (NUREG/CR-6897, ANL-04/28)
          - Assessment of Seismic Analysis Methodologies for Deeply Embedded Nuclear Power Plant Structures (NUR
          - Technical Review of On-Line Monitoring Techniques for Performance Assessment (NUREG/CR-6895)
          - Spent Fuel Transportation Package Response to the Caldecott Tunnel Fire Scenario (NUREG/CR-6894, PNN
          - Modeling Adsorption Processes: Issues in Uncertainty, Scaling, and Prediction (NUREG/CR-6893)
          - Irradiation-Assisted Stress Corrosion Cracking Behavior of Austenitic Stainless Steels Applicable to
          - Crack Growth Rates of Irradiated Austenitic Stainless Steel Weld Heat Affected Zone in BWR Environme
          - Reevaluation of Station Blackout Risk at Nuclear Power Plants (NUREG/CR-6890)
          - The Effect of Elevated Temperature on Concrete Materials and Structures —  A Literature Review (NURE
          - Materials Aging Issues and Aging Management for Extended Storage and Transportation of Spent Nuclear
          - Performance of Metal and Polymeric O-Ring Seals in Beyond-Design-Basis Temperature Excursions  (NURE
          - 2010
          - Experimental Studies of Reinforced Concrete Structures Under Multi-Directional Earthquakes and Desig
          - An Evaluation of Ultrasonic Phased Array Testing for Cast Austenitic Stainless Steel Pressurizer Sur
          - Rod Bundle Heat Transfer Facility – Steady-State Steam Cooling Experiments (NUREG/CR-7152)
          - Development of a Fault Injection-Based Dependability Assessment Methodology for Digital I&C Systems 
          - Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE) (NUREG/CR-715
          - Effects of Degradation on the Severe Accident Consequences for a PWR Plant with a Reinforced Concret
          - Confirmatory Battery Testing: The Use of Float Current Monitoring to Determine Battery State-of-Char
          - Nuclear Power Plant Security Assessment Guide (NUREG/CR-7145)
          - Laminar Hydraulic Analysis of a Commercial Pressurized Water Reactor Fuel Assembly (NUREG/CR-7144, S
          - Characterization of Thermal-Hydraulic and Ignition Phenomena in Prototypic, Full-Length Boiling Wate
          - Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Co
          - The U.S. Nuclear Regulatory Commission's Cyber Security Regulatory Framework for Nuclear Power React
          - Assessment of Current Test Methods for Post-LOCA Cladding Behavior  (NUREG/CR-7139, ANL-11/52)
          - 2009
          - Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion (NUREG/
          - Compensatory and Alternative Regulatory MEasures for Nuclear Power Plant FIRE Protection (CARMEN-FIR
          - The Estimation of Very-Low Probability Hurricane Storm Surges for Design and Licensing of Nuclear Po
          - Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR-60 Reactor (NU
          - New Source Term Model for the RESRAD-OFFSITE Code Version 3  (NUREG/CR-7127)
          - Human-Performance Issues Related to the Design and Operation of Small Modular Reactors (NUREG/CR-712
          - Validation of LAPUR 6.0 Code (NUREG/CR-7124)
          - A Literature Review of the Effects of Smoke from a  Fire on Electrical Equipment  (NUREG/CR-7123)
          - Radionuclide Behavior in Soils and Soil-to-Plant Concentration Ratios for Assessing Food Chain Pathw
          - Kerite Analysis in Thermal Environment of FIRE (KATE-Fire): Test Results – Final Report (NUREG/CR-71
          - 2007
          - Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire) (NUREG/CR-7100)
          - LAPUR 6.0 Benchmark Against Data from the GENESIS Facility (NUREG/CR-7047)
          - Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States
          - Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma C
          - Development of Quantitative Software Reliability Models for Digital Protection Systems of Nuclear Po
          - A Large Scale Validation of a Methodology for Assessing Software Reliability (NUREG/CR-7042)
          - Evaluation of JNES Equipment Fragility Tests for Use in Seismic Probabilistic Risk Assessments for U
          - Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8 (NUREG
          - Verification of RESRAD-OFFSITE (NUREG/CR-7038, ANL-10/27)
          - 2007
          - Analysis of Severe Roadway Accidents Involving Long Duration Fires (NUREG/CR-7035)
          - Analysis of Severe Railway Accidents Involving Long Duration Fires (NUREG/CR-7034)
          - Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Allo
          - Radionuclide Release from Slag and Concrete Waste Materials – Part 2: Relationship Between Laborator
          - 2011
          - Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis (NUREG/CR-7107)
          - A Framework for Low Power/Shutdown Fire PRA – Final Report (NUREG/CR-7114)
          - An Assessment of Ultrasonic Techniques for Far-Side Examinations of Austenitic Stainless Steel Pipin
          - A Summary of Aging Effects and Their Management in Reactor Spent Fuel Pools, Refueling Cavities, Tor
          - State-of-the-Art Reactor Consequence Analyses Project (NUREG/CR-7110)
          - An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses—Cr
          - Guidance on Developing Effective Radiological Risk Communication Messages: Effective Message Mapping
          - Expanded Materials Degradation Assessment (EMDA) (NUREG/CR-7153, Volume 1 - 5)
          - Guidance Document: Conducting Paleoliquefaction Studies for Earthquake Source Characterization (NURE
          - Correlation of Seismic Performance in Similar SSCs (Structures, Systems, and Components) (NUREG/CR-7
          - Results of Open Testing for the Program to Assess the Reliability of Emerging Nondestructive Techniq
          - Results of Blind Testing for the Program to Assess the Reliability of Emerging Nondestructive Techni
          - Development of A Statistical Testing Approach for Quantifying Safety-Related Digital System on Deman
          - Developing a Bayesian Belief Network Model for Quantifying the Probability of Software Failure of a 
          - Review of Spent Fuel Reprocessing and Associated Accident Phenomena (NUREG/CR-7232)
          - Modeling of Radionuclide Transport in Freshwater Systems Associated with Nuclear Power Plants (NUREG
          - Seismic Design Standards and  Calculational Methods in the United States and Japan (NUREG/CR-7230)
          - Testing to Evaluate Battery and Battery Charger Short Circuit Current Contributions to a Fault on th
          - Open Secondary Testing of Window-Type Current Transformers (NUREG/CR-7228)
          - 2015
          - Primary Water Stress Corrosion Cracking of High-Chromium, Nickel-Base Welds Near Dissimilar Metal We
          - Stability of Circumferential Flaws in Once-Through Steam Generator Tubes Under Thermal Loading Durin
          - Axial Moderator Density Distributions, Control Blade Usage, and Axial Burnup Distributions for Exten
          - Tsunami Hazard Assessment: Best Modeling Practices and State-of-the-Art Technology (NUREG/CR-7223)
          - Tsunami Hazard Assessment Based on Wave Generation, Propagation, and Inundation Modeling for the U.S
          - Integrating Model Abstraction into Subsurface Monitoring Strategies (NUREG/CR-7221)
          - SNAP/RADTRAD 4.0: Description of Models and Methods (NUREG/CR-7220)
          - Cladding Behavior during Postulated Loss-of-Coolant Accidents (NUREG/CR-7219, ANL-16/09)
          - Rod Bundle Heat Transfer Facility Two-Phase Mixture Level Swell and Uncovery Test Experiments Data R
          - Application of Automated Analysis Software to Eddy Current Inspection Data from Steam Generator Tube
          - Spent Fuel Pool Project Phase II: Pre-Ignition and Ignition Testing of a 1x4 Commercial 17x17 Pressu
          - Review of Exemptions and General Licenses for Fissile Material in 10 CFR 71 (NUREG/CR-7239, ORNL/TM-
          - Spent Fuel Pool Project Phase 1: Pre-Ignition and Ignition Testing of a Single Commercial 17x17 Pres
          - Impact of Operating Parameters on Extended BWR Burnup Credit (NUREG/CR-7240)
          - Response of Nuclear Power Plant Instrumentation Cables Exposed to Fire Conditions (NUREG/CR-7244, SA
          - A Preliminary Report on Fire Protection Research Program Fire Barriers and Fire Retardant Coatings T
          - Development and Verification of Fire Tests for Cable Systems and System Components: Quarterly Report
          - Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels (NUREG/C
          - Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-consi
          - Phenomenon Identification and Ranking Tables (PIRTs) for Power Oscillations Without Scram in Boiling
          - Solubility and Leaching of Radionuclides in Site Decommissioning Management Plan (SDMP) Soil and Pon
          - Combined Estimation of Hydrogeologic Conceptual Model and Parameter Uncertainty (NUREG/CR-6843, PNNL
          - Fracture Analysis of Vessels — Oak Ridge FAVOR, V04.1, Computer Code: User's Guide (NUREG/CR-6855, O
          - 2005
          - Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels and Alloy 690 from Hal
          - Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation — ARIANE and REBUS 
          - Technical Basis for Regulatory Guidance on Design-Basis Hurricane-Borne Missile Speeds for Nuclear P
          - Radionuclide Release from Slag and Concrete Waste Materials, Part I: Conceptual Models of Leaching f
          - SCALE/TRITON Primer: A Primer for Light Water Reactor Lattice Physics Calculations (NUREG/CR-7041, O
          - An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses—Is
          - BWR Anticipated Transients Without Scram in the MELLLA+ Expanded Operating Domain, Part 2: Sensitivi
          - Spent Fuel Transportation Package Response to the MacArthur Maze Fire Scenario, Final Report (NUREG/
          - Technical Manual and User’s Guide for MILDOS, Version 4.1 (NUREG/CR-7258, ANL/EVS-18/5)
          - MILDOS Version 4.1 Computational Verification Report (NUREG/CR-7259, ANL/EVS-19/3)
          - MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel (
          - Review of Probable Maximum Precipitation Procedures and Databases Used to Develop Hydrometeorologica
          - Effects of Environmental Conditions on Manual Actions for Flood Protection and Mitigation (NUREG/CR-
          - Enhancing Guidance for Evacuation Time Estimate Studies (NUREG/CR-7269)
          - NUREG/CR-7272
          - NUREG/CR-7210
          - NUREG/CR-7265
          - Validation of a Computational Fluid Dynamics Method Using Horizontal Dry Cask Simulator Data 
          - Application of Point Precipitation Frequency Estimates to Watersheds (NUREG/CR-7271, ORNL/SPR-2019/1
          - Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages (NUR
          - Jet Impingement in High-Energy Piping Systems (NUREG/CR-7275)
          - Developing a Technical Basis for Embedded Digital Devices and Emerging Technologies (NUREG/CR-7273)
          - User's Manual for RESRAD-OFFSITE Code Version 4 (NUREG/CR-7268)
          - Review of Radiation-Induced Concrete Degradation and Potential Implications for Structures Exposed t
          - Review of Accident Tolerant Fuel Concepts with Implications to Severe Accident Progression and Radio
          - Phenomena Identification Ranking Tables for Accident Tolerant Fuel Designs Applicable to Severe Acci
          - Radiation Evaluation Methodology for Concrete Structures (NUREG/CR-7281 ORNL/SPR-2020/1572)
          - Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications: Final (N
          - Regional Tectonics and Seismicity of Southwestern Iowa: Final Report: 1978-1982 (NUREG/CR-3021)
          - Nonradiological Health Consequences from Evacuation and Relocation (NUREG/CR-7285)
          - The Price-Anderson Act: 2021 Report to Congress, Public Liability Insurance and Indemnity Requiremen
          - Exploring Advanced Computational Tools and Techniques with Artificial Intelligence and Machine Learn
          - Research to Develop Flood Barrier Testing Strategies for Nuclear Power Plants (NUREG/CR-7279, INL/EX
          - Nuclear Data Assessment for Advanced Reactors (NUREG/CR-7289, ORNL/TM-2021/2002)
          - Assessments on Eddy Current Detection of Cracking Near Volumetric Indications in Steam Generator Tub
          - Human Factors in Nondestructive Examination (NUREG/CR-7295, PNNL-32505)
          - Evaluation of In-Service Radon Barriers over Uranium Mill Tailings Disposal Facilities (NUREG/CR-728
          - Aging Management and Performance of Stainless Steel Bellows in Nuclear Power Plants (NUREG/CR-6726, 
          - Reactor Pressure Vessel Fluence Evaluation Methodology for Extended Beltline Locations (NUREG/CR-728
          - Application of Radar-Rainfall Estimates to Probable Maximum Precipitation in the Carolinas (NUREG/CR
          - Basis for Technical Guidance to Evaluate Evapotranspiration Covers (NUREG-CR/7297)
          - Multi-Mechanism Flood Hazard Assessment: Critical Review of Current Practice and Approaches and Exam
          - Technical Bases for Consequence Analyses Using MACCS (MELCOR Accident Consequence Code System) (NURE
          - Fuel Qualification for Molten Salt Reactors (NUREG/CR-7299, ORNL/TM-2022/2754)
          - State-of-the-Art Reactor Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Short
          - Ultrasonic Modeling and Simulation for Nuclear Nondestructive Evaluation (NUREG/CR-7301, PNNL-33732)
          - SCALE 6.2 Lattice Physics Performance Assessment (NUREG/CR-7284, ORNL/TM-2017/278)
          - Baseline Risk Index for Initiating Events (BRIIE) (NUREG/CR-6932, INL/EXT-06-11950)
          - Primary Water Stress Corrosion Cracking of High-Chromium Nickel-Base Welds – 2018 (NUREG/CR-7276)
          - Primary Water Stress Corrosion Cracking of High-Chromium Nickel-Base Welds at or Near Interfaces – 2
          - Synthesis of Extreme Storm Rainfall and Probable Maximum Precipitation in the Southeastern U.S. Pilo
          - Convection-Permitting Modeling for Intense Precipitation Processes (NUREG/CR-7290)
          - Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses (NUREG/CR
          - Numerical Modeling of Local Intense Precipitation Processes (NUREG/CR-7287)
          - Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, "Fuel Qualification for Advanced Rea
          - Radiation Accident Dose and Simulated Loss-of-Coolant Accident Test of Low Voltage Cables (NUREG/CR-
          - Geochemical Studies of Commercial Low-Level Radioactive Waste Disposal Sites - Topical Report (NUREG
          - Validating Actinides and Fission Products for Burnup Credit Criticality Safety Analyses - Nuclide Co
          - Evaluating Flaw Detectability Under Limited-Coverage Conditions (NUREG/CR-7304, PNNL-34367)
          - A Methodology for Allocating Nuclear Power Plant Control Nuclear to Human or Automatic Control (NURE
          - Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water R

#### Spotlight

#### Choose a Section

                - The Commission
                - License Renewal Generic Environmental Review
                - Job Openings
                - Palisades Restart
                - Environmental Justice 
                - Pilgrim Water Releases
                - Monticello Tritium Issue
                - International Strategy 2021-2025
                - Tribal Liaison Program
                - MAP-X
                - Transformation at the NRC - Updated
                - OIG Hotline
                - Rulemaking Activity
                - Commission Documents
                - Advanced Reactors
                - Nuclear Security
                - Strategic Plan

Page Last Reviewed/Updated Thursday, March 25, 2021