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

File	Title
Volume 1	Phenomena Identification and Ranking Technique (PIRT) Exercise for Ranking Low-Power Shutdown Plant Operating States and Outage Types. Chapters 1 – 7
Volume 2	Phenomena Identification and Ranking Technique (PIRT) Exercise for Ranking Low-Power Shutdown Plant Operating States and Outage Types. Appendices A – K

Page Last Reviewed/Updated Thursday, March 25, 2021

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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 Oxic Condition
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-Water Contamin
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 Surveillance System 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

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

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