Section 1. TMI Action Plan Items ( )
- Description
- Task I.A: OPERATING PERSONNEL: Operating Personnel and Staffing: Shift Technical Advisor
- Task I.A.2: OPERATING PERSONNEL: Training and Qualifications of Operating Personnel: Immediate Upgrading of Operator and Senior Operator Training and Qualifications: Qualifications - Experience
- Task I.A.3: OPERATING PERSONNEL: Licensing and Requalification of Operating Personnel: Revise Scope of Criteria for Licensing Examinations
- Task I.A.4: OPERATING PERSONNEL: Initial Simulator Improvement: Short-Term Study of Training Simulators
- Task I.B: SUPPORT PERSONNEL: Management for Operations: Organization and Management Long-Term Improvements: Prepare Draft Criteria
- Task I.B.2: SUPPORT PERSONNEL: Inspection of Operating Reactors: Revise OIE Inspection Program: Verify the Adequacy of Management and Proced
- Task I.C: OPERATING PROCEDURES: Short-Term Accident Analysis and Procedures Revision: Small Break LOCAs
- Task I.D: CONTROL ROOM DESIGN: Control Room Design Reviews
- Task I.E: ANALYSIS AND DISSEMINATION OF OPERATING EXPERIENCE: Office for Analysis and Evaluation of Operational Data
- Task I.F: QUALITY ASSURANCE: Expand QA List
- Task I.G: PREOPERATIONAL AND LOW-POWER TESTING: Training Requirements
- Task II.A: SITING: Siting Policy Reformulation
- Task II.B: CONSIDERATION OF DEGRADED OR MELTED CORES IN SAFETY REVIEW: Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation
- Task II.C: RELIABILITY ENGINEERING AND RISK ASSESSMENT: Interim Reliability Evaluation Program
- Task II.D: REACTOR COOLANT SYSTEM RELIEF AND SAFETY VALVES: Testing Requirements
- Task II.E: SYSTEM DESIGN: Auxiliary Feedwater System: Auxiliary Feedwater System Evaluation
- Task II.E.2: SYSTEM DESIGN: Emergency Core Cooling System: Reliance on ECCS
- Task II.E.3: SYSTEM DESIGN: Decay Heat Removal: Reliability of Power Supplies for Natural Circulation
- Task II.E.4: SYSTEM DESIGN: Containment Design: Dedicated Penetrations
- Task II.E.5: SYSTEM DESIGN: Design Sensitivity of B&W Reactors: Design Evaluation
- Task II.E.6: SYSTEM DESIGN: In Situ Testing of Valves: Test Adequacy Study
- Task II.F: INSTRUMENTATION AND CONTROLS: Additional Accident Monitoring Instrumentation
- Task II.G: ELECTRICAL POWER: Power Supplies for Pressurizer Relief Valves, Block Valves, and Level Indicators
- Task II.H: TMI-2 CLEANUP AND EXAMINATION: Maintain Safety of TMI-2 and Minimize Environmental Impact
- Task II.J: GENERAL IMPLICATIONS OF TMI FOR DESIGN AND CONSTRUCTION ACTIVITIES: Vendor Inspection Program: Establish a Priority System for C
- Task II.J.2: GENERAL IMPLICATIONS OF TMI FOR DESIGN AND CONSTRUCTION ACTIVITIES: Construction Inspection Program: Reorient Construction Inspe
- Task II.J.3: GENERAL IMPLICATIONS OF TMI FOR DESIGN AND CONSTRUCTION ACTIVITIES: Construction Inspection Program: Assign Resident Inspectors to All Construction Sites
- Task II.J.4: GENERAL IMPLICATIONS OF TMI FOR DESIGN AND CONSTRUCTION ACTIVITIES: Revise Deficiency Reporting Requirements: Revise Deficiency Reporting Requirements
- Task II.K: MEASURES TO MITIGATE SMALL-BREAK LOSS-OF-COOLANT ACCIDENTS AND LOSS-OF-FEEDWATER ACCIDENTS: IE Bulletins: Review TMI-2 PNs and D
- Task III.A: EMERGENCY PREPAREDNESS AND RADIATION EFFECTS: Improve Licensee Emergency Preparedness - Short-Term: Upgrade Emergency Preparedness: Implement Action Plan Requirements for Promptly Improving Licensee Emergency Preparedness
- Task III.A.2: EMERGENCY PREPAREDNESS AND RADIATION EFFECTS: Improving Licensee Emergency Preparedness - Long-Term: Amend 10 CFR 50 and 10 CFR
- Task III.A.3: EMERGENCY PREPAREDNESS AND RADIATION EFFECTS: Improving NRC Emergency Preparedness: NRC Role in Responding to Nuclear Emergencie
- Task III.B: EMERGENCY PREPAREDNESS OF STATE AND LOCAL GOVERNMENTS: Transfer of Responsibilities to FEMA
- Task III.C: PUBLIC INFORMATION: Have Information Available for the News Media and the Public: Review Publicly Available Documents
- Task III.D: RADIATION PROTECTION: Radiation Source Control: Primary Coolant Sources Outside the Containment Structure: Review Information Su
- Task III.D.2: RADIATION PROTECTION: Public Radiation Protection Improvement: Radiological Monitoring of Effluents: Evaluate the Feasibility and Perform a Value-Impact Analysis of Modifying Effluent-Monitoring Design Criteria
- Task III.D.3: RADIATION PROTECTION: Worker Radiation Protection Improvement: Radiation Protection Plans
- Task IV.A: STRENGTHEN ENFORCEMENT PROCESS: Seek Legislative Authority
- Task IV.B: ISSUANCE OF INSTRUCTIONS AND INFORMATION TO LICENSEES: Revise Practices for Issuance of Instructions and Information to Licensees
- Task IV.C: EXTEND LESSONS LEARNED TO LICENSED ACTIVITIES OTHER THAN POWER REACTORS: Extend Lessons Learned from TMI to Other NRC Programs
- Task IV.D: NRC STAFF TRAINING: NRC Staff Training
- Task IV.E: SAFETY DECISION-MAKING: Expand Research on Quantification of Safety Decision-Making
- Task IV.F: FINANCIAL DISINCENTIVES TO SAFETY: Increased OIE Scrutiny of the Power-Ascension Test Program
- Task IV.G: IMPROVE SAFETY RULEMAKING PROCEDURES: Develop a Public Agenda for Rulemaking
- Task IV.H: NRC PARTICIPATION IN THE RADIATION POLICY COUNCIL: NRC Participation in the Radiation Policy Council
- Task V.A: DEVELOPMENT OF SAFETY POLICY: Develop NRC Policy Statement on Safety
- Task V.B: POSSIBLE ELIMINATION OF NONSAFETY RESPONSIBILITIES: Study and Recommend, as Appropriate, Elimination of Nonsafety Responsibiliti
- Task V.C: ADVISORY COMMITTEES: Strengthen the Role of Advisory Committee on Reactor Safeguards
- Task V.D: LICENSING PROCESS: Improve Public and Intervenor Participation in the Hearing Process
- Task V.E: LEGISLATIVE NEEDS: Study the Need for TMI-Related Legislation
- Task V.F: ORGANIZATION AND MANAGEMENT: Study NRC Top Management Structure and Process
- Task V.G: CONSOLIDATION OF NRC LOCATIONS: Achieve Single Location, Long-Term
Section 2. Task Action Plan Items ( )
- Description
- Item A-1: Water Hammer (former USI)
- Item A-2: Asymmetric Blowdown Loads on Reactor Primary Coolant Systems (former USI)
- Item A-3: Westinghouse Steam Generator Tube Integrity (former USI)
- Item A-4: CE Steam Generator Tube Integrity (former USI)
- Item A-7: Mark I Long-Term Program (former USI)
- Item A-6: Mark I Short-Term Program (former USI)
- Item A-7: Mark I Long-Term Program (former USI)
- Item A-8: Mark II Containment Pool Dynamic Loads Long-Term Program (former USI)
- Item A-9: ATWS (former USI)
- Item A-10: BWR Feedwater Nozzle Cracking (former USI)
- Item A-11: Reactor Vessel Materials Toughness (former USI)
- Item A-12: Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports (former USI)
- Item A-13: Snubber Operability Assurance
- Item A-14: Flaw Detection
- Item A-15: Primary Coolant System Decontamination and Steam Generator Chemical Cleaning
- Item A-16: Steam Effects on BWR Core Spray Distribution
- Item A-17: Systems Interactions in Nuclear Power Plants (former USI)
- Item A-18: Pipe Rupture Design Criteria
- Item A-19: Digital Computer Protection System
- Item A-20: Impacts of the Coal Fuel Cycle
- Item A-21: Main Steamline Break Inside Containment - Evaluation of Environmental Conditions for Equipment Qualification
- Item A-22: PWR Main Steamline Break - Core, Reactor Vessel and Containment Building Response
- Item A-23: Containment Leak Testing
- Item A-24: Qualification of Class 1E Safety-Related Equipment (former USI)
- Item A-25: Non-Safety Loads on Class 1E Power Sources
- Item A-27: Reload Applications
- Item A-28: Increase in Spent Fuel Pool Storage Capacity
- Item A-29: Nuclear Power Plant Design for the Reduction of Vulnerability to Industrial Sabotage
- Item A-30: Adequacy of Safety-Related DC Power Supplies
- Item A-31: RHR Shutdown Requirements (former USI)
- Item A-32: Missile Effects
- Item A-33: NEPA Review of Accident Risks
- Item A-34: Instruments for Monitoring Radiation and Process Variables During Accidents
- Item A-35: Adequacy of Offsite Power Systems
- Item A-36: Control of Heavy Loads Near Spent Fuel (former USI)
- Item A-37: Turbine Missiles
- Item A-38: Tornado Missiles
- Item A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits (former USI)
- Item A-40: Seismic Design Criteria (former USI)
- Item A-41: Long-Term Seismic Program
- Item A-42: Pipe Cracks in Boiling Water Reactors (former USI)
- Item A-43: Containment Emergency Sump Performance (former USI)
- Item A-44: Station Blackout (former USI)
- Item A-45: Shutdown Decay Heat Removal Requirements (former USI)
- Item A-46: Seismic Qualification of Equipment in Operating Plants (former USI)
- Item A-47: Safety Implications of Control Systems (former USI)
- Item A-48: Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment
- Item A-49: Pressurized Thermal Shock (former USI)
- Item B-1: Environmental Technical Specifications
- Item B-3: Event Categorization
- Item B-4: ECCS Reliability
- Item B-5: Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containments
- Item B-6: Loads, Load Combinations, Stress Limits
- Item B-7: Secondary Accident Consequence Modeling
- Item B-8: Locking Out of ECCS Power Operated Valves
- Item B-9: Electrical Cable Penetrations of Containment
- Item B-10: Behavior of BWR Mark III Containments
- Item B-11: Subcompartment Standard Problems
- Item B-12: Containment Cooling Requirements (Non-LOCA)
- Item B-13: Marviken Test Data Evaluation
- Item B-14: Study of Hydrogen Mixing Capability in Containment Post-LOCA
- Item B-15: CONTEMPT Computer Code Maintenance
- Item B-16: Protection Against Postulated Piping Failures in Fluid Systems Outside Containment
- Item B-17: Criteria for Safety-Related Operator Actions
- Item B-18: Vortex Suppression Requirements for Containment Sumps
- Item B-19: Thermal-Hydraulic Stability
- Item B-20: Standard Problem Analysis
- Item B-21: Core Physics
- Item B-22: LWR Fuel
- Item B-23: LMFBR Fuel
- Item B-121: Seismic Qualification of Electrical and Mechanical Equipment
- Item B-25: Piping Benchmark Problems
- Item B-26: Structural Integrity of Containment Penetrations
- Item B-27: Implementation and Use of Subsection NF
- Item B-28: Radionuclide/Sediment Transport Program
- Item B-29: Effectiveness of Ultimate Heat Sinks
- Item B-30: Design Basis Floods and Probability
- Item B-31: Dam Failure Model
- Item B-32: Ice Effects on Safety-Related Water Supplies
- Item B-33: Dose Assessment Methodology
- Item B-34: Occupational Radiation Exposure Reduction
- Item B-35: Confirmation of Appendix I Models for Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light Water Cooled Power Reactors
- Item B-36: Develop Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units for Engineer
- Item B-37: Chemical Discharges to Receiving Waters
- Item B-38: Reconnaissance Level Investigations
- Item B-142: Transmission Lines
- Item B-40: Effects of Power Plant Entrainment on Plankton
- Item B-41: Impacts on Fisheries
- Item B-42: Socioeconomic Environmental Impacts
- Item B-43: Value of Aerial Photographs for Site Evaluation
- Item B-44: Forecasts of Generating Costs of Coal and Nuclear Plants
- Item B-45: Need for Power - Energy Conservation
- Item B-46: Cost of Alternatives in Environmental Design
- Item B-47: Inservice Inspection of Supports-Classes 1, 2, 3, and MC Components
- Item B-48: BWR Control Rod Drive Mechanical Failures
- Item B-49: Inservice Inspection Criteria and Corrosion Prevention Criteria for Containments
- Item B-50: Post-Operating Basis Earthquake Inspection
- Item B-51: Assessment of Inelastic Analysis Techniques for Equipment and Components
- Item B-52: Fuel Assembly Seismic and LOCA Responses
- Item B-53: Load Break Switch
- Item B-54: Ice Condenser Containments
- Item B-55: Improved Reliability of Target Rock Safety Relief Valves
- Item B-56: Diesel Reliability
- Item B-57: Station Blackout
- Item B-58: Passive Mechanical Failures
- Item B-59: (N-1) Loop Operation in BWRs and PWRs
- Item B-60: Loose Parts Monitoring Systems
- Item B-61: Allowable ECCS Equipment Outage Periods
- Item B-62: Reexamination of Technical Bases for Establishing SLs, LSSSs, and Reactor Protection System Trip Functions
- Item B-63: Isolation of Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary
- Item B-64: Decommissioning of Reactors
- Item B-65: Iodine Spiking
- Item B-66: Control Room Infiltration Measurements
- Item B-67: Effluent and Process Monitoring Instrumentation
- Item B-68: Pump Overspeed During LOCA
- Item B-69: ECCS Leakage Ex-Containment
- Item B-70: Power Grid Frequency Degradation and Effect on Primary Coolant Pumps
- Item B-71: Incident Response
- Item B-72: Health Effects and Life Shortening from Uranium and Coal Fuel Cycles
- Item B-73: Monitoring for Excessive Vibration Inside the Reactor Pressure Vessel
- Item C-1: Assurance of Continuous Long-Term Capability of Hermetic Seals on Instrumentation and Electrical Equipment
- Item C-2: Study of Containment Depressurization by Inadvertent Spray Operation to Determine Adequacy of Containment External Design Pressu
- Item C-3: Insulation Usage Within Containment
- Item C-4: Statistical Methods for ECCS Analysis
- Item C-5: Decay Heat Update
- Item C-6: LOCA Heat Sources
- Item C-7: PWR System Piping
- Item C-8: Main Steam Line Leakage Control Systems
- Item C-9: RHR Heat Exchanger Tube Failures
- Item C-10: Effective Operation of Containment Sprays in a LOCA
- Item C-11: Assessment of Failure and Reliability of Pumps and Valves
- Item C-12: Primary System Vibration Assessment
- Item C-13: Non-Random Failures
- Item C-14: Storm Surge Model for Coastal Sites
- Item C-15: NUREG Report for Liquid Tank Failure Analysis
- Item C-16: Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling System Selection
- Item C-17: Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes
- Item D-1: Advisability of a Seismic Scram
- Item D-2: Emergency Core Cooling System Capability for Future Plants
- Item D-3: Control Rod Drop Accident
Section 3. New Generic Issues ( )
- Description
- Issue 1: Failures in Air-Monitoring, Air-Cleaning, and Ventilating Systems
- Issue 2: Failure of Protective Devices on Essential Equipment
- Issue 3: Set Point Drift in Instrumentation
- Issue 4: End-of-Life and Maintenance Criteria
- Issue 5: Design Check and Audit of Balance-of-Plant Equipment
- Issue 6: Separation of Control Rod from Its Drive and BWR High Rod Worth Events
- Issue 7: Failures Due to Flow-Induced Vibrations
- Issue 8: Inadvertent Actuation of Safety Injection in PWRs
- Issue 9: Reevaluation of Reactor Coolant Pump Trip Criteria
- Issue 10: Surveillance and Maintenance of TIP Isolation Valves and Squib Charges
- Issue 11: Turbine Disc Cracking
- Issue 12: BWR Jet Pump Integrity
- Issue 13: Small-Break LOCA from Extended Overheating of Pressurizer Heaters
- Issue 14: PWR Pipe Cracks
- Issue 15: Radiation Effects on Reactor Vessel Supports
- Issue 16: BWR Main Steam Isolation Valve Leakage Control Systems
- Issue 17: Loss of Offsite Power Subsequent to a LOCA
- Issue 18: Steam Line Break with Consequential Small LOCA
- Issue 19: Safety Implications of Nonsafety Instrument and Control Power Supply Bus
- Issue 20: Effects of Electromagnetic Pulse on Nuclear Power Plants
- Issue 21: Vibration Qualification of Equipment
- Issue 22: Inadvertent Boron Dilution Events
- Issue 23: Reactor Coolant Pump Seal Failures
- Issue 24: Automatic ECCS Switchover to Recirculation
- Issue 25: Automatic Air Header Dump on BWR Scram System
- Issue 26: Diesel Generator Loading Problems Related to SIS Reset on Loss of Offsite Power
- Issue 27: Manual vs. Automated Actions
- Issue 28: Pressurized Thermal Shock
- Issue 29: Bolting Degradation or Failure in Nuclear Power Plants
- Issue 30: Potential Generator Missiles - Generator Rotor Retaining Rings
- Issue 31: Natural Circulation Cooldown
- Issue 32: Flow Blockage in Essential Equipment Caused by Corbicula
- Issue 33: Correcting Atmospheric Dump Valve Opening Upon Loss of Integrated Control System Power
- Issue 34: RCS Leak
- Issue 35: Degradation of Internal Appurtenances in LWRs
- Issue 36: Loss of Service Water
- Issue 37: Steam Generator Overfill and Combined Primary and Secondary Blowdown
- Issue 38: Potential Recirculation System Failure as a Consequence of Ingestion of Containment Paint Flakes or Other Fine Debris
- Issue 39: Potential for Unacceptable Interaction Between the CRD System and Non-Essential Control Air System
- Issue 40: Safety Concerns Associated with Pipe Breaks in the BWR Scram System
- Issue 41: BWR Scram Discharge Volume Systems
- Issue 42: Combination Primary/Secondary System LOCA
- Issue 43: Reliability of Air Systems
- Issue 44: Failure of Saltwater Cooling System
- Issue 45: Inoperability of Instrumentation Due to Extreme Cold Weather
- Issue 46: Loss of 125 Volt DC Bus
- Issue 47: Loss of Offsite Power
- Issue 48: LCO for Class 1E Vital Instrument Buses in Operating Reactors
- Issue 49: Interlocks and LCOs for Redundant Class 1E Tie-Breakers
- Issue 50: Reactor Vessel Level Instrumentation in BWRs
- Issue 51: Proposed Requirements for Improving the Reliability of Open Cycle Service Water Systems
- Issue 52: SSW Flow Blockage by Blue Mussels
- Issue 53: Consequences of a Postulated Flow Blockage Incident in a BWR
- Issue 54: Valve Operator-Related Events Occurring During 1978, 1979, and 1980
- Issue 55: Failure of Class 1E Safety-Related Switchgear Circuit Breakers to Close on Demand
- Issue 56: Abnormal Transient Operating Guidelines as Applied to a Steam Generator Overfill Event
- Issue 57: Effects of Fire Protection System Actuation on Safety-Related Equipment
- Issue 58: Inadvertent Containment Flooding
- Issue 59: Technical Specification Requirements for Plant Shutdown when Equipment for Safe Shutdown is Degraded or Inoperable
- Issue 60: Lamellar Tearing of Reactor Systems Structural Supports
- Issue 61: SRV Line Break Inside the BWR Wetwell Airspace of Mark I and II Containments
- Issue 62: Reactor Systems Bolting Applications
- Issue 63: Use of Equipment Not Classified as Essential to Safety in BWR Transient Analysis
- Issue 64: Identification of Protection System Instrument Sensing Lines
- Issue 65: Probability of Core-Melt Due to Component Cooling Water System Failures
- Issue 66: Steam Generator Requirements
- Issue 67: Steam Generator Staff Actions: Integrity of Steam Generator Tube Sleeves
- Issue 68: Postulated Loss of Auxiliary Feedwater System Resulting from Turbine-Driven Auxiliary Feedwater Pump Steam Supply Line Rupture
- Issue 69: Make-Up Nozzle Cracking in B&W Plants MTEB
- Issue 70: PORV and Block Valve Reliability
- Issue 71: Failure of Resin Demineralizer Systems and Their Effects on Nuclear Power Plant Safety
- Issue 72: Control Rod Drive Guide Tube Support Pin Failures
- Issue 73: Detached Thermal Sleeves
- Issue 74: Reactor Coolant Activity Limits for Operating Reactors
- Issue 75: Generic Implications of ATWS Events at the Salem Nuclear Plant
- Issue 76: Instrumentation and Control Power Interactions
- Issue 77: Flooding of Safety Equipment Compartments by Back-Flow Through Floor Drains
- Issue 78: Monitoring of Fatigue Transient Limits for Reactor Coolant System
- Issue 79: Unanalyzed Reactor Vessel Thermal Stress During Natural Convection Cooldown
- Issue 80: Pipe Break Effects on Control Rod Drive Hydraulic Lines in the Drywells of BWR Mark I and II Containments
- Issue 81: Impact of Locked Doors and Barriers on Plant and Personnel Safety
- Issue 82: Beyond Design-Basis Accidents in Spent Fuel Pools
- Issue 83: Control Room Habitability
- Issue 84: CE PORVs
- Issue 85: Reliability of Vacuum Breakers Connected to Steam Discharge Lines Inside BWR Containments
- Issue 86: Long-Range Plan for Dealing with Stress Corrosion Cracking in BWR Piping
- Issue 87: Failure of HPCI Steam Line Without Isolation
- Issue 88: Earthquakes and Emergency Planning
- Issue 89: Stiff Pipe Clamps
- Issue 90: Technical Specifications for Anticipatory Trips ICSB
- Issue 91: Main Crankshaft Failures in Transamerica DeLaval Emergency Diesel Generators
- Issue 92: Fuel Crumbling During LOCA
- Issue 93: Steam Binding of Auxiliary Feedwater Pumps
- Issue 94: Additional Low Temperature Overpressure Protection for Light-Water Reactors
- Issue 95: Loss of Effective Volume for Containment Recirculation Spray
- Issue 96: RHR Suction Valve Testing
- Issue 97: PWR Reactor Cavity Uncontrolled Exposures
- Issue 98: CRD Accumulator Check Valve Leakage
- Issue 99: RCS/RHR Suction Line Valve Interlock on PWRs
- Issue 100: Once-Through Steam Generator Level
- Issue 101: BWR Water Level Redundancy
- Issue 102: Human Error in Events Involving Wrong Unit or Wrong Train
- Issue 103: Design for Probable Maximum Precipitation
- Issue 104: Reduction of Boron Dilution Requirements
- Issue 105: Interfacing Systems LOCA at LWRs
- Issue 106: Piping and Use of Highly Combustible Gases in Vital Areas
- Issue 107: Main Transformer Failures
- Issue 108: BWR Suppression Pool Temperature Limits
- Issue 109: Reactor Vessel Closure Failure
- Issue 110: Equipment Protective Devices on Engineered Safety Features
- Issue 111: Stress Corrosion Cracking of Pressure Boundary Ferritic Steels in Selected Environments
- Issue 112: Westinghouse RPS Surveillance Frequencies and Out-of-Service Times
- Issue 113: Dynamic Qualification Testing of Large Bore Hydraulic Snubbers
- Issue 114: Seismic-Induced Relay Chatter
- Issue 115: Enhancement of the Reliability of Westinghouse Solid State Protection System
- Issue 116: Accident Management
- Issue 117: Allowable Time for Diverse Simultaneous Equipment Outages
- Issue 118: Tendon Anchorage Failure
- Issue 119: Piping Review Committee Recommendations: Piping Rupture Requirements and Decoupling of Seismic and LOCA Loads
- Issue 120: On-Line Testability of Protection Systems
- Issue 121: Hydrogen Control for Large, Dry PWR Containments
- Issue 122: Davis-Besse Loss of All Feedwater Event of June 9, 1985: Short-Term Actions: Potential Inability to Remove Reactor Decay Heat: F
- Issue 123: Deficiencies in the Regulations Governing DBA and Single-Failure Criteria Suggested by the Davis-Besse Event of June 9, 1985
- Issue 124: Auxiliary Feedwater System Reliability
- Issue 125: Davis-Besse Loss of All Feedwater Event of June 9, 1985: Long-Term Actions: Availability of the Shift Technical Advisor
- Issue 126: Reliability of PWR Main Steam Safety Valves
- Issue 127: Maintenance and Testing of Manual Valves in Safety-Related Systems
- Issue 128: Electrical Power Reliability
- Issue 129: Valve Interlocks to Prevent Vessel Drainage During Shutdown Cooling
- Issue 130: Essential Service Water Pump Failures at Multiplant Sites
- Issue 131: Potential Seismic Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse-Designed Plants
- Issue 132: RHR System Inside Containment
- Issue 133: Update Policy Statement on Nuclear Plant Staff Working Hours
- Issue 134: Rule on Degree and Experience Requirement
- Issue 135: Steam Generator and Steam Line Overfill
- Issue 136: Storage and Use of Large Quantities of Cryogenic Combustibles On Site
- Issue 137: Refueling Cavity Seal Failure
- Issue 138: Deinerting of BWR Mark I and II Containments During Power Operations Upon Discovery of RCS Leakage or a Train of a Safety System
- Issue 139: Thinning of Carbon Steel Piping in LWRs
- Issue 140: Fission Product Removal Systems
- Issue 141: Large-Break LOCA With Consequential SGTR
- Issue 142: Leakage Through Electrical Isolators in Instrumentation Circuits
- Issue 143: Availability of Chilled Water Systems and Room Cooling
- Issue 144: Scram Without a Turbine/Generator Trip
- Issue 145: Actions to Reduce Common Cause Failures
- Issue 146: Support Flexibility of Equipment and Components
- Issue 147: Fire-Induced Alternate Shutdown/Control Room Panel Interactions
- Issue 148: Smoke Control and Manual Fire-Fighting Effectiveness
- Issue 149: Adequacy of Fire Barriers
- Issue 150: Overpressurization of Containment Penetrations
- Issue 151: Reliability of Anticipated Transient Without SCRAM Recirculation Pump Trip in BWRs
- Issue 152: Design Basis for Valves That Might Be Subjected to Significant Blowdown Loads
- Issue 153: Loss of Essential Service Water in LWRs
- Issue 154: Adequacy of Emergency and Essential Lighting
- Issue 155: More Realistic Source Term Assumptions
- Issue 156: Systematic Evaluation Program: Settlement of Foundations and Buried Equipment
- Issue 157: Containment Performance
- Issue 158: Performance of Power-Operated Valves Under Design Basis Conditions
- Issue 159: Qualification of Safety-Related Pumps While Running on Minimum Flow
- Issue 160: Spurious Actions of Instrumentation Upon Restoration of Power
- Issue 161: Use of Non-Safety-Related Power Supplies in Safety-Related Circuits
- Issue 162: Inadequate Technical Specifications for Shared Systems at Multiplant Sites When One Unit Is Shut Down
- Issue 163: Multiple Steam Generator Tube Leakage
- Issue 164: Neutron Fluence in Reactor Vessel
- Issue 165: Safety and Safety/Relief Valve Reliability
- Issue 166: Adequacy of Fatigue Life of Metal Components
- Issue 167: Hydrogen Storage Facility Separation
- Issue 168: Environmental Qualification of Electrical Equipment
- Issue 169: BWR MSIV Common Mode Failure Due to Loss of Accumulator Pressure
- Issue 170: Fuel Damage Criteria for High Burnup Fuel
- Issue 171: ESF Failure from LOOP Subsequent to a LOCA
- Issue 172: Multiple System Responses Program
- Issue 173: Spent Fuel Storage Pool: Operating Facilities
- Issue 174: Fastener Gaging Practices: SONGS Employees' Concern
- Issue 175: Nuclear Power Plant Shift Staffing
- Issue 176: Loss of Fill-Oil in Rosemount Transmitters
- Issue 177: Vehicle Intrusion at TMI
- Issue 178: Effect of Hurricane Andrew on Turkey Point
- Issue 179: Core Performance
- Issue 180: Notice of Enforcement Discretion
- Issue 181: Fire Protection
- Issue 182: General Electric Extended Power Uprate
- Issue 183: Cycle-Specific Parameter Limits in Technical Specifications
- Issue 184: Endangered Species
- Issue 185: Control of Recriticality Following Small-Break LOCA In PWRs
- Issue 186: Potential Risk and Consequences of Heavy Load Drops
- Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump in Nuclear Power Plan
- Issue 188: Steam Generator Tube Leaks/Ruptures Concurrent with Containment Bypass
- Issue 189: Susceptibility of Ice Condenser Containments to Early Failure from Hydogen Combustion During A Severe Accident
- Issue 190: Fatigue Evaluation of Metal Components for 60-Year Plant Life
- Issue 191: Assessment of Debris Accumulation on PWR Sump Performance
- Issue 192: Secondary Containment Drawdown Time
- Issue 193: BWR ECCS Suction Concerns
- Issue 194: Implications of Updated Probabilistic Seismic Hazard Estimates
- Issue 195: Hydrogen Combustion in Foreign BWR Piping
- Issue 196: Boral Degradation
- Issue 197: Iodine Spiking Phenomena
- Issue 198: Hydrogen Combustion in PWR Piping
- Issue 199: Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States
- Issue 200: Tin Whiskers
- Issue 201: Small-Break LOCA and Loss of Offsite Power Scenario
- Issue 202: Spent Fuel Pool Leakage Limits
- Issue 203: Potential Safety Issues with Cranes that Lift Spent Fuel Casks
- Issue 204: FLOODING OF NUCLEAR POWER PLANT SITES FOLLOWING UPSTREAM DAM FAILURES
Section 4. Human Factors Issues ( )
- Description
- Task HF1: STAFFING AND QUALIFICATIONS: Shift Staffing
- Task HF2: TRAINING: Evaluate Industry Training
- Task HF3: OPERATOR LICENSING EXAMINATIONS: Develop Job Knowledge Catalog
- Task HF4: PROCEDURES: Inspection Procedure for Upgraded Emergency Operating Procedures
- Task HF5: MAN-MACHINE INTERFACE: Local Control Stations
- Task HF6: MANAGEMENT AND ORGANIZATION: Develop Regulatory Position on Management and Organization
- Task HF7: HUMAN RELIABILITY: Human Error Data Acquisition
- Task HF8: Maintenance and Surveillance Program
Section 5. Chernobyl Issues ( )
- Description
- Task CH1: ADMINISTRATIVE CONTROLS AND OPERATIONAL PRACTICES: Administrative Controls to Ensure That Procedures Are Followed and That Procedures Are Adequate: Symptom-Based EOPs
- Task CH2: DESIGN: Reactivity Accidents: Reactivity Transients
- Task CH3: CONTAINMENT: Containment Performance During Severe Accidents: Containment Performance
- Task CH4: EMERGENCY PLANNING: Size of the Emergency Planning Zones
- Task CH5: SEVERE ACCIDENT PHENOMENA: Source Term: Mechanical Dispersal in Fission Product Release
- Task CH6: GRAPHITE-MODERATED REACTORS: Graphite-Moderated Reactors: The Fort St. Vrain Reactor and the Modular HTGR
Section 6. Nuclear Material Safety and Safeguards GSIs ( )
- Description
- NMSS-0001: Door Interlock Failure Resulting from Faulty Microselectron High Dose Rate Remote Afterloader
- NMSS-0002: Significant Quantities of Fixed Contamination Remain in Krypton-85 Leak-Detection Devices after Venting
- NMSS-0003: Corrosion of Sealed Sources Caused by Sensitization of Stainless Steel Source Capsules During Shipment
- NMSS-0004: Overexposures Caused by Sources Stolen from Facility of Bankrupt Licensee
- NMSS-0005: Potential for Erroneous Calibration, Dose Rate, or Radiation Exposure Measurements with Victoreen Electrometers
- NMSS-0006: Criticality in Low-Level Waste
- NMSS-0007: Criticality Benchmarks Greater than 5% Enrichment
- NMSS-0008: Year 2000 Computer Problem - Non-Reactor Licensees
- NMSS-0009: Amersham Radiography Source Cable Failures
- NMSS-0010: Troxler Gauge Source Rod Weld Failures
- NMSS-0011: Spent Fuel Dry Cask Weld Cracks
- NMSS-0012: Inadequate Transportation Packaging Puncture Tests
- NMSS-0013: Use of Different Dose Equivalent Models to Show Compliance
- NMSS-0014: Surety Estimates for Groundwater Restoration at in Situ Leach Facilities
- NMSS-0015: Adequacy of 10 CFR 150 Criticality Requirements
- NMSS-0016: Adequacy of 0.05 Weight Percent Limit in 10 CFR 40
- NMSS-0017: Misleading Marketing Information to General Licensees
- NMSS-0018: Problems Encountered When Manually Editing Treatment Planning Data on Nucletron Microselectron-HDR Model 105.999
- NMSS-0019: Control Unit Failures of Classic Nucletron HDR Units
- NMSS-0020: Leaking Pools
- NMSS-0021: Unlikely Events
- NMSS-0022: Gamma Stereotactic Radiosurgery
Section 7. Fukushima Issues ( )
Tables ( )
- Description
- TABLE II: LIST OF ALL THREE MILE ISLAND NUCLEAR PLANT ACTION PLAN ITEMS, TASK ACTION PLAN ITEMS, NEW GENERIC ISSUES, HUMAN FACT
- Table III: SUMMARY OF THE STATUS OF ALL GENERIC SAFETY ISSUES
- TABLE IV: LISTING OF AEOD REPORTS AND RELATED GENERIC ISSUES
- TABLE V: SUMMARY OF CONSOLIDATED GENERIC ISSUES
Appendices ( )
- Description
- Appendix A: Releases from Containment
- Appendix B: Applicability of NUREG-0933 Issues to Operating and Future Reactor Plants
- Appendix C: Priority Ranking Numerical Thresholds Used in Prioritizations Completed Before June 30, 1993
- Appendix D: Related Generic Activities
- Appendix E: Generic Communication and Compliance Activities
- Appendix F: Nuclear Material Safety and Safeguards GSIs
- Appendix G: Generic Issues Program Current and Historical Procedures
NUREG 0933: Revision History
NMSS-0012: Inadequate Transportation Packaging Puncture Tests
DESCRIPTION
This issue was identified when two holders of Certificates of Compliance for shipping packages performed puncture tests using a bar that was not properly mounted, as specified in 10 CFR 71.73(c)(3). As a result, NRC
Bulletin 97-02[1] was issued to holders of Certificates of Compliance for shipping packages under 10 CFR 71 with the request to review puncture test assessments for each certified package design. For designs based on physical tests, certificate holders were requested to determine whether the puncture tests were performed in accordance with 10 CFR 71.73(c)(3).
The responses to the Bulletin[2] identified packages that had not been puncture-tested in full accordance with the regulations. In addition, there were some packages for which the certificate holders were unable to determine whether the tests were performed correctly. In these cases, the certificate holders submitted justification for continued use of their packages.
CONCLUSION
This issue was given a medium priority ranking and resolution was pursued.[3] In response to Bulletin
97-02,[4] some certificate holders retested their packages to demonstrate adequacy. It was determined that, even though previous tests had not been performed exactly as specified in the regulations, the differences had no significant effect on the test results. Based on these findings and the excellent safety record for these
packages, the staff concluded that no further action was needed with respect to previously-approved packages. These packages were authorized for use under certain restrictions pursuant to 10 CFR 71.13. Thus, the issue
was resolved.[5]
[1] Bulletin 97-02, "Puncture Testing of Shipping Packages Under 10 CFR Part 71," U.S. Nuclear Regulatory Commission, September 23, 1997. [ML082460420]
[2] Bulletin 97-02, "Puncture Testing of Shipping Packages Under 10 CFR Part 71," U.S. Nuclear Regulatory Commission, September 23, 1997. [ML082460420]
[3] Memorandum for J. Craig from F. Combs, "Submittal of New Generic Issues for Tracking in the Generic Issues Management and Control System (GIMCS)," June 4, 1998. [9806090180]
[4] Bulletin 97-02, "Puncture Testing of Shipping Packages Under 10 CFR Part 71," U.S. Nuclear Regulatory Commission, September 23, 1997. [ML082460420]
[5] Memorandum for C. Rossi from D. Cool, "Status of NMSS Issues in the Generic Issue Management and Control System," June 25, 1999. [9907010194]
Page Last Reviewed/Updated 11/18/2025
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Page Last Reviewed/Updated 11/18/2025