NUREG-0933 Table of Contents

The on-line NRC version of NUREG-0933 consists of 11 pages with hyperlinks. The following is a compendium of those contents pages.


Section 1: TMI Action Plan Items

This section contains the TMI Action Plan items that were documented in NUREG-0660. All items in Chapters I, II, III, and IV that were identified for prioritization and listed in this section follow the numbering system established in NUREG-0660. Items found to be closely related have been combined where possible to form single issues for prioritization purposes. As a result, some of these combined issues contain items with the lead responsibility assigned to several offices. However, the lead responsibility and a summary of the findings for each item listed can be found in Table II of the Introduction. Items clarified in NUREG-0737 are listed in this section for accounting purposes only.


Chapters I, II, III, and IV presented a detailing of plans for NRC staff or licensee action whereas Chapter V addressed NRC policy, organization, and management and originally called for 17 specific actions to be taken by the Commissioners. In recognition of the interrelationships that required correlated planning, these 17 items were later grouped into seven subject areas by the staff and forwarded to the Commission in SECY-80-230B. This revision to Chapter V was agreed upon by the Commission and was published as Rev. 1 to NUREG-0660 in July 1980. All items of Chapter V listed in this section follow the numbering system established in NUREG-0660, Rev. 1.


Task I.A: Operating Personnel

Task I.A.2: Training and Qualifications of Operating Personnel

Task I.A.3: Licensing and Re-Qualification of Operating Personnel

Task I.A.4: Simulator Use and Development

Task I.B: Support Personnel

Task I.B.2: Inspection of Operating Reactors

Task I.C: Operating Procedures

Task I.D: Control Room Design

Task I.E: Analysis and Dissemination of Operating Experience

Task I.F: Quality Assurance

Task I.G: Pre-operational and Low-Power Testing


Task II.A: Siting

Task II.B: Consideration of Degraded or Melted Cores in Safety Review

Task II.C: Reliability Engineering and Risk Assessment

Task II.D: Reactor Coolant System Relief and Safety Valves

Task II.E: System Design

Task II.E.2: Emergency Core Cooling System

Task II.E.3: Decay Heat Removal

Task II.E.4: Containment Design

Task II.E.5: Design Sensitivity of B&W Reactors

Task II.E.6: In Situ Testing of Valves

Task II.F: Instrumentation and Controls

Task II.G: Electrical Power

Task II.H: TMI-2 Cleanup and Examination

Task II.J: General Implications of TMI For Design And Construction Activities

Task II.J.2: Construction Inspection Program

Task II.J.3: Management For Design and Construction

Task II.J.4: Revise Deficiency Reporting Requirements

Task II.K: Measures to Mitigate Small-break Loss-of-Coolant Accidents and Loss-of-Feedwater Accidents


Task III.A: Emergency Preparedness and Radiation Effects

Task III.A.2: Improving Licensee Emergency Preparedness - Long-term

Task III.A.3: Improving NRC Emergency Preparedness

Task III.B: Emergency Preparedness of State and Local Governments

Task III.C: Public Information

Task III.D: Radiation Protection

Task III.D.2: Public Radiation Protection Improvement

Task III.D.3: Worker Radiation Protection Improvement


Task IV.A: Strengthen Enforcement Process

Task IV.B: Issuance of Instructions and Information to Licensees

Task IV.C: Extend Lessons Learned to Licensed Activities Other than Power Reactors

Task IV.D: NRC Staff Training

Task IV.E: Safety Decision-Making

Task IV.F: Financial Disincentives to Safety

Task IV.G: Improve Safety Rulemaking Procedures

Task IV.H: NRC Participation in The Radiation Policy Council


Task V.A: Development of Safety Policy

Task V.B: Possible Elimination of Non-Safety Responsibilities

Task V.C: Advisory Committees

Task V.D: Licensing Process

Task V.E: Legislative Needs

Task V.F: Organization and Management

Task V.G: Consolidation of NRC Locations


Section 2. Task Action Plan Items

This section contains all Task Action Plan items documented in NUREG-0371 and NUREG-0471 as well as all USIs documented in other NRC publications. Items A-1 through A-41 are listed in NUREG-0371 and all items with prefixes "B," "C," and "D" are listed in NUREG-0471. USIs identified after publication of NUREG-0371 and NUREG-0471 are listed in the following documents: NUREG-0510 (A-42 through A-44); NUREG-0705 ( A-45 through A-48); and NUREG-1090 (A-49). A total of 142 items are listed in this section.


The Generic Issues Tracking System (GITS) Report issued on December 17, 1981, provided a status report on the majority of the 142 items as well as their classification into four categories: Environmental, Licensing Improvement, Safety, and USI. The safety issues identified in the GITS Report provided the basis for all prioritization work contained in this section. The lead responsibility and a summary of the findings for each item listed in this section can be found in Table II of the Introduction.


Item A-1: Water Hammer

Item A-2: Asymmetric Blowdown Loads on Reactor Primary Coolant Systems

Item A-3: Westinghouse Steam Generator Tube Integrity

Item A-4: CE Steam Generator Tube Integrity

Item A-5: B&W Steam Generator Tube Integrity

Item A-6: Mark I Short-Term Program

Item A-7: Mark I Long-Term Program

Item A-8: Mark II Containment Pool Dynamic Loads Long-Term Program

Item A-9: ATWS

Item A-10: BWR Feedwater Nozzle Cracking

Item A-11: Reactor Vessel Materials Toughness

Item A-12: Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports

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

Item A-18: Pipe Rupture Design Criteria

Item A-19: Digital Computer Protection System

Item A-20: Impacts of The Coal Fuel Cycle Description

Item A-21: Main Steam Line Break Inside Containment - Evaluation of Environmental Conditions for Equipment Qualification

Item A-22: PWR Main Steam Line Break - Core, Reactor Vessel, and Containment Building Response

Item A-23: Containment Leak Testing

Item A-24: Qualification of Class 1E Safety-Related Equipment

Item A-25: Non-Safety Loads on Class 1E Power Sources

Item A-26: Reactor Vessel Pressure Transient Protection

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

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

Item A-37: Turbine Missiles

Item A-38: Tornado Missiles

Item A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits

Item A-40: Seismic Design Criteria

Item A-41: Long-term Seismic Program

Item A-42: Pipe Cracks in Boiling Water Reactors

Item A-43: Containment Emergency Sump Performance

Item A-44: Station Blackout

Item A-45: Shutdown Decay Heat Removal Requirements

Item A-46: Seismic Qualification of Equipment in Operating Plants

Item A-47: Safety Implications of Control Systems

Item A-48: Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment

Item A-49: Pressurized Thermal Shock


Item B-1: Environmental Technical Specifications

Item B-2: Forecasting Electricity Demand

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

Issue 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-24: 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 Engineered Safety Features Systems and for Normal Ventilation Systems

Item B-37: Chemical Discharges to Receiving Waters

Item B-38: Reconnaissance Level Investigations

Item B-39: 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: Costs 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

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 Pressure

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

This section contains a compilation of issues that are not listed in any single report. these issues have surfaced since publication of NUREG-0371, NUREG-0471, and NUREG-0660.


The first 27 of these issues were transmitted from the Generic Issues Branch (GIB) to SPEB on September 30, 1981 for prioritization by SPEB. This transmittal provided the basis for the formation of this group of New Generic Issues.


Several of the first 27 issues had origins in reports published by ACRS, some of which were important enough to be considered as Candidate USIs. This group has been expanded to include new issues that have been raised by the NRR staff as well as those forwarded to NRR from other offices such as AEOD. It will continue to expand to include new safety issues as they arise. the lead responsibility and a summary of the findings for each issue listed in this section can be found in Table II of the Introduction.

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: The Loss of Offsite Power

Issue 48: LCO For Class 1E Vital Instrument Buses in Operating Reactors

Issue 49: Interlocks And LCOS For 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 System

Issue 52: SSW Flow Blockage by Blue Mussels

Issue 53: Consequences of a Postulated Flow Blockage Incident in a BWR

Issue 54: Survey of 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: Containment Flooding

Issue 59: Technical Specification Requirements for Plant Shutdown When Equipment for Safe Shutdown fs 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

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

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 Backflow 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

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 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 the 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 Anchor Head Failure

Issue 119: Piping Review Committee Recommendations

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

Issue 123: Deficiencies in the Regulations Governing DBA and Failure Criterion

Issue 124: Auxiliary Feedwater System Reliability

Issue 125: Davis-Besse Loss of All Feedwater Event of June 9, 1985 -- Long Term Actions

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 -- 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 Mark II Containments During Power

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: Generic Concerns Arising from TMI-2 Cleanup

Issue 156: Systematic Evaluation Program

Issue 157: Containment Performance

Issue 158: Performance of Safety-related Power-operated Valves under Design Basis Conditions

Issue 159: Qualification of Safety-related Pumps While Running on Minimum Flow

Issue 160: Spurious Actuations 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 Shutdown

Issue 163: Multiple Steam Generator Tube Leakage

Issue 164: Neutron Fluence in Reactor Vessel

Issue 165: Spring-actuated Safety and 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

Issue 174: Fastener Gaging Practices

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 LOCAs in PWRs

Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification

Issue 188: Steam Generator Tube Leaks or Ruptures, Concurrent with Containment Bypass from Main Steam Line or Feedwater Line Breaches

Issue 189: Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen 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


Section 4. Human Factors Issues

The issues presented in this section include those outlined in the Human Factors Program Plan (HFPP) and documented in NUREG-0985, Revision 1. This plan describes the human factors-related work required to complete the NUREG-0660 human factors tasks as well as the additional human factors-related efforts, identified during implementation of NUREG-0660 tasks, that require NRC attention. The lead responsibility and a summary of the findings for each item listed in this section can be found in Table II of the Introduction.


Human Factors Program Plan

Task HF1: Staffing and Qualifications

Task HF2: Training

Task HF3: Operator Licensing Examinations

Task HF4: Procedures

Task HF5: Man-machine Interface

Task HF6: Management and Organization

Task HF7: Human Reliability

Item HF8: Maintenance and Surveillance Program


Section 5. Chernobyl Issues

The staff's assessment of the implications of the Chernobyl accident on the safety regulation of U.S. commercial nuclear power plants, as reported in NUREG-1251, led to the conclusion that no immediate changes in NRC's regulations regarding the design or operation of U.S. commercial reactors were needed. However, further consideration of certain issues was recommended, most of which were found to be already under consideration as a part of ongoing NRC work.


This section includes all the work recommended in NUREG-1251 and outlined in the staff's follow-up program, SECY-89-081. As noted in NUREG-1251, the Chernobyl experience will continue to be taken into account in various areas of reactor safety. The follow-up program was limited to work on those issues whose relationship to the events at Chernobyl was direct, clear, and substantial, but with reasonable extrapolation to account for the large differences in specific design and operational features. Other work that was generally related to severe accidents was to be pursued (or considered for pursuit), in accordance with established procedures, outside the Chernobyl follow-up program.


The tasks contained in this section follow the numbering sequence of the various chapters in NUREG-1251. The issues identified for further pursuit under each task follow the labeling of the follow-up program.


Task CH1: Administrative Controls and Operational Practices

Task CH2: Design

Task CH3: Containment

Task CH4: Emergency Planning

Task CH5: Severe Accident Phenomena

Task CH6: Graphite-moderated Reactors



Appendix A sections are excerpted from Appendix VI, Section 2 of WASH-1400 (NUREG-75/104), "Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," dated October 1975.


Section 2 Releases from Containment



PWR Types 1 through 9

BWR Types 1 through 5



This appendix contains a listing of those residual GSIs that are applicable to operating and future reactor plants and includes: issues that have been resolved with

Requirements, USI, HIGH- and MEDIUM-priority issues scheduled for resolution; nearly-resolved issues scheduled for resolution (NOTES 1 and 2); and

issues that are scheduled for prioritization (NOTE 4).





(a) The priority rank is always HIGH when any of the following risk (or risk related) thresholds are estimated to be exceeded (or when extraordinary uncertainty suggests that they may well be exceeded):

(1) 11000 person-rem estimated public dose per remaining reactor lifetime

(2) 50,000 person-rem total estimated for all affected reactors for their remaining lifetime (e.g., 500 person-rem/reactor for 100 reactors)

(3) 10-5/reactor-year large-scale core-melt

(4) 5 x 10-4/year large-scale core-melt (total for all affected reactors)

(b) Always at least MEDIUM priority:

10 or more percent of the always-HIGH criteria

(c) Always at least LOW priority:

1 or more percent of the always-HIGH Criteria

(d) Never higher than MEDIUM priority:

Less than 10% of the always-HIGH criteria

(e) Never higher than LOW priority:

Less than 1% of the always-HIGH criteria

(f) Always DROP category:

Less than 1% of the always-HIGH criteria



This appendix documents those activities related to generic issues, i.e., related generic activities (RGA), that did not meet the criteria for designation as generic issues (GI), but were important enough to require the development of Action Plans by NRR to address the concerns. The plan for documenting these RGAs was delineated in SECY-96-107.











Appendix E: Generic Communication and Compliance Activities

This appendix documents those generic communication and compliance activities (GCCA) completed by NRR that did not meet the criteria for designation as generic issues (GI), but were important enough to require the issuance of Information Notices (IN) and/or Generic Letters (GL) to licensees. The plan for documenting closed GCCAs was delineated in SECY-96-107.


GCCA-0001: Assessment Of Condition Of Safety-Related Structures And Civil Engineering Features

GCCA-0002: Environmental Licensing And Regulatory Concentrations In Building Wakes

GCCA-0003: Rrg, 50.54(P) Guidance

GCCA-0004: Relocation Of Selected Ts Requirements Related To Instrumentation (Gl)

GCCA-0005: Bwr - Scram Solenoid Pilot Valve Problems

GCCA-0006: Shift Staffing Issue Followup

GCCA-0007: Lessons Learned From Operational Safeguards Response Evaluations

GCCA-0008: Air Entrainment In Terry Turbine Lubricating Control Oil System

GCCA-0009: Wrong Replacement Parts Relief Valves And Refueling Mast

GCCA-0010: Spent Fuel Pool Overflow Into Ventilation System

GCCA-0011: Ipeee For Severe Accident Vulnerabilities

GCCA-0012: Surry Ventilation Filter Issue

GCCA-0013: Common Mode Failure Of Copes Volcan Porvs

GCCA-0014: Deficiencies Identified During Electrical Distribution System

GCCA-0015: Seismic Adequacy Of Thermo-Lag Panels

GCCA-0016: Capability Of Offsite Power During Design Basis Events

GCCA-0017: Potential For Loss Of Automatic Esf Actuation

GCCA-0018: Potential For Mov Failure - Stem Protection Pipe Changes

GCCA-0019: Unanticipated And Unauthorized Movement Of Fuel

GCCA-0020: Fraudulent Commercial Grade Certificate Of Compliance

GCCA-0021: Chatter Of Itt Barton 288a And 289a Differential Pressure

GCCA-0022: Frequency Of Use Of Air-Operated Gate Valves

GCCA-0023: Pressure Locking And Thermal Binding Of Gate Valves

GCCA-0024: Degraded Decay Heat Removal Capability Via Natural Circulation

GCCA-0025: Falsification Of Asnt Certificate By American Power Services

GCCA-0026: Adequacy Of Emergency And Essential Lighting

GCCA-0027: Address Concerns Regarding Asme Code

GCCA-0028: Circumferential Cracking Of Steam Generator Tubes

GCCA-0029: Reactor Coolant Pump Turning Vane Bolt Locking Device Failure

GCCA-0031: Potential Cable Damage From Excess Side Wall Pressure

GCCA-0032: Evaluate Missiles From Mirror Insulation During High Energy Pipe Breaks

GCCA-0033: Results Of Recent Nrc Sponsored Flame Spread And Fire Endurance Testing

GCCA-0036: Failure Of Automatic Ventilation System Operation Following A Loss Of Offsite Power

GCCA-0037: Failure To Test Swing Buses During Integrated Emergency Diesel Generator Surveillance

GCCA-0038: Lightning Dissipation Systems

GCCA-0039: Switchgear Fire And Partial Loss Of Offsite Power

GCCA-0040: Spent Fuel Transfer Canal Shielding Deficiency At Boiling Water Reactor

GCCA-0041: Changes In The Operator Licensing Program

GCCA-0042: Unplanned, Unmonitored Release Of Radioactivity From The Exhaust Ventilation System Of A Bwr

GCCA-0043: Susceptibility Of Low Pressure Coolant And Core Spray Injection Valves To Pressure Locking

GCCA-0044: Potentially Nonconforming Fasteners Supplied By A&G Engineering Ii, Inc.

GCCA-0045: Current Fire Endurance Test Results For 3m Interam Raceway Fire Barrier Systems

GCCA-0046: Potential For Data Collection Equipment To Affect Protection System Performance

GCCA-0047: Boraflex Degradation In Spent Fuel Pool Storage Racks

GCCA-0049: Legal Actions Against Thermal Science, Inc., Manufacturer Of Thermo-Lag

GCCA-0050: Transient Involving Open Safety Relief Valve Followed By Complications

GCCA-0052: Unexpected Opening Of An Srv And Complications Involving Suppression Pool Strainer Blockage

GCCA-0053: Decay Heat Management Practices During Refueling

GCCA-0054: Potential For Loss Of Automatic Engineered Safety Features Actuation

GCCA-0056: Augmented Reactor Vessel Inspection

GCCA-0057: Consideration Of Position Changeable Valves

GCCA-0058: Problem Of Grease Leakage In Pre-Stressed Concrete Containment

GCCA-0060: Inadequate Testing Of Safety-Related Logic Circuits

GCCA-0061: Boraflex Degradation In Spent Fuel Pool Storage Racks

GCCA-0062: Ansys And Gtstrudl Computer Program Error Notifications

GCCA-0063: Inadequate Control Of Molded-Case Circuit Breakers

GCCA-0064: Relocation Of Rcs Pressure/Temperature Limits

GCCA-0065: Reconsideration Of Plant Security Requirements

GCCA-0066: Fires In Emergency Diesel Generator Exciters

GCCA-0067: Inadequate Capacity Of Ccw Leads To Freon Release To The Control Room

GCCA-0068: Bwr Stability With Flow Slightly Less Than Natural Circulation Flow

GCCA-0070: Evaluate Impact Of Rcp Support Column Tilt On Leak Before Break Analyses

GCCA-0071: Fish Mouth Burst And Bowing Of Previously-Plugged Steam Generator Tubes

GCCA-0072: Blockage Of Untested Eccs Piping

GCCA-0073: Porv Inoperability Masked By Downstream Indications During Testing

GCCA-0074: Loss Of Rc Inventory And Potential Loss Of Emergency Mitigation Functions While In A Shutdown Condition

GCCA-0075: Control Rod Drive Mechanism Penetration Cracking

GCCA-0076: Augmented Examination Of Reactor Vessel

GCCA-0077: Closed Head Vent Causes Inaccurate Level Indication During Reduced Inventory

GCCA-0078: Shutdown Cooling Flow Bypassing Core Results In Temperature And Pressure Increases

GCCA-0079: Potential Containment Leak Path Through Hydrogen Analyzer

GCCA-0080: Inadequate Testing And Design Of Tornado Dampers

GCCA-0081: Assessment Of Corrosion Of B&W Fuel Used In 2-Year Fuel Cycles

GCCA-0082: Environmental Effects On Main Steam Safety Valve Set Point

GCCA-0083: Inadvertent Draining Of Reactor Vessel And Isolation Of Shutdown Cooling System

GCCA-0084: Recent Problems With Overhead Cranes

GCCA-0085: Removing Refueling Floor Shielding Plugs Prior To And Soon After Shutdown

GCCA-0086: Damage To Valve Internals Caused By Thermally-Induced Pressure Locking

GCCA-0087: Damage In Foreign Steam Generator Internals

GCCA-0088: Interface Between Operators And Nuclear Engineers During Tests And Startup

GCCA-0089: Valve Stem Coupling Of Gimpel Auxiliary Feedwater Turbine Trip Throttle Valves

GCCA-0091: Use Of Individual Plant Examinations (Ipes) For Regulatory Decision Making

GCCA-0092: Overwithdrawal Of Tip

GCCA-0093: Spent Fuel Pool Cooling

GCCA-0094: South Texas Stuck Rod Event Following Reactor Trip

GCCA-0095: Radwaste Facility Equipment Degradation At Millstone Unit 1

GCCA-0096: Wolf Creek Reactor Trip With One Train Essential Service Water System Inoperable

GCCA-0097: Stuck Control Rod Problems

GCCA-0099: Slow Five Percent Scram Insertion Times Caused By Viton Diaphragms In Scram Solenoid Pilot Valves

GCCA-0100: Potential Clogging Of Hpsi Throttle Valves During Containment Sump Recirculation Phase

GCCA-0101: Steam Generator Tube Inspection Results

GCCA-0102: Reactor Operation Believed To Be Inconsistent With That Described In The Fsar

GCCA-0103: Movement Of Dry Storage Casks Over Spent Fuel, Fuel In The Reactor Core, Or Safety-Related Equipment

GCCA-0104: Inaccuracy Of Diagnostic Equipment For Motor-Operated Butterfly Valves

GCCA-0105: Cross-Tied Safety Injection Accumulators

GCCA-0106: Hydrogen Gas Ignition During Welding Of A Vsc-24 Multi-Assembly Sealed Basket



This appendix documents those non-reactor GSIs identified, prioritized, and resolved by NMSS. As stated in SECY-98-001, the prioritization procedure for these issues is contained in NMSS Policy and Procedures Letter 1-57, "NMSS Generic Issues Program."


Issue No. Title Priority

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 Fields

NMSS-0015 Adequacy of 10 CFR 150 Criticality Requirements

NMSS-0016 Adequacy of 0.05 Weight Percent Limit in 10 CFR 40



Abbreviations used throughout NUREG-0933.



A listing of the 1853 references used in preparation of NUREG-0933.


Copyright 1996-2006.  The Virtual Nuclear Tourist. All rights reserved. Revised: December 19, 2005.