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
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
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:
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)
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:
Issue 73: Detached Thermal Sleeves
Issue 74: Reactor Coolant Activity Limits for Operating Reactors
Issue 75: Generic Implications of ATWS Events at the
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
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:
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
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
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
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
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.
The staff's assessment of the implications of the
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
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:
RELEASES FROM CONTAINMENT
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
2.1 GENERAL REMARKS
2.2 ACCIDENT DESCRIPTIONS
PWR Types 1 through 9
BWR Types 1 through 5
APPENDIX B:
APPLICABILITY OF NUREG-0933 ISSUES TO OPERATING AND FUTURE REACTOR PLANTS
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).
APPENDIX C:
PRIORITY RANKING NUMERICAL THRESHOLDS USED IN PRIORITIZATIONS COMPLETED BEFORE
JUNE 30, 1993
TABLE 1 RISK THRESHOLDS
(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
APPENDIX D:
RELATED GENERIC ACTIVITIES
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.
GA-001: BOILING WATER REACTOR INTERNALS
GA-002: REACTOR PRESSURE VESSEL FRACTURE TOUGHNESS
GA-003: DRY CASK STORAGE OF SPENT FUEL
GA-004: THERMO-LAG FIRE BARRIERS
GA-005: RCS DRAINDOWN
GA-006: SRP REVISION
GA-007: PRA IMPLEMENTATION PLAN
RGA-007.1.2(D): GRADED QUALITY ASSURANCE
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
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:
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:
GCCA-0095: Radwaste Facility Equipment Degradation At Millstone Unit 1
GCCA-0096:
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
APPENDIX F:
NUCLEAR MATERIAL SAFETY AND SAFEGUARDS GSIs (REV. 4)
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
Glossary
Abbreviations used throughout NUREG-0933.
References
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.