Safety evaluation report . 58 , Related to the operation of Callaway plant, unit no.1

목차 (Table of Contents)

• CONTENTS
• Abstract = i
• 1. INTRODUCTION AND GENERAL DISCUSSION = 1-1
• 1.1 Introduction = 1-1
• 1.2 General Plant Description = 1-4
• 1.3 Comparison With Similar Facility Designs = 1-6
• 1.4 Identification of Agents and Contractors = 1-6
• 1.5 Summary of Principal Review Matters = 1-7
• 1.6 Modifications to the Facility During the Course of the Staff Review = 1-8
• 1.7 Summary of Outstanding Issues = 1-8
• 1.8 Confirmatory Issues = 1-9
• 1.9 License Conditions = 1-11
• 1.10 Unresolved Safety Issues = 1-12
• 2. SITE CHARACTERISTICS = 2-1
• 2.1 Geography and Demography = 2-1
• 2.1.1 Site Location Description = 2-1
• 2.1.2 Exclusion Area Authority and Control = 2-1
• 2.1.3 Population Distribution = 2-1
• 2.1.4 Conclusion = 2-5
• 2.2 Nearby Industrial, Transportation, and Military Facilities = 2-6
• 2.2.1 Transportation Routes = 2-6
• 2.2.2 Nearby Facilities = 2-6
• 2.2.3 Conclusion = 2-7
• 2.3 Meteorology = 2-7
• 2.3.1 Regional Climatology = 2-7
• 2.3.2 Local Meteorology = 2-9
• 2.3.3 Onsite Meteorogical Measurements Program = 2-9
• 2.3.4 Short-Term (Accident) Diffusion Estimates = 2-10
• 2.3.5 Long-Term (Routine) Diffusion Estimates = 2-11
• 2.4 Hydrology = 2-11
• 2.4.1 Hydrologic Description = 2-11
• 2.4.2 Flood Potential = 2-12
• 2.4.3 Water Supply = 2-14
• 2.4.4 Ultimate Heat Sink = 2-14
• 2.4.5 Site Groundwater = 2-16
• 2.4.6 Accident Effects = 2-17
• 2.4.7 Conclusion = 2-17
• 2.5 Geology and Seismology = 2-18
• 2.5.1 Basic Geologic and Seismic Information = 2-18
• 2.5.2 Sismology = 2-22
• 2.5.3 Capable Faults = 2-25
• 2.5.4 Stability of Subsurface Materials and Foundations = 2-26
• 2.5.5 Stability of Slopes = 2-31
• 2.5.6 Embankments and Dams = 2-32
• 2.6 References = 2-32
• 3. DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPNENTS = 3-1
• 3.1 General = 3-1
• 3.1.1 Conformance with General Design Criteria = 3-1
• 3.1.2 Conformance With Industry Codes and Standards = 3-1
• 3.2 Classification of Structures, Systems, and Components = 3-1
• 3.2.1 Seismic Classification = 3-1
• 3.2.2 System Quality Group Classification = 3-2
• 3.3 Wind and Tornado Loadings = 3-4
• 3.3.1 Wind Design Criteria = 3-4
• 3.3.2 Tornado Design Criteria = 3-4
• 3.4 Water Level (Flood) Design = 3-5
• 3.4.1 Water Level (Flood) Design = 3-5
• 3.4.2 Water Level (Flood) Design Procedures = 3-6
• 3.5 Missile Protection 3-6
• 3.5.1 Internally Generated Missiles = 3-6
• 3.5.2 Structures, Systems, and Components To Be Protected From Externally Generated Missiles = 3-9
• 3.5.3 Barrier Design Procedures = 3-10
• 3.6 Protection Against Effects Associated With the Postulated Rupture of Piping = 3-11
• 3.6.1 Determination of Break Locations and Dynamic Effects = 3-13
• 3.6.2 Postulated Pipe Breaks Outside Containment = 3-13
• 3.7 Seismic Design = 3-14
• 3.7.1 Seismic Input = 3-14
• 3.7.2 Seismic Structural System and Subsystme Analysis = 3-14
• 3.7.3 Seismic Mechanical Subsystem Analysis = 3-16
• 3.7.4 Seismic Instrumentation Program = 3-17
• 3.8 Design of Seismic Category I Structures = 3-17
• 3.8.1 Concrete Containment = 3-17
• 3.8.2 Concrete and Structural Steel Internal Structures = 3-18
• 3.8.3 Other Category I Structures = 3-19
• 3.8.4 Foundations = 3-20
• 3.9 Mechanical Systems and Components = 3-21
• 3.9.1 Special Topics for Mechanical Components = 3-21
• 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment = 3-15
• 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures 3-24
• 3.9.4 Control Rod Drive System 3-27
• 3.9.5 Reactor Pressure Vessel Internals = 3-27
• 3.9.6 Inservice Testing of Pumps and Valves = 3-28
• 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment = 3-29
• 3.11 Environmental Qualification of Safety-Related Electrical Equipment = 3-30
• 4. REACTOR = 4-1
• 4.1 General = 4-1
• 4.2 Fuel Design = 4-1
• 4.2.1 Design Basis 4-2
• 4.2.2 Description and Design Drawings = 4-10
• 4.2.3 Design Evaluation = 4-10
• 4.2.4 Testing, Inspection, and Surveillance Plans = 4-19
• 4.2.5 Evaluation Findings = 4-20
• 4.3 Nuclear Design = 4-21
• 4.3.1 Design Bases = 4-21
• 4.3.2 Design Description = 4-21
• 4.3.3 Analytical Methods = 4-25
• 4.3.4 Summary of Evaluation Findings = 4-26
• 4.4 Thermal-Hydraulic Design = 4-26
• 4.4.1 Performance and Safety Criteria = 4-26
• 4.4.2 Design Bases = 4-26
• 4.4.3 Thermal-Hydraulic Design Methodlogy = 4-27
• 4.4.4 Design Abnormalities = 4-29
• 4.4.5 ThermalHydraulic Omparison = 4-31
• 4.4.6 N-1 Loop Operation = 4-32
• 4.4.7 Conclusion = 4-33
• 4.5 Reactor Materials = 4-33
• 4.5.1 Control Rod Structural Materials = 4-33
• 4.5.2 Reactor Internals and Core Support Materials = 4-34
• 4.6 Functional Design of Reactivity Control Systems = 4-36
• 4.7 References = 4-38
• 5. REACTOR COOLANT SYSTEM = 5-1
• 5.1 Summary Description = 5-1
• 5.2. Integrity of Reactor Coolant Pressure Boundary = 5-1
• 5.2.1 Compliance With Codes and Code Cases 5-1
• 5.2.2 Overpressure Protection = 5-2
• 5.2.3 Reactor Coolant Pressure Boundary Materials = 5-4
• 5.2.4 Preservice and Inservice Inspection and Testing of the Reactor Coolant Pressure Boundary = 5-5
• 5.2.5 Reactor Coolant Pressure Boundary Leakage Detection System = 5-7
• 5.3 Reactor Vessel = 5-9
• 5.3.1 Reactor Vessel Materials and Compliance to Appendices G and H, 10 CFR Part 50 = 5-9
• 5.3.2 Pressure-Temperature Limits = 5-12
• 5.3.3 Reactor Vessel Integrity = 5-13
• 5.4 Component and Subsystem Design = 5-14
• 5.4.1 Reactor Coolant Pumps = 5-14
• 5.4.2 Steam Generators = 5-14
• 5.4.3 Residual Heat Removal System = 5-18
• 5.4.4 Pressurizer Relief Tank (Quench Tank) = 5-20
• 5.4.5 Loose Parts Monitoring System = 5-21
• 6. ENGINEERED-SAFETY-FEATURES = 6-1
• 6.1 Engineered Safety Features Materials = 6-1
• 6.1.1 Metallic Materials = 6-1
• 6.1.2 Organic Materials = 6-2
• 6.2 Containment Systems = 6-2
• 6.2.1 Containment Functional Design = 6-2
• 6.2.2 Containment Heat Removal System 6-10
• 6.2.3 Containment Isolation Systme = 6-12
• 6.2.4 Combustible Gas Control System = 6-18
• 6.2.5 Containment Leakage Testing Program = 6-20
• 6.3 Emergency Core Cooling System = 6-20
• 6.3.1 Design Basis = 6-20
• 6.3.2 System Design = 6-21
• 6.3.3 Design Evaluation for Single Failures = 6-21
• 6.3.4 Tests and Inspection = 6-23
• 6.3.5 Performance Evaluation = 6-24
• 6.3.6 Conclusions = 6-25
• 6.4 Control Room Habitability = 6-25
• 6.5 Engineered Safety Feature Atmosphere Cleanup System = 6-26
• 6.5.1 ESF FIliter System Description and Evaluation = 6-26
• 6.5.2 Containment Spray System = 6-27
• 6.6 Inservice Inspection of Class 2 and 3 Components = 6-29
• 6.6.1 Evaluation of Compliance for Callaway Unit No.1 with 10 CFR 50.55a(g) = 6-29
• 6.6.2 Conclusions = 6-30
• 7. INSTRUMENTATION AND CONTROLS = 7-1
• 7.1 General = 7-1
• 7.1.1 Acceptance Criteria = 7-1
• 7.1.2 Conformance to Criteria and Guidelines = 7-1
• 7.1.3 Specific Findings = 7-1
• 7.2 Reactor Trip System = 7-3
• 7.2.1 Description = 7-3
• 7.2.2 Resolution of Issues = 7-7
• 7.2.3 Evaluation Findings = 7-9
• 7.3 Engineered Safety Features Actuation System = 7-10
• 7.3.1 Description = 7-10
• 7.3.2 Resolution of Issues = 7-12
• 7.3.3 Evaluation Findings = 7-19
• 7.4 Systems Required for Safe Shutdown = 7-21
• 7.4.1 Description = 7-21
• 7.4.2 Remote Shutdown Capability = 7-23
• 7.4.3 Resolution of Issues = 7-24
• 7.4.4 Evaluation Findings = 7-25
• 7.5 Information Systems Important to Safety = 7-26
• 7.5.1 Description = 7-26
• 7.5.2 Resolution of Issues = 7-28
• 7.5.3 Evaluation Findings = 7-29
• 7.6 Interlock Systems Important to Safety 7-30
• 7.6.1 Residual Heat Removal Isolation Valves = 7-30
• 7.6.2 Accumulator Motor-Operated Valves = 7-31
• 7.6.3 Interlocks for RCS Pressure Control During Low Temperature Operation = 7-31
• 7.6.4 Switchover from Injection to Recirculation = 7-31
• 7.6.5 Isolation of Essential Service Water to the Air Compressors = 7-32
• 7.6.6 Isolation of the NonsafetyRelated Portion of the Component Coolling Water (CCW) System = 7-32
• 7.6.7 Resolution of Issues = 7-32
• 7.6.8 Evaluation Findings = 7-33
• 7.7 Control Systems = 7-34
• 7.7.1 Reactor Control = 7-34
• 7.7.2 Rob Control = 7-35
• 7.7.3 Pressurizer Pressure Control = 7-35
• 7.7.4 Pressurizer Water Level Control = 7-35
• 7.7.5 Steam Generator Water Level Control = 7-35
• 7.7.6 Steam Dump Control = 7-36
• 7.7.7 Plant Control System Interlocks = 7-37
• 7.7.8 Incore Instrumentation = 7-37
• 7.7.9 Boron Concentration Monitoring System = 7-37
• 7.7.10 Plant Control Signals for Monitoring and Indicating = 7-38
• 7.7.11 Resolution of Issues = 7-39
• 7.7.12 Evaluation Findings = 7-40
• 8. ELECTRIC POWER SYSTEMS = 8-1
• 8.1 General Considerations = 8-1
• 8.2 Offsite Power System = 8-1
• 8.2.1 Compliance with GDC 5 = 8-1
• 8.2.2 Compliance with GDC 17 = 8-1
• 8.2.3 Compliance with GDC 18 = 8-5
• 8.2.4 Evaluation Findings = 8-5
• 8.3 Onsite Emergency Power Systems = 8-5
• 8.3.1 Onsite-AC Power System's Compliance with GDC 17 = 8-5
• 8.3.2 Onsite-DC System Compliance with GDC 17 = 8-16
• 8.3.3 Common Electrical Features and Requirements = 8-19
• 8.3.4 Evaluation Findings = 8-25
• 9. AUXILIARY SYSTEMS = 9-1
• 9.1 Fuel Storage and Handling = 9-1
• 9.1.1 New-Fuel Storage = 9-1
• 9.1.2 Spent Fuel Storage = 9-2
• 9.1.3 Spent Fuel-Pool Cooling and Cleanup System = 9-6
• 9.1.4 Fuel Handling System 9-9
• 9.2 Water Systems = 9-10
• 9.2.1 Station Service Water System 9-4
• 9.2.2 Component Cooling Water System = 9-11
• 9.2.3 Demineralized Water Makeup System = 9-13
• 9.2.4 Potable and Sanitary Water Systems = 9-13
• 9.2.5 Ultimate Heat Sink = 9-14
• 9.2.6 Condensate Storage Facilities = 9-5
• 9.3 Process Auxiliaries = 9-15
• 9.3.1 Compressed Air System = 9-15
• 9.3.2 Process Sampling System = 9-16
• 9.3.3 Equipment and Floor Drainage System = 9-17
• 9.3.4 Chemical and Volume Control System = 9-18
• 9.4 Air Conditioning, Heating, Cooling, and Ventilation Systems = 9-19
• 9.4.1 Control Room Area Ventilation System (Control Building HVAC System) = 9-19
• 9.4.2 Fuel Pool Area Ventilation System (Fuel Building HVAC System) = 9-21
• 9.4.3 Auxiliary Building and Radwaste Area Ventilation System = 9-22
• 9.4.4 Turbine Area Ventilation System (Turbine Building HVAC) = 9-24
• 9.4.5 Engineered Safety Feature Ventilation (Diesel Generator Building Ventilation and Essential Service Water Pumphouse Ventilation) = 9-25
• 9.5 Other Auxiliary Systems = 9-26
• 9.5.1 Fire Protection Review = 9-26
• 9.5.2 Communication System = 9-35
• 9.5.3 Lighting System = 9-38
• 9.5.4 Emergency Diesel Engine Fuel Oil Storage and Transfer System = 9-40
• 9.5.5 Diesel Generator Cooling Water System = 9-43
• 9.5.6 Emergency Diesel Engine Starting System = 9-45
• 9.5.7 Emergency Diesel Engine Lubricating Oil System = 9-46
• 9.5.8 Emergency Diesel Engine Combustion Air Intake and Exhaust System = 9-49
• 10. STEAM AND POWER-CONVERSION SYSTEM = 10-1
• 10.1 Summary Description = 10-1
• 10.2 Turbine Generator = 10-1
• 10.2.1 Overspeed Protection System = 10-1
• 10.2.2 Turbine Disc Integrity = 10-2
• 10.3 Main Steam Supply System = 10-3
• 10.3.1 Main Steam Supply System (up to and Including the Main Steam Isolation Valves) = 10-4
• 10.3.2 Main Steam Supply System (Downstream of the Main Steam Isolation Valves) = 10-4
• 10.3.3 Steam and Feedwater System Materials = 10-5
• 10.4 Other Features = 10-6
• 10.4.1 Main Condenser = 10-6
• 10.4.2 Main Condenser Evacuation System = 10-6
• 10.4.3 Turbine Gland Sealing System = 10-7
• 10.4.4 Turbine Bypass System = 10-8
• 10.4.5 Circulating Water System = 10-8
• 10.4.6 Condensate Cleanup System = 10-9
• 10.4.7 Condensate and Feedwater Systems = 10-10
• 10.4.8 Steam Generator Blowdown System = 10-11
• 10.4.9 Auziliary Feedwater System = 10-11
• 11. RADIOACTIVE WASTE MANAGEMENT = 11-1
• 11.1 Summary Description = 11-1
• 11.2 Liquid-Waste-Management System Description and Evaluation = 11-1
• 11.3 Gaseous-Waste-Management Systems Description and Evaluation = 11-6
• 11.4 Solid-Waste-Management System Description and Evaluation = 11-12
• 11.5 Process and Effluent Radiological Monitoring and Sampling Systems Descriptions and Evaluation = 11-13
• 12. RADIATION PROTECTION = 12-1
• 12.1 Ensuring that Occupational Radiation Exposures Are As Low As Is Reasonably Achievable = 12-1
• 12.1.1 Policy Considerations = 12-1
• 12.1.2 Design Considerations = 12-2
• 12.1.3 Operational Considerations = 12-2
• 12.1.4 Decommisioning = 12-3
• 12.2 Radiation Sources = 12-3
• 12.3 Radiation Protection Design Features = 12-4
• 12.3.1 Facility Design Features = 12-4
• 12.3.2 Shielding = 12-4
• 12.3.3 Ventilation = 12-5
• 12.3.4 Are Radiation and Airborne Radioactivity Monitoring Instrumentation = 12-5
• 12.4 Dose Assessment = 12-6
• 12.5 Health Physics Program = 12-7
• 12.5.1 Program and Staff Oranization = 12-7
• 12.5.2 Health Physic Facilities = 12-8
• 12.5.3 Health Physics Instrumentation = 12-8
• 12.5.4 Procedures = 12-9
• 13. CONDUCT OF OPERATIONS = 13-1
• 13.1 Organizational Structure and Qualifications = 13-1
• 13.1.1 Management and Technical Resources = 13-1
• 13.1.2 Operating Organization = 13-7
• 13.2 Training = 13-15
• 13.2.1 Staff Training Program = 13-15
• 13.2.2 Licensed Operator Training Program = 13-16
• 13.3 Emergency Planning = 13-17
• 13.4 Review and Audit = 13-17
• 13.4.1 Onsite Review Committee = 13-18
• 13.4.2 Nuclear Safety Review Board = 13-18
• 13.4.3 Independent Safety Engineering Goup = 13-19
• 13.5 Plant Procedures = 13-19
• 13.5.1 Administrative Procedures = 13-19
• 13.5.2 Operating and Maintenance Procedures = 13-20
• 13.6 Industrial Security = 13-22
• 14. INITIAL TEST PROGRAM = 14-1
• 15. ACCIDENT ANALYSIS = 15-1
• 15.1 General = 15-1
• 15.1.1 Input Parameters for Transient and Accident Analyses = 15-1
• 15.1.2 Methods of Analysis = 15-2
• 15.2 Moderate Frequency Transients = 15-3
• 15.2.1 Increased Cooling Transients = 15-3
• 15.2.2 Decreased Cooling Transients = 15-3
• 15.2.3 Increased Core Reactivity Transients = 15-4
• 15.3 Infrequent Transients and Postulated Accidents = 15-5
• 15.3.1 Uncontrolled Rod Cluster Control Assembly (Rob) Bank Withdrawal at Power = 15-6
• 15.3.2 Inadvertent Loading of a Fuel Assembly Into Improper Position = 15-6
• 15.3.3 Rupture fo a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) = 15-7
• 15.3.4 Steam Line Breaks = 15-8
• 15.3.5 Feedwater Line Break = 15-8
• 15.3.6 Reactor Coolant Pump Locked Rotor Accident = 15-8
• 15.3.7 Loss-of-Coolant Accident (LOCA) = 15-9
• 15.3.8 Anticipated Transients Without Seram = 15-9
• 15.3.9 Conclusions = 15-11
• 15.4 Radiological Consequences of Design-Basis Accidents = 15-11
• 15.4.1 Main Steam Line Failure Outside Containment = 15-11
• 15.4.2 Rod Ejection Accident = 15-14
• 15.4.3 Failure of a Small Line Carrying Primary Coolant Outside Containment = 15-14
• 15.4.4 Steam Generator Tube Rupture Accident = 15-16
• 15.4.5 Loss-of-Coolant Accident = 15-17
• 15.4.6 Fuel Handling Accident = 15-18
• 16. TECHNICAL SPECIFICATIONS = 16-1
• 17. QUALITY ASSURANCE = 17-1
• 17.1 General = 17-1
• 17.2 Organization for the QA Program = 17-1
• 17.3 Quality Assurance Program 17-3
• 17.4 Conclusion 17-5
• 17.5 Outstanding Quality Assurance Issue for Callaway Unit 1 = 17-6
• 18. REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS = 18-1
• 19. COMMON DEFENSE AND SECURITY = 19-1
• 20. FINANCIAL QUALIFICATIONS = 20-1
• 20.1 Introduction = 20-1
• 20.2 Estimated Operating Costs of Facility = 20-1
• 20.3 Estimated Costs To Decommission Facility = 20-2
• 20.4 Reasonable Assurance of Funds = 20-2
• 20.5 Conclusion = 20-2
• 21. FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS = 21-1
• 21.1 General = 21-1
• 21.2 Preoperational Storage of Nuclear Fuel = 21-1
• 21.3 Operating Licenses = 21-1
• 22. TMI-2 REQUIREMENTS = 22-1
• 22.1 Introduction = 22-1
• 22.2 Discussion and Conclusions = 22-2
• Ⅰ.A.1.1 Shift Technical Advisor = 22-2
• Ⅰ.A.1.2 Shift Supervisor Administrative Duties = 22-2
• Ⅰ.A.1.3 Shift Manning = 22-3
• Ⅰ.A.2.1 Immediate Upgrading of Operator and Senior Operator Training and Qualifications = 22-3
• Ⅰ.A.3.1 Revised Scope and Criteria for Licensing Examinations = 22-4
• Ⅰ.B.1.2 Independent Safety Engineering Group = 22-4
• Ⅰ.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents = 22-4
• Ⅰ.C.2 Shift and Relief Turnover Procedures = 22-5
• Ⅰ.C.3 Shift Supervisor Respon = 22-5
• Ⅰ.C.4 Control Room Access = 22-5
• Ⅰ.C.5 Procedures for Feedback of Operating Experience to Plant Staff = 22-5
• Ⅰ.C.6 Verify Correct Performance of Operating Activities = 22-5
• Ⅰ.C.7 NSSS Vendor Review of Procedures = 22-6
• Ⅰ.C.8 Pilot Monitoring of Selected Emergency Procedures for Near-Term Operating License Applicants = 22-6
• Ⅰ.D.1 Control Room Design Review = 22-6
• Ⅰ.G.1 Special Low-Power Testing and Training = 22-8
• Ⅱ.B.1 Reactor Coolant System Vents = 22-8
• Ⅱ.B.2 Plant Shielding To Provide Access to Vital Areas and Pretect Safety Equipment for Postaccident Operation = 22-8
• Ⅱ.B.3 Postaccident Sampling Capability = 22-9
• Ⅱ.B.4 Training for Mitigating Core Damage = 22-10
• Ⅱ.D.1 Performance Testing of Boiling Water Reactor and Pressurized Water Reactor Relief and Safety Valves = 22-10
• Ⅱ.D.3 Direct Indication of Relief and Safety Valve Position = 22-11
• Ⅱ.E.1.1 Auxiliary Feedwater System (AFWS) Realibility Evaluation = 22-12
• Ⅱ.E.1.2 Auxiliary Feedwater Initiation and Indication = 22-19
• Ⅱ.E.3.1 Emergency Power Supply for Pressurizer Heaters = 22-20
• Ⅱ.E.4.1 Dedicated Containment Penetrations = 22-20
• Ⅱ.E.4.2 Containment Isolation Dependability = 22-20
• Ⅱ.F.1 Additional Accident Monitoring Instrumentation = 22-22
• Attachment 1: Noble Gas Effluent Monitor = 22-23
• Attachment 2: Sampling and Analysis of Plant Effluents = 22-23
• Attachment 3: Containment High-Range Radiation Monitor = 22-24
• Attachment 4: Containment Pressure Monitor = 22-24
• Attachment 5: Containment Water Level Monitor = 22-24
• Attachment 6: Containment Hydrogen Monitor = 22-25
• Ⅱ.F.2 Instrumentation for Detection of Inadequate Core Cooling = 22-25
• Ⅱ.G.1 Emergency Power for Pressurizer Equipment = 22-27
• Ⅱ.K.1 IE Bulletins on Measures To Mitigate Small-Break LOCAs and Loss-of Feedwater Accidents = 22-27
• Ⅱ.K.2.13 Thermal Mechanical Report-Effect of High-Pressure Injection on Vessel Integrity for Small-Break Loss-of-Coolant Accident With No Auxiliary Feedwater = 22-28
• Ⅱ.K.2.17 Potential for Voiding in the Reactor Coolant System During Transients = 22-28
• Ⅱ.K.2.19 Sequential Auxiliary Feedwater Flow Analysis = 22-29
• Ⅱ.K.3.1 Installation and Testing of Automatic Power-Operated Relief Valve Isolation System = 22-29
• Ⅱ.K.3.2 Report on Overall Safety Effect of Power-Operated Relief Valve Isolation System = 22-29
• Ⅱ.K.3.3 Reporting of Safety Valve and Relief Valve Failure and Challenges = 22-29
• Ⅱ.K.3.5 Automatic Trip of Reactor Coolant Pumps During LOCA = 22-29
• Ⅱ.K.3.9 Proportional Integral Derivative (PID) Controller Modification = 22-30
• Ⅱ.K.3.10 Proposed Anticipatory Trip Modification = 22-30
• Ⅱ.K.3.11 Justification of Use of Certain PORVs = 22-30
• Ⅱ.K.3.12 Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip = 22-30
• Ⅱ.K.3.17 Report on Outages of ECCS; Licensee Report and Proposed Technical Specification Changes = 22-30
• Ⅱ.K.3.25 Effect of Loss of AC Power on Pump Seals = 22-30
• Ⅱ.K.3.30 Revised Small-Break LOCA Methods To Show Compliance with 10 CFR 50.46 = 22-31
• Ⅲ.A.1.1 Upgrade Emergency Preparedness = 22-31
• Ⅲ.A.1.2 Upgrade Emergency Support Facilities = 22-31
• Ⅲ.A.2 Improving License Emergency Preparedness--Long Term = 22-31
• Ⅲ.D.1.1 Integrity of Systems Outside Containment Likely To Contain Radioactive Material = 22-32
• Ⅲ.D.3.3 Improved Inplant Iodine Instrumentation Under Accident Conditions = 22-32
• Ⅲ.D.3.4 Control Room Habitability Requirements = 22-33
• 23. CONCLUSIONS = 23-1
• APPENDICES
• Appendix A. Chronology of NRC Staff Radiological Review of Callaway
• Appendix B. Bibliography
• Appendix C. Nuclear Regulatory Commission Unresolved Safety Issues
• Appendix D. NRC Staff Contributors and Consultants
• Appendix E. Lawrence Livermore National Laboratory Report
• Appendix F. Emergency Preparedness Evaluation Report
• LIST OF FIGURES
• 2.1 Regional Census Map Showing Cities and Towns = 2-2
• 2.2 Site Vicinity Map = 2-3
• 2.3 Site Layout = 2-4
• 11.1 Liquid Waste Management System Schematic = 11-8
• 11.2 Gageous Waste Management System = 11-10
• 13.1 Corporate Organization of Union Electric Company for the Operation of Callaway = 13-2
• 13.2 Organization of the Nuclear Function = 13-3
• 13.3 Responsibilities of The General Manager, Engineering = 13-5
• 13.4 Responsibilities of the Manager, Operations = 13-6
• 13.5 Responsibilities of the Manager, Quality Assurance = 13-8
• 13.6 Organization of the Callway Plant Staff = 13-9
• 17.1 Callaway Quality Assurance Organization = 17-2
• F.1 Proposed Siren System = F-6
• F.2 Notification Points for In-Transit Population = F-8