ML20049H280
| ML20049H280 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/12/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8202230051 | |
| Download: ML20049H280 (3) | |
Text
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O O
APPENDIX 6A IMPROVED DELAY HEAT CORRELATION FOR LOCA ANALYSIS (LATER) 1 i
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O 8202230051 820212 PDR ADOCK 05000447 K pyg
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 4
APPENDIX 6A CONTENTS Section Title Page 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS 6A-1 6A.1
SUMMARY
, 6A.l.2 Objective 6A-2 6A.2 APPLICATION 6A-3 6A.2.1 Exemption Request 6A-3 6A.2.2 Applicability 6A-3 6A.2.2.1 Operating Plant Analysis 6A-3 l 6A.2.2.1.1 Analysis for New Fuel Only 6A-3 6A.2.2.1.2 Analysis for All Fuel 6A-3 6A.2.2.1.3 Limiting Break Analysis 6A-3 i 6A.2.2.1.4 Complete Recalculation 6A-3 6A.2.2.2 Requisition Plant Analysis 6A-4 6A.2.2.2.1 Plants with Con 1pleted Analysis 6A-4
( Plants without Completed Analysis 6A-4 6A.2.2.2.2 6A.3 INTRODUCTION 6A-4 6A.3.1 Licensing Philosophy for Evaluation Models 6A-5 6A.3.2 Other Appendix K Conservatisms 6A-6 6A.4 OBJECTIVE 6A-7 6A.4.1 Prevent MAPLHGR Derate 6A-7 6A.4.2 Efficient Fuel Utilization 6A-7 6A.4.3 Cost Effective Allocation of Resources 6A-8 6A.5 TECHNICAL BASIS 6A-9 6A.5.1 Basis for U.S. BNR Core Power Limits 6A-9 6A.5.1.1 Peak Pellet Linear Heat Generation Rate 6A-10 6A.5.1.2 Core Average Power Density 6A-ll 6A.5.1.3 Core Average Specific Power 6A-13 6A.S.2 Technical Advantages of the 1979 Decay Heat Standard 6A-13 O 6A.5.2.1 Background 6A-13 6A-i
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O CONTENTS (Continued)
Section Title Page 6A.5.2.2 Decay Heat Power from Fission Products 6A-14 6A.5.2.3 Effect of Neutron Capture 6A-15 6A.5.2.4 Pulsed and Infinite Reactor Operation 6A-15 6A.6 METHOD OF ANALYSIS 6A-17 6A.6.1 Summary of Decay Heat Terms 6A-17 6A.6.2 Total Energy per Fission 6A-18 6A.6.3 Fission Heat Induced by Delayed Neutrons 6A-20 6A.6.4 Fission Product Decay Heat 6A-21 6A.6.5 Decay Heat Due to Actinide Decay 6A-23 6A.6.6 Decay of Activated Structure and Poison Materials 6A-25 6A.6.7 Application Considerations 6A-26 6A.6.8 Evaluation of Uncertainties 6A-30 6A.7 CONCLUSION 6A-47 6A.8 REFERENCES 6A-48 O
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I GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
SUMMARY
TABLE OF CONTENTS I
- Chapter / !
! Section Title Volume !
1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT ,
l
1.1 INTRODUCTION
1 f I 1.1.1 Type of License Required- !
I 1.1.2 Identification of Applicant i 1.1.3 Number of Plant Units 1.1.4 Description of Location i
1.1.5 Type of Nuclear Steam Supply System l 1.1.6 Type of Containment 1
l
! 1.1.7 Core Thermal Power Levels 1.1.8 Scheduled Completion and Operation Dates i
1.2 GENERAL PLANT DESCRIPTION 1 1.2.1 Principal Design Criteria 1.2.2 Plant Description
! 1.3 COMPARISON TABLES 1 1.3.1 Comparisons with similar Facility f
, Designs
! 1.3.2 Comparisons of Final and Preliminary Information l
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS' 1 1.4.1 Applicant ,
1.4.2 Architect Engineer - Nuclear Island Design ,
1.4.3 Nuclear Steam Supply System Supplier 1.4.4 Turbine-Generator Vendor 1.4.5 Consulta its 1.5 REQUIREMENTS FOR F'JRTIIER TECIINICAL INFORMATION 1 1.5.1 Current Development Programs 1.5.2 PSAR Commitment Items I 1.6 MATERIAL INCORPORATED BY REFERENCE 1 l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7.2 Piping and Instrumentation Diagrams 1.7.3 Abbreviations and Symbols 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1 1.8.1 Compliance Assessment Method 1.9 STANDARD DESIGNS 1 1.9.1 Interfaces 1.9.2 Exceptions O
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SUMMARY
TABLE OF CONTENTS (Continued)
V)
(
Chapter /
Section Title Volume 2 SITE CHARACTERISTICS 2.0
SUMMARY
l 2.1 GEOGRAPHY AND DEMOGRAPHY l 2.1.1 Site Location and Description 2.1.2 Exclusion Area Authority and Control 2.1.3 Population Distribution 2.2 NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 1 2.2.1 Location and Routes 2.2.2 Descriptions 2.2.3 Evaluation of Potential Accidents 2.3 METEOROLOGY l 2.3.1 Regional Climatology 2.3.2 Local Meteorology
[
\'J 2.3.3 Onsite Meteorological Measurements Prograta 2.3.4 Short-Term Atmospheric Diffusion Estimates 2.3.5 Long-Term Atmospheric Diffusion Estimates 2.4 HYDROLOGIC ENGINEERING 1 2.4.1 Hydrologic Description 2.4.2 Floods 2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers 2.4,4 Potential Dam Failures, Seismically Induced 2.4.5 Probable Maximum Surge and Seiche Flooding 2.4.6 Probable Maximum Tsunami Flooding 2.4.7 Ice Effects 2.4.8 Cooling Water Canals and Reservoirs 2.4.9 Channel Diversions
['
2.4.10 Flooding Protection Requirements v
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
O Section Title Volume 2.4.11 Low Water Considerations 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters 2.4.13 Ground Water 2.4.14 Technical Specifications and Emergency, Operation Requirements 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING 1 2.5.1 Basic Geologic and Seismic Information 2.5.2 Vibratory Ground Motion 2.5.3 Surface Faulting 2.5.4 Stability of Subsurface Materials and Foundations 2.5.5 Stability of Slopes 2.5.6 Embankments and Dams l
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
SUMMARY
TABLE OF CONTENTS (Continued) 7-N,,Y Chapter /
Section Title Volume 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITII NRC GENERAL DESIGN CRITERIA 2 3.1.1 Summary Description 3.1.2 Criterion Conformance 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 2 3.2.1 Seismic Classification 3.2.2 System Quality Group Classifications 3.2.3 System Safety Classifications 3.2.4 Quality Assurance 3.2.5 Correlation of Safety Classes with Industry Codes 3.3 WIND AND TORNADO LOADINGS 2 3.3.1 Wind Loadings ss- 3.3.2 Tornado Loadings 3.3.3 BOP Interface 3.3.4 References 3.4 WATER LEVEL (FLOOD) DESIGN 2 3.4.1 Flood Protection 3.4.2 Analytical and Test Procedures 3.4.3 BOP Interface 3.4.4 References 3.5 MISSILE PROTECTION 2 3.5.1 Missile Selection and Description 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles 3.5.3 Barrier Design Procedures 3.5.4 BOP Interface 3.5.5 References vii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 2 3.6.1 Postulated Piping Failures in Fluid Systems Inside and outside of Containment 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.3 References 3.7 SEISMIC DESIGN 3 3.7.1 Seismic Input 3.7.2 Seismic System Analysis 3.7.3 Seismic Subsystem Analysis 3.7.4 Seismic Instrumentation 3.7.5 BOP Interface 3.7.6 References 3.8 DESIGN OF SEISMIC CATEGORY I STRUCTURES 3 3.8.1 Concrete Containment 3.8.2 Steel containment 3.8.3 Concrete and Steel Internal Structures of Steel Containment 3.8.4 Other Seismic Category I Structures 3.8.5 Foundations 3.8.6 BOP Interface 3.9 MECHANICAL SYSTEMS AND COMPONENTS 4 3.9.1 Special Topics for Mechanical Components 3.9.2 Dynataic Testing and Analysis 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures 3.9.4 Control Rod Drive System 3.9.5 Reactor Pressure Vessels Internals 3.9.6 Inservice Testing of Pumps and Valves 3.9.7 BOP Interface 3.9.8 References v ii
' GESSAR II 22A7007
, 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 3.10 SEISMIC QUALIFICATIONS OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT (INCLUDING HYDRODYNAMIC EFFORTS) 5 3.10.1 Seismic Qualification Criteria (Including Hydrodynamic Loads) 1 3.10.2 Methods and Procedures for Qualifying Electrical Equipment and Instrumentation 3.10.3 Methods and Procedure of Seismic Analysis or Testing of Supports of Electrical Equipment and Instrumentation (Including ilydrodynamic Loads) 3.10.4 Seismic Qualification Tests and Analyses (Including Hydrodynamic Loads) 3.11 ENVIRONMENTAL DESIGN OF SAFETY-RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 5 3.11.1 Equipment Identification and Environmental Conditions 3.11.2 Qualification Tests and Analyses 3.11.3 Qualification Results 3.11.4 Loss of Ventilation 3.11.5 Estimated Chemical and Radiation Environment APPENDIX 3A SEISMIC SOIL-STRUCTURE INTERACTION ANALYSIS OF THE NUCLEAR ISLAND 5 APPENDIX 3B CONTAINMENT LOADS 6,7 APPENDIX 3C COMPUTER PROGRAMS USED IN THE DESIGN OF SEISMIC CATEGORY I STRUCTURES 8 APPENDIX 3D ANALYSIS OF RECIRCULATION MOTOR AND PUMP UNDER ACCIDENT CONDITIONS 8 APPENDIX 3E DESCRIPTION OF SAFETY RELATED MECHANICAL AND ELECTRICAL EQUIPMENT 8 APPENDIX 3F DYNAMIC BUCKLING CRITERIA FOR CONTAINMENT VESSEL 8
)
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume APPENDIX 3G PIPE FAILURE ANALYSIS 8 APPENDIX 3H EFFECT OF CONCRETE ANNULUS BELOW ELEVATION (-) 5 FT., 3 IN. ON SEISMIC DESIGN LOADS AND BUILDING RESPONSES 8 9
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SUMMARY
TABLE OF CONTENTS (Continued) 2 i Chapter /
! Section- Title Volume o
i 4 REACTOR 4.1
SUMMARY
DESCRIPTION 9 4.1.1- Reactor Vessel j 4.1.2 Reactor Internal Components 3
4.1.3 Reactivity Control Systems 4.1.4 Analysis Techniques 4 4.1.5 References 4.2 FUEL SYSTEM DESIGN 9 4.2.1 General and Detailed Design Base 4.2.2 General Design Description 4.2.3 Design Evaluations 4.2.4 Testing and Inspection 4.2.5 Operating and Developmental
' Experience 4.2.6 References 4.3 NUCLEAR DESIGN 9 4.3.1 Design Bases 4.3.2 Description 4.3.3 Analytical Methods 4.3.4 Changes
< 4.3.5 References i
4.4 THERMAL - HYDRAULIL DESIGN 9
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4.4.1 Design Basis 4.4.2 Description of Thermal Hydraulic Design of the Reactor Core 4.4.3 Description of the Thermal and Hydraulic Design of the Reactor
- Coolant System 4.4.4 Evaluation 4.4.5 Testing and Verification 4.4.6 Instrumentation Requirements
{' 4.4.7 References f
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
Chapter /
Section Title Volume 4.5 REACTOR MATERIALS 9 4.5.1 Control Rod System Structural Materials 4.5.2 Reactor Internal Materials 4.5.3 Control Rod Drive llousing Supports 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS 9 4.6.1 Information for Control Rod Drive System (CRDs) 4.6.2 Evaluations of the CRDs 4.6.3 Testing and Verification of the CRDs 4.6.4 Information for Combined Performance of Reactivity Systems 4.6.5 Evaluation of Combined Performance 4.6.6 References APPENDIX 4A CONTROL ROD PATTERNS AND ASSOCIATED POWER DISTRIBUTION FOR TYPICAL BWR 9 4A.1 Introduction 4A.2 Power Distribution Strategy 4A.3 Results of Core Simulation Studies 4A.4 Glossary of Terms 4A.5 References O
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22A7007 GESSAR II 238 NUCLEAR ISLAND Rev. O I
SUMMARY
TABLE OF CONTENTS (Continued)
Chapter / ,
- Section Title Volume 1
l 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1
SUMMARY
DESCRIPTION 10 J 5.1.1 Schematic Flow Diagram 5.1.2 Piping and Instrumentation Diagram l 5.1.3 Elevation Drawing l 5.2 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY 10 j 5. 2.1 - Compliance with Codes and Code cases j 5.2.2 Overpressure Protection 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2.5 Reactor Coolant Pressure Boundary and ECCS System Leakage Detection
, f-~g System I
\m / 5.2.6 References 1
5.3 REACTOR. VESSEL 10 5.3.1 Reactor Vessel Materials 5.3.2 Pressure / Temperature Limits l
5.3.3 Reactor Vessel Integrity 5.3.4 References 5.4 COMPONENT AND SUBSYSTEM DESIGN 10
! 5.4.1 Reactor Recirculation Pumps 5.4.2 Steam Generators (PWR) l 5.4.3 Reactor Coolant Piping 5.4.4 Main Steam Line Flow Restrictors 5.4.5 Main Steam Line Isolation System
- 5.4.6 Reactor Core Isolation Cooling System (RCIC) 5.4.7 Residual Heat Removal System 5.4.8 Reactor Water Cleanup System 5.4.9 Main Steam Lines and Feedwater Piping l () 5.4.10 Pressurizer xiii 4
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SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 5.4.11 Pressurizer Relief Valve Discharge System 5.4.12 Valves 5.4.13 Safety and Relief Valves 5.4.14 Component Supports 5.4.15 References I
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SUMMARY
TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 6 ENGINEERED SAFETY FEATURES 6.0 GENERAL 11 6.1 ENGINEERED SAFETY FEATURE MATERIALS 11 6.1.1 Metallic Materials 6.1.2 Organic Materials 6.2 CONTAINMENT SYSTEMS 11 6.2.1 Containment Functional Design 6.2.2 Containment Heat Removal System 6.2.3 Secondary Containment Functional Design 6.2.4 Containment Isolation System 6.2.5 Combustible Gas Control in Containment 6.2.6 Containment Leakage Testing
() 6.2.7 6.2.8 Suppression Pool Makeup System References 6.3 EMERGENCY CORE COOLING SYSTEMS 11 6.3.1 Design Bases and Summary Description 6.3.2 System Design 6.3.3 ECCS Performance Evaluation 6.3.4 Tests and Inspections 6.3.5 Instr" mentation Requirements 6.3.6 References 6.4 HABITABILITY SYSTEMS 11 6.4.1 Design Basis l
I 6.4.2 System Design 6.4.3 Systems Operational Procedures 6.4.4 Design Evaluations l 6.4.5 Testing and Inspection 6.4.6 Instrumentation Requirements 6.4.7 Nuclear Island / BOP Interface D
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TABLE OF CONTENTS (Continued)
Chapter /
O Section Title Volume 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 11 6.5.1 Standby Gas Treatment System (SGTS) 6.5.2 Containment Spray Systems 6.5.3 Fission Product Control Systems 6.5.4 Ice Condensers as a Fission Product Control System 6.5.5 References 6.6 INSERVICE INSPECTION OF CLASS 2 AND 3 COMPONENTS ll 6.6.1 Components Subject to Examination 6.6.2 Accessibility 6.6.3 Examination Techniques and Procedures 6.6.4 Inspection Intervals 6.6.5 Examination Categories and Requirements 6.6.6 Evaluation of Examination Results 6.6.7 System Pressure Tests 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures 6.7 MAIN STEAM POSITIVE LEAKAGE CONTROL SYSTEM (MSPLCS) 11 6.7.1 Design Bases 6.7.2 System Description 6.7.3 System Evaluation 6.7.4 Inspection and Testing 6.7.5 Instrumentation Requirements 6.8 PNEUMATIC SUPPLY SYSTEM ll 6.8.1 Design Bases 6.8.2 System Description 6.8.3 System Evaluation 6.8.4 Inspection and Testing Requirements 6.8.5 Instrumentation Requirements APPENDIX 6A IMPROVED DECAY HEAT CORRELATION FOR LOCA ANALYSIS ll xvi
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O x
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TABLE OF CONTENTS (Continued) j Chapter /
Section Title Volume 7 INSTRUMENTATION AND CONTROL SYSTEMS
7.1 INTRODUCTION
(All Systems) 12 7.1.1 Identification of Safety-Related Systems 7.1.2 Identification of Safety and Power Generation Criteria 7.2 REACTOR PROTECTION (TRIP) SYSTEM (RPS) 12 7.2.1 Description 7.2.2 Conformance Analysis 7.3 EtaGINEERED SAFETY FEATURES SYSTEM, INSTRUMENTATION AND CONTROL 13 7.3.1 Description 7.3.2 Analysis
-HPCS -Shield Building
-ADS Annu us Mixing (m) -LPCS - econdary Cgntain-ment Isolation
-RHR/LPCI -Primary Containment
-CRVICS Isolation LCS
-MSPLCS -Standby Power
-RHR/ Containment -D-G Support Systems Spray
-Essential Service
-RHR/ Suppression Pool Water 1 "9 -ESF Area Cooling
-Suppression Pool -Pneumatic Supply Makeup
-CB Atmospheric
-Combustible Gas ntrol Control ed Water ,
-SGTS 7.4 SYSTEMS REQUIRED FOR SAFE SHUTDOWN 14 7.4.1 Description 7.4.2 Analysis
-RCIC -RHR/ Shutdown Cooling n
-SLC -Remote Shutdown xvii
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Chapter /
Section Title Volume 7.5 SAFETY-RELATED DISPLAY INSTRUMENTATION 14 7.5.1 Description 7.5.2 Analysis
-Nuclenet Control -BOP Benchboard
-Supervisory Moni-
-Standby Information toring Console E UU1 -Display Control
-Rx Core Cooling BB System 7.6 ALL OTi!ER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY 14 7.6.1 Description 7.6.2 Analysis 7.6.3 Additional Design Considerations Analyses 7.6.4 References
-Neutron Monitoring -FPCCS
-Process Radiation -DW/ Containment Monitoring Vacuum Relief
-Refueling Interlocks -Vent & Pressure Control
-Leak Detection
-Rod Pattern Control
~^
-Suppression Pool
-IIP /LP System e perature Interlock Monitcring
-Recirculation Pump Trip 9
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TABLE OF CONTENTS (Continued)
(s' Chapter /
Section Title Volume 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 14 7.7.1 Description 7.7.2 Analysis 7.7.3 References
-RPV Instrumentation -Leak Detection
-Rod Control & -Rod Block Trip Information -Fire Protection
-Recirculation Flow
-Drywell Chiller &
Control Cooling
-Feedwater Control
-Plant Instrument Air
-Performance Moni-toring System -Neutron Monitoring
-Radwaste 7.8 NI/ BOP INTERFACES 14
(~N, 7.8.1 Essential Service Water (Supply) s_,/ System Instrumentation and Controls 7.8.2 Diesel Generator Fuel Oil Transfer System APPENDIX 7A I&C ELEMENTARY DIAGRAMS 15 (vh xix
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
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TABLE OF CONTENTS (Continued)
O Chapter /
Section Title Volume 8 ELECTRIC POWER ,
8.1 INTRODUCTION
16 8.1.1 Utility Grid Description ,
8.1.2 Onsite Electric Power System 8.1.3 Design Bases.
8.2 OFFSITE POWER SYSTEM 16 8.2.1 Description 8.2.2 Analysis 8.2.3 Nuclear Island - BOP Interface 8.3 ONSITE POWER SYSTEMS 16 6
8.3.1 AC Power Systems 8.3.2 DC Power Systems 8.3.3 Fire Protection of Cable Systems O
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TABLE OF CONTENTS (Continued)
(m, .
Chapter /
Section Title Volume 9 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 17 9.1.1 New Fnel Storage (High Density) 9'.l.2 Spent Fuel Storage (High Density) 9.1.3 Fuel Pool Cooling and Cleanup System 9.1.4 Fuel-Handling System .
c 9.2 WATER SYSTEMS 17 9.2.1 Essential Service Water (ESW) System 9.2.2 Closed Cooling Water System
. ,7 9.2.3 Domineralized Water Makeup System 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 9.2.6 Condensate Storage Facilities and Distribution System
()
9.2.7 9.2.8 Plant Chilled Water Systems Heated Water Systems 9.2.9 Nuclear Island / BOP Interfaces 9.3 PROCESS AUXILIARIES 17 9.3.1 Compressed Air Systems 9.3.2 Process Sampling System 9.3.3 Floor and Equipment Drainage Systems 9.3.4 Chemical and Volume Control System (PWR) 9.3.5 Standby Liquid Control System 9.4 AIR CONDITIONINi;, HEATING, COOLING AND VENTILATION SYSTEM 17 9.4.1 Control Room HVAC System 9.4.2 Fuel Building HVAC System 9.4.3 Auxiliary Building HVAC Systems 9.4.4 Turbine Building Area Ventilation System 9.4.5 Reactor Building HVAC System
_s xxi
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Section Title Volume 9.4.6 Radwaste Building IIVAC System 9.4.7 Diesel-Generator Buildings llVAC Systems 9.5 OTIIER AUXILIARY SYSTEMS 17 9.5.1 Fire Protection System 9.5.2 Communications Systems 9.5.3 Lighting Systems 9.5.4 Diesel-Generator Fuel Oil Storage and Transfer System 9.5.5 Diesel-Generator Cooling Water System 9.5.6 Diesel-Generator Starting Air System 9.5.7 Diesel Engine Lubrication System 9.5.8 Diesel Generator Combustion Air Intake and Exhaust System 9.5.9 Suppression Pool Cleanup System 9.5.10 Nuclear Island - BOP Interface APPENDIX 9A FIRE !!AZARD ANALYSIS 18 l
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SUMMARY
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Section Title Volume 10 STEAM AND POWER CONVERSION SYSTEM 10.1
SUMMARY
DESCRIPTION 19 10.2 TURBINE GENERATOR 19 10.2.1 Design Bases Functional Limitations by Design or Operational Charac-teristics of the Reactor Coolant System 10.2.2 System Description 10.2.3 Turbine Disk Integrity 10.2.4 Evaluation 10.3 MAIN STEAM SUPPLY 19 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 19 10.4.1 Main Condensers g- 10.4.2 Condenser Air Removal System
\.s / 10.4.3 Main Condenser Evacuation System 10.4.4 Turbine Bypass System 10.4.5 Circulating Water System 10.4.6 Condensate Cleanup System 10.4.7 Condensate and Feedwater System 10.4.8 Steam Generator Blowdown System (PWR) 10.4.9 Auxiliary Feedwater System (PWR)
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Section Title Volume 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS 19 11.1.1 Fission Products 11.1.2 Activation Products 11.1.3 Tritium 11.1.4 Fuel Fission Production Inventory and Fuel Experience 11.1.5 Process Leakage Sources 11.1.6 Radwaste System 11.1.7 Radioactive Sources in the Gas Treatraent System 11.1.8 Source Terms for Component Failures 11.1.9 Other Releases 11.1.10 References 11.2 LIQUID WASTE MANAGEMENT SYSTEM 19 11.2.1 Design Basis 11.2.2 System Descriptions 11.2.3 Estimated Releases 11.3 GASEOUS WASTE MANAGEMENT SYSTEMS 19 11.3.1 Design Bases 11.3.2 Main Condenser Steam Jet Air Ejector Low-Temp RECHAR System Description 11.3.3 RECHAR System Operating Procedure 11.3.4 Radioactive Releases 11.3.5 References 11.4 SOLID RADWASTE SYSTEM 19 11.4.1 Design Bases 11.4.2 System Description O
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Section Title Volume 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS 19 11.5.1 Design Bases 11.5.2 System Description 11.5.3 Effluent Monitoring and Sampling 11.5.4 Process Monitoring and Sampling 11.5.5 Calibration and Maintenance 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAM 19 v
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Section Title Volume 12 RADIATION PROTECTION 12.1 ENSURING TilAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY AClIIEVABLE (ALARA) 19 12.1.1 Policy Considerations 12.1.2 Design Considerations 12.1.3 Operational Considerations 12.2 RADIATION SOURCEE 19 12.2.1 Contained Sources 12.2.2 Airborne Radioactive Material Sources 12.2.3 References 12.3 RADIATION PROTECTION DESIGN FEATURES 19 12.3.1 Facility Design Features 12.3.2 Shielding 12.3.3 Ventilation 12.3.4 Area Radiation and Airborne Radioactivity Monitors 12.3.5 References 12.4 DOSE ASSESSMENT 19 12.5 IIEALTil PilYSICS PROGRAM 19 O
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I l Sectio _n Title Volume i
13 CONDUCT OF OPERATIONS 19 4
6
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TABLE OF CONTENTS (Continued)
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Section Title Volume 14 INITIAL TEST PROGRAM 14.1 TEST PROGRAM 20 14.1.1 Administrative Procedures (Testing) 14.1.2 Administrative Procedures (Modifications) 14.1.3 Test Objectives and Procedures 14.1.4 Fuel Loading and Initial Operation 14.1.5 Administrative Procedures (System Operation) 14.2 SPECIFIC INFORMATION TO BE INCLUDED IN SAFETY ANALYSIS REPORTS 20 14.2.1 Summary of Test Program and Objectives 14.2.2 Organization and Staffing 14.2.3 Test Procedures 14.2.4 Conduct of Test Program 14.2.5 Review, Evaluation and Approval of Test Results 14.2.6 Test Records 14.2.7 Conformance of Test Programs with Regulatory Guides 14.2.8 Utilization of Reactor Operating and Testing Experiences in the Develop-ment of Test Program 14.2.9 Trial Use of Plant Operating and Emergency Procedures
! 14.2.10 Initial Fuel Loading and Initial Criticality 14.2.11 Test Program Schedule 14.2.12 Individual Test Descriptions O
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Section Title Volume 15 ACCIDENT ANALYSES 15.0 GENERAL 21 15.0.1 Analytical Objective 15.0.2 Analytical Categories 15.0.3 Event Evaluation 15.0.4 Nuclear Safety Operational Analysis (NSOA) Relationship 15.0.5 References 15.1 DECREASE IN REACTOR COOLANT TEMPERATURE 21 15.1.1 Loss of Feedwater Heating 15.1.2 Feedwater Controller Failure -
Maximum Demand 15.1.3 Pressure Regulator Failure - Open 15.1.4 Inadvertent Safety / Relief Valve
/% Opening
'- 15.1.5 Spectrum of Steam System Piping Failures Inside and Outside of Containment in a PWR 15.1.6 Inadvertent RHR Shutdown Cooling Operation 15.1.7 References 15.2 INCREASE IN REACTOR PRESSURE 21 15.2.1 Pressure Regulator Failure - Closed 15.2.2 Generator Load Rejection 15.2.3 Turbine Trip 15.2.4 MSLIV Closures 15.2.5 Loss of Condenser Vacuum 15.2.6 Loss of Offsite AC Power 15.2.7 Loss of Feedwater Flow 15.2.8 Feedwater Line Break 15.2.9 Failure of RHR Shutdown Cooling O
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Section Title Volume 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 21 15.3.1 Recirculation Pump Trip 15.3.2 Recirculation Flow Control Failure -
Decreasing Flow 15.3.3 Recirculation Pump Seizure 15.3.4 Recirculation Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 21 15.4.1 Rod Withdraaal Error - Low Power 15.4.2 Rod Withdrawal Error at Power 15.4.3 Control Rod Maloperation (System Malfunction or Operator Error) 15.4.4 Abnormal Startup of Idle Recirculation Pump 15.4.5 Recirculation Flow Control with Increasing Flow 15.4.6 Chemical and Volume Control System Malfunctions 15.4.7 Misplaced Bundles Accident 15.4.8 Spectrum of Rod Ejection Assemblies 15.4.9 Control Rod Drop Accident (CRDA) 15.5 INCREASE IN REACTOR COOLANT INVENTORY 21 15.5.1 Inadvertent HPCS Startup 15.5,2 Chemical Volume Control System Malfunction for Operator Error) 15.5.3 BWR Transients Which Increase Reactor Coolant Inventory XXX
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O
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TABLE OF CONTENTS (Continued)
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Section Title Volume 15.6 DECREASE IN REACTOR COOLANT INVENTORY 21 15.6.1 Inadvertent Safety / Relief Valve-Opening 15.6.2 Failure of'Small Lines Carrying Primary Coolant Outside Containment 15.6.3 Steam Generator Tube Failure 15.6.4 Steam System Piping Break Outside Containment 15.6.5 Loss-of-Coolant Accidents (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary) - Inside Containment 15.6.6 Feedwater Line Break - Outside Containment 15.7 RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS 21 15.7.1 Radioactive Waste' System Leak or Failure 15.7.2 Liquid Radioactive System Failure 15.7.3 Postulated Radioactive Releases Due to Liquid Radwaste Tank Failure 15.7.4 Fuel-Handling Accident 15.7.5 Spent Fuel Cask Drop Accidents.
APPENDIX 15A PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS 21 APPENDIX 15B BWR/6 GENERIC ROD WITHDRAWAL ERROR ANALYSIS 21 i
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Section Title Volume 16 STANDARD TECHNICAL SPECIFICATIONS FOR GENERAL ELECTRIC BOILING WATER REACTORS 16.1 DEFINITIONS 22 16.1.1 Action 16.1.2 Average Planar Exposure 16.1.3 Average Planar Linear Heat Generation Rate 16.1.4 Channel Calibration 16.1.5 Channel Check 16.1.6 Channel Functional Test 16.1.7 Core Alteration 16.1.8 Critical Power Ratio 16.1.9 Dose Equivalent I-131 16.1.10 E-Average Disintegration Energy 16.1.11 Emergency Core Cooling System (ECCS)
Response Time 16.1.12 Frequency Notation 16.1.13 Identified Leakage 16.1.14 Isolation System Response Time 16.1.15 Limiting Control Rod Pattern 16.1.16 Linear Heat Generation Rate 16.1.17 Logic System Functional Test 16.1.18 Maximum Total Peaking Factor 16.1.19 Minimum Critical Power Ratio j 16.1.20 Operable - Operability 16.1.21 Operational Condition (Condition) 16.1.22 Physics Test 16.1.23 Pressure Boundary Leakage 16.1.24 Primary Containment Integrity 16.1.25 Rated Thermal Power 16.1.26 Reactor Protection System Response Time 16.1.27 Recirculation Pump Trip System Response Time xxxii
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O Chapter /
Section Title Volume 16.1.28 Reportable Occurrence 16.1.29 Rod Density 16.1.30 Secondary Containment Integrity 16.1.31 Shutdown Margin 16.1.32 Staggered Test Basis 16.1.33 Thermal Power 16.1.34 Total Peaking Factor 16.1.35 Unidentified Leakage 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 22 16.2.1 Safety Limits 16.2.2 Limiting Safety System Settings 16.B2 SAFETY LIMITS 22 16.B2.1 Bases 7
(,) 16.B2.2 Limiting Safety System Settings 16.3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 22 16.3/4.0 Applicability 16.3/4.1 Surveillance Requirements 16.3/4.2 Power Distribution Limits 16.3/4.3 Instrumentation 16.3/4.4 Reactor Coolant System 16.3/4.5 Emergency Core Cooling Systems 16.3/4.6 Containment Systems 16.3/4.7 Plant Systems 16.3/4.8 Electrical Power Systems 16.3/4.9 Refueling Operations 16.3/4.10 Special Test Exceptions O
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Section Title Volume 16.B3/4.0 Applicability 16.B3/4.1 Reactivity Control Systems 16.B3/4.2 Power Distribution Limits 16.B3/4.3 Instrumentation 16.B3/4.4 Reactor Coolant System 16.B3/4.5 Emergency Core Cooling System 16.B3/4.6 Containment Systems 16.B3/4.7 Plant Systems 16.B3/4.8 Electrical Power Systems 16.B3/4.9 Refueling Operations 16.B3/4.10 Special Test Exceptions 16.5 DESIGN FEATURES 22 16.5.1 Site 16.5.2 Containment 16.5.3 Reactor Core 16.5.4 Reactor Coolant System 16.5.5 Meteorological Tower Location 16.5.6 Fuel Storage 16.5.7 Component Cyclic or Transient Limit O
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SUMMARY
TABLE OF CONTENTS (Continued)
! Chapter / ;
! Section Title Volume. i
- 17 QUALITY ASSUIWJCE I 17.1 QUALITY ASSURANCE DURING DESIGN AND CONSTRUCTION 22 !
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! 17.2 QUALITY ASSURANCE DURING THE OPERATING PHASE 22 !
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