ML17191A301

Dresden Nuclear Power Station, Units 2 and 3, Final Safety Analysis Report

ML17191A301
Person / Time
Site:	Dresden
Issue date:	07/10/2017
From:	Commonwealth Edison Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML17191A301 (76)

Text

f FSAR INDEX . -A -Section I *. -ACAD/CAM 6.8.3.3 Acceleratioa Response Spectrum Earthquake 12. 1. l :3 Access Control Access to the Site Access, Station Accident Analysis for 7 x 7 Fuel Accident Analysis for G2 8 x 8 Fuel 14.2.2.5 Accident Analysis for ENC 9 x 9 Fuel 14.2.2.6 Accident Analysis, Radwaste 9.2.5 Accident, Control Rod Drop Procedural 14. 2 .1. 3 Acoustic Monitors 4.5.2 Acronyms and Initialisms 1.1.2.1 Action taken due to Reportable 13.6.2.2 Action taken due to Safety Exceeded 13.6.2.1 Administrative Controls 13.6 Administrative 12 .1. 4. 5 Admission Valves 6.2.3.4 Airborne Effects the Refueling Pool 14.2.2.6 Air Cleanup Appendix 8 (8-28) Air Ground Level Appendix A (2.1.1) System 10.11 11. 2. 2 Ejector Off-Gas Monitoring 7.6.2.3 Monitoring, Reactor Bldg 7. 6. 2. 5 Airlock Doors 5.3.2.2 i 1043v

.., ' FSAR INDEX -A -Analysis and Acceptance Criteria Inst & Control Analysis of Off-Site Electric Power Supply Analysis Supporting ECCS Clad Melt Criteria Analytical Methods Analytical Stability Model ANL Test Data on Clad Flailure Approval of Changes APRM Archifect -Engineer Organization Area Radiation Monitoring System As-Built Safety-Related Piping Analysis ASKE Class A Nuclear Vessels Atmospheric Control System Atmospheric Pressure, Fuel Loading Atmospheric Weather/Wind Authority to Terminate Power Production Authorization of Changes Automatic Depressurization System Automatic Vacuum Relief Auxiliary and Emergency Systems Auxiliary Power supplies Auxiliary Power System Auxiliary Systems Auxiliary Transformers ii 1043v Section 7.2.6.3 8.2.1.4 6.2.7.6 3.3.3 7.2.2.3 6.2.7:25-28 13.6 7.4 .2 Appendix E (2.3.1) 7.6.3 9.1.2 9.5.3 12 .1. 2 .4 4.1.0.1 6.8 13.8.2.1 Appendix 13.6.1 13.6 6.2.6 5.2.2.9 10.1 13.7.3.42 1.2.4.3 8.2.1.3 1.2.4.4 8.2.1.3 G

-. ! FSAR INDEX -A -Auxiliaries, Turbine Generator Availability Analysis Average Power Range Monitor (APRM) iii 1043v

• Section 13.7.3.43 6.2.7.4 7.4.5.2

• *

• Balance of Plant -Aux Systems Bases for Design Biological Shield Batteries, Station Battery Tests and Inspection FSAR INDEX -B -Bio -Assay and Medical Exam Program Bodega Bay Tests Boron Blowoff Details, Rx Bldg. Burnable Neutron Absorber Burning in Drywell Bypass Valves, Turbine 1057v i Section 13.7.3.42 12 .1.1. 3 12.2.2.l 8.2.3.2 8.3 9. 5. 5. 7 5.2.3.5a & b 9.6.1.3.2 5.3.2:1 3.5.5 6.8.1:12 7.2.6.2 FSAR INDEX -c -* Cable Pans, Electric Cask Pad CB & I CECO and GE Startup Organization Channel Hydrodynamic Conformance Change Room Facilities / -characteristics After Reactor Slowdown Charcoal Beds, Off-Gas CHASTE Chimney Chimney Effluent Monitoring '

• Circuit Breakers Circulating Water Cladding Integrity Safety Limit (Fuel) Class I Structures & Equipment Class II .Structures & Equipment Classification of Nuclear Systems Cleanup Demineralizer System Cleanup System Cleanup System (Rx Water) C02 Fire Protection System Coefficiency of Reactivity Cold Loop Startup -Transient Analysis

• Common Auxiliary Systems i 1058v Section 8.2.2.3 10 .1.2 5.2.3:24 & 25 13 .1. 2 7.2.3.2 7.2.4.2 9.5.5.4 5.2.3.3 9.2.5 6.8.3.3.4 12 .1. 2 .3 7.6.2.4 9.1 & 9.2.2.2 8.2.2 11. 2. 2 3.2.2.3 12 .1. 2 12 .1. 3 Appendix E (Exhibit 2. 7) 13.7.3.22 10.2 10.3 10.7.2:1 & 2 3.3.5.1 4.3.3:lla .& b 1.2.4.4

• * ** FSAR INDEX -c -Conununication System "' Computer, Process CONCEN Conclusions on Site and Environs Condensate Demineralizer System Condensate -Feedwater System Condensate -Feedwater Tests and Inspections Condensate Makeup Piping Conduct of Operations Conduct of Operations Construction Tests Containment Containment Atmospheric Control System I Containment Cooling System Containment Design Basis Containment Heat Removal Systems Containment Isolation Valves Containment Leakage Rate Testing Containment Penetrations Containment Response to LOCA Containment Shield Containment Spray System ii 1058v Sectfon 10.14 7 .11 8.2.2.4 6.8.3.3.4 2.4 7.8.2 13.7.3.13 11.1 11.3 11.3 10.12.2:2 13.1 thru 8 13.1 13.7.3 1.2.1.3 5.2.3:7 7. 7. 2: 1 6.8 6.2.4 Appendix 8 (B-26) Appendix B (B-26) 5. 2 .4.3; Appendix B ( B-2 7) Appendix B ( B-27) 5.2.4.2 5.2.3.2 12.2.2.2 13.7.3.34 6.2.4.2.2

• *

• FSAR INDEX -c -Containment Systems Containment Ventilation System Containment Vs Hydrogen Contractors Control and Instrumentation Control and Instrumentation, other Systems Control Curtains Control of Access to Radiation Zones Control Methods (Reactor) Control Rods Control Rod Block Function Control Rod Drive Control Rod Drive Housing Section i.2.2.4 5.l;*Appendix C 5.2.4.4 6.8.1.3 1.3 1.2.1.4 1.2.2.6 7.10 3.5.2.2 9.5.5.l 3.5.2 3.5.2.1 7.3.2:1 13.7.3.21 10.6.3 Control Rod Drive Housing Supports 6.6 Control Rod Drive Housing Support Inspection & Testing 6.6.4 Control Rod .. i>ri ve Hydraulic System 10. 6 13.7.3.17 Control Rod Drive Mechanism 3.5.3.2 Control Rod Drive System 10.6.2:1 Control Rod Drop 14.2.1 Control Rod Drop Accident Procedural Safeguards 14.2.1.3 Control Rod Housing Support 6.1.2.4 Control Rod Hydraulic System 13.7:3.17 Control Rod Isometric 3.5.2:1 Control Rod Movement 7.3.2 iii 1058v FSAR INDEX -c -* Control Rod Sequence Control Rod Surveillance and Testing Control Rod Worth Control Rod Velocity Limiter Control Room Control Room Ventilation Cooling Lake . Core Cooling Core Cooling System

• Core Internals, Thermal Shock Efforts Core Lattice Unit Core Nuclear Dynamic Characteristic Core Release, Non-Line Break Scenario Core Spray Tests and Inspection Core Spray System Core Thermal and Hydraulic Performance Crane, Reactor Building Crib.House Criteria & Bases for Design CPR Histogram for 8 x 8

• ii ii 1058v Section 14.5.2 3.5.4 3.3.4.4 6.1.2.3 6.5 12 .1. 2. 2 12.2.2.4 12.2.3 14.2.5 12.2.2.5 2.2.4.1 2.2.1:2 2.2.4:1 14.2.3.9 6.2 3.6.3.3 3.4.2:2 3.3.5 12.3.2.2 6.2.3.4 6.2.3 13.7.3.32 6.2.3:6 8.2.3 14.5.4 10.1.2.2.2 2.2.6:2 12 .1. 3 .. 3 12 .1.1. 3 3.2.2:2

• *

• FSAR INDEX -D -Data Analysis and Acceptance Criteria DC Systems Decay Ratio Dernineralizer System Description of Control Rods Description of ECCS Description of Fuel Assemblies Description of Hain Stearn Description of Reactor Vessel Internals Description of Safety Features Design Basis Accidents Design Basis Automatic Depressurization Design Basis Earthquake (Piping) Design Basis of Core Spray Design Bases Dependent On Site Characteristics Design Basis of Fuel Mechanical Characteristics Design Basis of Isolation Condenser Design Basis of LPCI Design Basis of Hain Stearn Design Basis of Nuclear Characteristics, Design Basis of Primary Containment System Design Basis Reactivity Control Mechanical Characteristics Design Basis of (Reactor) i 1045v Section 7.2.6.3 13.7.3.2 & 8.2.3.2 7.2 12.2.2.7 3.5.2.1 3.5.3 6.2.2 3.4.2 3.6.2 14.1 14.2 6.2.6 12. 1. 2. 4 .,4 6.2.3 1.2.2.1 3.4.1 4.6.1 6.2.4.1 4.4.1 3.3.1 3.5.1 3.2.1.1 3.2.1.3 FSAR INDEX -D -* Design Basis of Reactor Bldg. Design Basis of Reactor Recirculation System Design Basis of Reactor Vessel Internals Design Basis of Relief and Safety Valves Design Bases for Shielding Design Evaluation Containment System Design Evaluation (Fuel) Design Evaluation Main Stearn Design Evaluation Reactor Coolant System *Design Guide Limit Definition Design of Control Rods and Curtains Design of Electrical Systems

• Design Report, Reactor Designed Safeguards Determination of Radiation Environment Development of Technical Spec Diesel -Generator System Diesel Generator Tests and Inspection Discharge to the River Distances From Release Points Distribution System, Station Domestic Water Doppler Coefficient

• Dose, External ii 1045v Section 5.3.1 4.3.1 3.6.1 4.5.1 1.2.2:1 5.2.3 3.4.3 4.4.3 4.2.3 7.2.4.1. 3.5.2.3 8.2 Appendix 14. 2 .1. 2 12.3.3.0 3.2.4 8.2.3.1 8.3.1 13.7.3.39 8.3 9.3.3 2.2.5:1 8.2.2 13.7.3.8 D 3.3.5:1,2,3,4,5 Appendix A ( 2. 2 .1) ..

• *

• FSAR INDEX -D -Dose, Hydrogen Addition Dose to the Control Room, etal Dresden Lock and Dam Dresden Containment Certification Dresden Units 2 & 3 Map Dropout Velocities Drywell Drywell Pneumatlc System Drywell and Suppression Chamber Inspection and Testing Drywell Expansion Gap Drywell Missile Protection Drywell Spray Drywell -Torus Leak Rate Measurement Drywell Ventilation iii 1045v Section 14. 2 .1.8 12. 3 .8. 2.2.6.1 Appendix C 3.2.3.1 6.5.3 5.2.2.1 5.2.4.1 5.2.3.26 10.8.2 5.2.4 5.2.3.6 5.2.3.7 6 .2 .4 .2 .. 13.7.3.18 13.7.3.40 FSAR INDEX -E -* Earthquake Earthquake Analysis of Rx Vessel ECCS ECCS Clad Melt Criteria ECCS*Flood Protection ECCS Pipe Whip Criteria ECCS Pump NPSH Economic Generation Control Effect of KSIV Closure Time

• Effects of Postulated LOCA's EGC Operation El Centro Earthquake Electrical Penetration Seals Electric Power Electric System Electroslag Weld Report, Rx Vessel Elevated Release Point Discharge Emergency Core Cooling System

• i 1059v Section 5.2.3:8+9 12 .1.1: 2 Appendix D 1.2.2.5 6.2 6.2.7.5 Appendix B (B-23&25) 6.2.7.6 6.2.8 6.2.7.7 6. 2. 7 .9 7.3.3.l 7.3.6 14.2.3.8 1.2.5.2 7.3.3.2 7.3.6:2 12 .1.1: 2 5.2.2.4 5.3.2.3 1.2.1.5 1. 2. 2 .10 8.1 Appendix F i4. 2 .1. 7 6.1.2.1 6.2.2 6.2.7.1 Appendix B (8-25)

., FSAR INDEX -E -** Emergency Lighting Emergency Power Emergency Ventilation Engineered Safey Features Environs Radioactivity Monitoring Equipment Description, Computer Equipment Drain System Equipment Separation Equipment Supply -QA Essential Service System Exclusi_on Area Exfiltration

• Expansion Gap, Drywell

• External Dose ii 1059v Section 10.13.2 8.2.3 13.7.3.4.1 Appendix B (B-21&24) 2.3 7 .11. 3 9.3.2.1 12 .1.4. 4 Appendix E (3.3) 8. 2 .2*.4 2.2.1.6 5.3.3.l 5.2.3.6 Appendix (2.2.1)

FSAR INDEX -F Section

• Features of Plant Design 1.2.3:1 Feedwater Control System 7.8.3 Feedwater Flow, Reactor 7.5.2.4 Feedwater Nozzle Inner Bore 6.2.5.3.4 Feedwater Pumps 7.2.6.2 Feedwater Sparger Integrity 6.2.5.3.4 Feedwater System 11.1 11.3 14.2.3.5 Field Change Control Appendix E (3.4.3) Fire Alarm Systems 10.14.3 Fire Extinguishers, Portable 10.7.2 Fire Protection System 10.7

• 13.7.3.11 8.2.2.1 Fire Suppression Water System 10.7.2 & 10.7.3 Fission Product Release from the Fuel 14.2.4.2 Fission Product Transport 14. 2 .1. 6 Flange Leak Detection, Reactor Vessel 7. 5 .2 .6 Floor Drain Surge Tank 9.3.2:5I Floor Drain System 9.3.2.2 'Flow Control Recirc System 7.3.3 Flow Factor, *Kf 3.2.2.9 Flow Monitors (Recirculation) 7.4.5.2.2 Flow Regulating Station (Circulating Water/Canal) 2.2.4 Fluid Pipe Penetration 5*. 2. 2. 5

• Flux Response to Rods 14.5.3 i 1060v FSAR INDEX -F -* Fractional Control Rod Density FSAR Controlled Copy Recipient Fuel and Waste Storage Systems Fuel Assembly Isometric Fuel Cladding Integrity Safety Limit Fuel Cycle Fuel Damage Limits Fuel Design Analysis Fuel Handling

• Fuel Handling and Storage Fuel Loading Fuel Mechanical Characteristics Fuel Pool Cooling and Cleanup System Fuel Pool Damage Protection Fuel Recovery Plant Fuel Shipping Cask Fuel Storage and Fuel Handling Fuel Storage Criticality Fuel Storage Pool (Spent} --Fuel Storage Vault ii 1060v Section 3.3.4:4 1.1.1.4 Appendix B (B-29} 3.4.2:1 3.2.2.3 3.2.4.2 3.3.4.1 3.2.1.2 3.4.3.4 14.2.2.5.1 14.2.2.6 3.4.3.3 10.1 13 .1. 3. 2 13.7.3.20 1.2.1.8 1.2.2.8 13.8.2.1 3.4 10.2 13.7.3.19 10 .1.4 Appendix A (4.0} 10 .1. 2 .2 .2 10.1. 2 .3 10.1 Appendix B (B-30} 10.1. 2. 2 10.1. 2 .1 FSAR INDEX -G -*-Gadolinium Bearing Rods Gaseous Radioactive Wastes Gaseous Waste Effluents GE Startup Organization General Arrangement Crib House General Arrangement, Rx Bldg. General Arrangement, Turb. Bldg. General Conclusions General Description (Reactor) General Electric Safety Analysis General Electric Topical Reports Generating Station Emergency Plan (GSEP)

• Generator Load Rejection Gee;> logy Ground Level Radiation Dose-Guide CRD

• i 1067v Section 3.5.5.5 9.1 1.2.4.1 9.2 13 .1. 2 .1 12 .1. 3 :8 12.1.2:1-4 12.1.3:1 1.4 3.3.2 14.3 1.1.2.1 13.4.1 11.2.3.2 7.7.1.2 2.2.3 Appendix A (2.0) 6.5.2

• *

• FSAR INDEX -H -Halon System Head Cooling System (Rx) Health Physics Health Physics Instrument Inspection and Testing Heat Generation Rate Heating Boiler Heating, Ventilating, and A-C System Heat up High Density Spent Fuel Storage Rack High Neutron Flux High Primary Containment System Pressure High Radiation Sampling System (HRSS) H,igh Reactor Pressure Histogram of XN-3 Predictions HPCI HPCI Room Coolers HPCI Tests and Inspection HRSS Hydraulic Control System (CRD) Hydraulic (Reactor) Characteristics Hydro Tests Hydrodynamic Stability i 1046v Section 10.7.2 10.5 13.7.3.26 7.6.5 9.5.5 9.5.5.5 7.6.5.3 3.2.2.2 3.4.3.2 13.7.3.14 10.11 13.8.2.2 10 .1. 2: 2 7.7.1.2 7.7.1.2 9.6 7.7.1.2 3.2.2:11 6.2.5 13.7.3.33 6.2.5:1-5 8.2.3 10.9.3 6.2.5.4 9.6 3.5.3.3 3.2 13.7.3.16 7.2.2.2

• *

• Hydrogen Addition Hydrogen from Radiolysis Hydrogen from -H20 Reactions Hydrogen in Containment Effects Hydrology Hypochlorite Chemical 1046v FSAR INDEX -H -ii Section 14. 2 .1.8 6.8.1.2 6.8.1.1 6.8.1.3 2.2.4 10.9.2

• * ** FSAR INDEX -I -Identification, CRD Identification of Contractor IEEE 279 Impact Forces Industrial Facility Near Station In-Core Probe (TIP) Inerting System Initial Operating Personnel Initialisms and Acronyms Inservice Inspection Inspection and Testing of Condensate and Feedwater and Testing of Core Spray Inspection and Testing of CRD Housing Support Inspection and Testin_g of Diesel Generators and Batteries Inspection and Testing of Drywell and Suppression Chamber Inspection and Testing of Health Physics Instruments I Inspection and Testing of HPCI Inspection and Testing of Isolation Condenser Inspection and Testing of Low Pressure Coolant Injection Inspection and Testing of Off gas and Ventilation Inspection and Testing of Main Steam Inspection and Testing of Reactor i 106lv Section 14. 2 .1.1 1.3 7.4.5 14.2.3.7 2.2.2:2 5.2.2.7 8.2.2.3 6.8.3.2 13 .1.4 .1 1.1.2.1 4.3.4.2 11.3 6.2.3.4 6.6.4 8.3 5.2.4 7.6.5.3 6.2.5.4 4.5.4 6.2.4.4 9.2.4 4.4.4 3.6.4 Cr FSAR INDEX -I -Inspection and Testing of Reactor Coolant Inspection and Testing of Reactor Vessel Inspection and Testing of Recirculation System Inspection and Testing of Safety and Relief Valves Inspection and Testing of Secondary Containment Inspection and Testing of Standby Coolant Supply Inspection and Testing of Standby Liquid Controi System Inspection and Testing of Stearn Flow Restrictors Inspection and Testing of Turbine Inspection, Weld, Visual Institutional Facilities Near Station Instrument and Service Air System Instrumentation and Control Instrumentation and Control-Containment Integrated Plant Safety Assessment etal (IPSEP) Integrated System Design Evaluation Inter-Plant Effects of Accidents Interaction of Units 1,2, & 3 Interconnection, Electrical Network Intermediate Range Monitor (!RM) Introduction and Summary Iodine Activities Iodine (I-131) Release IRM ii 106lv Section 4.2.4 4.3.4 4.2.4 4.3.4 4.4.4 5.3.4 6.3.4 6.7.4 6.4.4 11. 2. 4 12.1.2.4.4.1 2.2.2:3 10.8 13.7.3.12 7.1 6.8.3.4 14.4.0 6 2. 7 1.2.4.5 1. 2 .4 8.2.1 7.4.4 1.1. 9.2.5 Appendix A (3-4) 7.4 I FSAR INDEX -I -Section Isokinetic Sample 7.6.2.4.2 Isolation Condenser Inspection and Testing 4.6.4 Isolation Condenser Vent Monitor 7.6.2.9 Isolation Condenser -Piping Diagram 4.6.2:1 Isolation Valves 5.2.2.6 5.2.4.3 13.7.3.18 Appendix B (B-27) Isotope N16 7.6.2 Isotopes in Liquid Waste Discharger 9.3.3 Investigative Function 13.6.2

• iii 106lv FSAR INDEX -J -Section

• Jet Pump Efficiency 4.3.3.1 Jet Pump Isometric 4.3.2:2 Jet Pump Operation 4.3.2.2 Jet Pump Stability 4.3.3.2 *

• i 1047v

• *

• 1048v FSAR INDEX -K -i Section

• *

• FSAR INDEX -L -Laboratory Radiation Measuring Inst Lake Land Use Leakage of Reactor Internals During Rec ire Line Break . Leakage Test, Rx Bldg Lighting System Limiting Safety System Settings Liquid Radioactive Waste Discharge Liquid Waste Effluents Liquid Waste Performance Analysis Load Diagrams Load Set Mechanism LOCA's Loe.al Limits During Operations Local Power Range Monitor (LPRK) Local Power Peaking Lock and Dam Loss-of-Control Room Loss-of-Coolant Accident Loss of EHC System Oil Pressure Loss of Feedwater Low Reactor Water Level i 1062v \ Section 7.6.5 2.2.4.1 2.2.1:2 2.2.2.2 3.6.3.5 13.7.3.41 10.13 3.2.4.1 7.6.2.8 9.3 1.2.4.2 9.3.3 12 .1. 2. 28 7.3.3.2.C 1.2.5.2 5.2.3:2 3.2.2* 7.4.5.1 3.3.4.2 2.2.6.1 2.2.6:1 14.2.5 14.2.4 11.2.3.2 7.7.1.2 11:3.3:2-3C 7. 7 .1: 2 FSAR INDEX

• LPCI LPCI Inspection and Testing LPCI Room Coolers LPRM * * *1062v -L -ii Section 6.2.4 13.7.3.33 6.2.4:1-6 8.2.3 6.2.4.4 10.9.3 7.4.5:2-8 7.4 FSAR INDEX ' -K -* Kain Condenser Condensate Kain Steam Kain Steam Flow Restrictors Kain Steam Isolation Valve "L--. Kain Steam Line Break Outside Drywell Kain Steam Line Flow Restrictor *Kain Steam Line Isolation Valve Closure Kain *Steam Line Koni toring Kain Steam Line Radiation Monitoring system Kain steam Line Restrictors

• Kain Steam System Inspection and Testing Maintenance Department* Makeup Water System MAPLHGR Flow Controller Mathematical Model Maximum Rate of Load Change Maximum Recycle System Maximum Rod Worth KCPR Mechanical Design Limits (Fuel) Mechanical Vacuum Pump System * /" i 1068v 'j Section 7.8.2 4.4 14.2.3:1 6.4 5.2.2:9 7.7.2:2 14.2.3:1 11.2.3:4-6 14.2.3 6.4.3:1 14.2.3.3 7.6.2.2 7.6.2:1 6.1.2.2 6.4.2 4.4.4 13 .1. 3. 4 10.12 13 .. 7. 3. 8 7.4 7.3.3.2 12.1.2:5-7 11.2.3.3 9.3.2:5J-M 3.3.4:6 7.4 3.4.3.1 11.2 .2 FSAR INDEX -M -Section

• Medical Exam Program 9. 5. 5. 7 Metal-Water Reactions 5.2.3.4 Meteorology 2.2.5 Meteorological Factors Appendix A (2.1) Midwest Fuel Recovery Plant Appendix A (4.0) Minimum Shift Manning Requirements 13 .1.4. 2 Missile Protection Appendix B (B-25) Mixture Impact Forces 14.2.3.7 Moderator Rod Worth 3.3.4:5 Moderator Temp. Coefficient of Reactivity 3.3.5:6 I-Moderator Void Coefficient of Reactivity 3.3.5:7

• Monitoring Systems, Personnel 9.5.5.2 Motor -Generator Sets 7.3.3 Movement of Control Rods 7.3.2 MSIV 11.2.3.2 MSIV Closure Time 14.2.3.8

• ii 1068v FSAR INDEX ' -N -Section

• N16 Isotope 7.6.2 NOT Requirements Appendix B (B-26) Nearby Facilities -Potential Hazards 2.2.2.3 NEBS Instrumentation Systems 13.7.3.36 Negative Feedback 7.2.2.1 Network Interconnection 8.2.1 Neutron Flux Level 7.4.2 Neutron Monitoring Reliability 8.2.3.2.3 New Features 1. 2. 5 New Fuel Storage Vault 10 .1. 2 .1 Noble Gas Release Appendix A (3.3) 3.2.3 4.3.2:3

• Normal Operation Characteristics NPSH NPSH for ECCS Pumps 6.2.7.9 NSS Supply, Material Appendix E (2.2.2) NSS Periodic and On-Demand Programs, Computer 7.11.3.4 Nuclear Analysis Methods 3.5.5.4 Nuclear and Process Parameters 14.5 Nuclear Characteristics 3.3 Nuclear Instrumentation 7 .4 Nyquist Plot of Open-Loop Response 7. 2. 3: 7

• i 1063v FSAR INDEX Section

• Off-Gas and Ventilation Inspection and Testing 9.2.4 Off-Gas Radiation Monitoring System 7.6.2:2 9.1 Off-Gas Treatment System 9.2.2:1 Off-Site Dose, Hydrogen Addition 14. 2 .1.8 Off-Site Electrical Power System 8.2.2.2 8.2.1.4 Off-Site Power and ECCS 6.2.7.5 Operability of the Units 1.2.5.3 On-Site Electrical Power System 8.2.2.1 On-Site Environs Radiation Monitoring System 9.5.4 Operating Basis Earthquake (Piping) 12 .1. 2. 4 Operating Basis (Reactor) 3.2.2.1 13 .1. 3 .1 3.4.3.2

• Operating Group Operating Limit Heat Generation Rate Operating Limits (Reactor) 3.2.1.3 Operating Procedures 13.3 Operational Description Recirc System 4.3.2.3 Operational Description of Recirculation Pumps 4.3.2.3 C & D Operational Design Guide and Conformance 7.2.4 Operational Training 13.2 Organization and Responsibility 13.1 Organization of Report 1.1. 2 Overall Quality Program . Appendix E (3.1) 138 KV System 8.2.1.3 13.7.3.3

• i 1049v

• *

• 115 Volt Systems FSAR INDEX 125 Volt DC Station Battery System ii 1049v Section ' 13.7.3.7 8.2.2:2

• *

• FSAR INDEX -p -Partical Closure of Main Steam Line Isolation Valves Particulate Release (Sr-90) Peak Fuel Enthalphy Pedestal, Reactor Penetrations, Testing of Performance Analysis (Rad Waste) Performance Analysis (Shielding) Performance Characteristic for Normal Operation Performance Evaluation of Reactor Vessel, Internals Performance Evaluation Recirc System Performance *Predictions Recirc System Peripheral Equipment, Computer Personnel Monitoring Systems Personnel Protection Equipment Personnel Qualifications Personnel Training Physical Description Reactor Coolant System Piping Pipe Penetrations Pipe Whip Criteria ECCS Plant Comparative Evaluation i 1069v Section 7.7.1.2 Appendix A ( 3. 5) . 14.2.1:1-3 12 .1.2. 5 Appendix B (8-27) 9.2.3 9.3.3 12.2.3 3.2.3 3.6.3 4.3.3 4.3.3.3 7.11.3.2 9.5.5.2 13.4.2.2 9.5.5.3 13.4.2.3 13 .1. 4 13.2.1:1 4.3.2.1 12 .1. 2 .4 12 .1. 3 .4 5.2.2.5 5.2.4.2 5.3.2.3 6. 2. 7 .7 Appendix B FSAR INDEX -p -Section

• Plant Description 1.2 Plant Design 1.2.3:1 Plant Effluents Appendix B (B-31) Plant Electrical Cabling 8.2.2.3 Plant Heating Boiler 13. 7 .. 3 .14 Plant Safety (SEP) 14.4.0 Plant Stability Analysis 7.2 Plot Plan 12.1.1:1 Plume Reflection Effects Appendix A (2.1.3) Pool, Spent Fuel Storage 10 .1. 2 Population Data 2.2.2.1 2.2.2:1

• Portable Fire Extinguishers 10.7.2 Portable Instrumentation 9.5.5.6 Post-Accident Radiation Levels 12.3.1-1 Potential Hazards Due To Nearby Facilities 2.2.2.3 Power Flow Map 3.2.3:3 Power Range Instruments 7.4.5 Power Transient Analysis 14. 2 .1.4 Pre-Operational Training 13.2.1 Pre-Operational Test Program 13.7 Precautionary Planning 13.4 Pressure Forces During Blowdown (Reactor) 3.6.3.2 Pressure, Reactor Vessel 7.5.2.2

• Pressure Regulator and Turbine-Generator Controls 7.8.1 ii 1069v-

• *

• FSAR INDEX -p -Pressure Suppression Chamber Primary Containment Isolation Surveillance and Testing Primary Containment Isolation System Primary Containment Sizing Primary Containment System Primary Piping Primary System Expansion Primary System Hydro Principal Design Criteria Procedural Safeguards Procedure Designations and Categories Process and Area Monitoring Process and System Equip Chart \ Process Computer Process Liquid Monitoring .Process Radiation Monitoring Property Plat Protection Protection Systems Pump Back System Purge, Vent, and Inerting System iii 1069v Section 5.2.2.3 7.7.2.4 7.7.2 5.2.3.1 5.2 4.3.4.l 13.7.3.29 13.7.3.16 1. 2 .1 14. 2 .1. 3 14.2.2.3 13.3.0:1 9.1.2 1.1.2:1 7 .11 8.2.2.4 7.6.2.7 7.6.2 1.2.2:1 9.5.5.3 7. 7 10.8.2 6.8.3.2 FSAR INDEX -Q -Section

• Quality Assurance Records Appendix E (3. 7) Quality Control Reports Appendix E *

• i lOSOv FSAR INDEX -R -* Racks, High Density Spent Fuel Storage Radiation Control Standards Radiation Dose (Fuel Pool) Radiation Levels, Post-Accident Radiation Monitoring Systems / Radiation Protection Procedures Radiation Protection Radiation (High) Sampling System Radiation Shielding (HRSS) Radiation Zones Radioactive Waste Control

• Radioactive Waste Disposal Radiological Effects Factors Radiolysis Radwaste Air Sparging System Radwaste Building Radwaste Process Systems Ventilation Ramp Rate Rate of Response (CRD)

• i 1064v Section 10.1. 2 13.4.2 10.1. 2. 2. 2 12.3 1.2.2.7 2.3 7.6 7.6.4 1.2.2.11 9.5 9.6 9.6.3.0 9.5.5.1 1. 2. 2 .12 9.1 1.2.1.6 13.7.3.35 14. 2 .1. 5 14.2.3.10 14.2.4.2 Appendix A (2.2) 6.8.1.2 10.8.2 12 .1. 3. 2 13.7.3.44 7.3.6.3 3.5.3.1 FSAR INDEX -R -* RBCCW (Reactor Building Closed Cooling Water) Reactivity Control Reactivity Insertion Accidents Reactor* Slowdown Reactor Building Reactor Building Air Monitoring Reactor Building Closed Cooling Water System Reactor Building Crane

• Reactor Building Leakage Rate Reactor Bldg Ventilation Reactor Building Ventilation Exhaust Reactor Building Ventilation Isolation Valves Reactor Building Ventilation Stack Monitoring

• Reactor Coolant System Reactor Coolant Slowdown Reactor Coolant Pressure Boundary Reactor Control Systems Reactor Core* Reactor Core and Channel Hydrodynamic Stability

• ii 1064v Section 7.6.2.7 10.10 13.7.3.15 3.3.4.3 3.3.5.1 3.5 1.2.5.1 5.2.3.3 5.3 5.3.2.1 12 .1. 2 .1 7.6.2.5 7.6.2.7 10.10 13.7.3.15 10 .1.2. 2 .2 5.3.4.1 13.7.3.41 13.7.3.44 9.2.2.1 5.3.2.4 7.6.2.6 9.1 4.1 4.1.0:1 14.2.3.6 Appendix B (8-18&26) 7.3 1.2.1.1 7.2.2.2 7. 2. 3. 3

,, FSAR INDEX -R -Section Reactor Core Conformance 7.2.4.3 Reactor Core Cooling System 1.2.1.2 Reactor Core Shutdown 14.2.3.4 Reactor Design Basis 3.2.1.1 Reactor Operating Limits 3.2.1.3 Reactor Pedestal 12 .1. 2. 5 Reactor Pressure Control 7.3.5 Reactor Pressure Vessel Design Appendix D Reactor Protection System 7. 7 .1 13.7.3.37 Reactor Protection System Surveillance and Testing 7.7.1.4 Reactor Recirculation System 13.7.3.31 Reactor Relief Valves 4.5.2 Reactor Shutdown Cooling System 10.4 Reactor Systems 1.2.2.3 3.1 Reactor Vessel 4.2 4.2.1:1 Reactor Vessel Components 13.7.3.27 Vessel Designed Cycles 4.2.1:1 Reactor Vessel Weld Report Appendix F Reactor Vessel Head Cooling System 10.5 13.7.3.26 7.6.2.7 Reactor Vessel Instrumentation Surveillance and Testing 7.5.4 Reactor Vessel Isometric 4.3.2:1 Reactor Vessel Hydro 13.7.3.16 Reactor Vessel Instrumentation 7.5 9' 13.7.3.28 iii 1064v FSAR INDEX -R -Reactor Vessel Internals Reactor Vessel Lateral Supports Reactor Vessel Nozzle Safe Ends Reactor Vessel Inspection and Reactor Vessel Supporting Structure and Stabilizers Reactor Water Cleanup Piping Diagram Reactivity Control Recipient, FSAR Controlled Copy Recirculation Flow Monitors Recirculation Line Break Recirculation Pumps Operational Description Recirculation Speed Control Network Recirculation System Recirculation System Analysis Recirculation System Inspection and Testing Records Recreational Facility Near Station Refueling Refueling Accident Refueling Accident Procedural Safeguards Refueling Pool Airborne Effects Regional and Site Meteorology Relative Bundle Power Histogram ii ii 1064v Section 3.6 4.2.2:1 4.2.2.1 4.2.2 12 .1. 2. 5 10.3.1:1 10.3.2 Appendix B (B-15) 1.1.1.4 7.4.5.2.2 3.6.3.5 4.3.2.3.C & D 7.3.3:1 4.3 13.7.3.31 4.3.3.4 4.3.4 13.5 Appendix (3.7.1) 2.2.2:3 10 .1.2 .3 14.2.2 14.2.2.3 14.2.2.6 2.2.S E 3.2.2:1 & 3 FSAR INDEX -R -Section

• Release of Activity to Environment (Liquid) 9.3.3 Appendix B (B-31) Relief and Safety Valves 4.5 13.7.3.30 Reliability of Protection Systems Appendix B ( B-12 )" Reportable Occurrence 13.6.2.2 Resumes of Startup Personnel Appendix H Review and Investigative Function 13.6.2 Ring Header 5.2.3:18-23 Rod Block Monitor (RBM) 7.4.S.3 7.4.S.4 Rod Drop Accident Analysis 12 .1.4. 6 14.2.1:4 Rod Movement Tests 7.2.6.2 7.9 13.7.3.38

• Rod Worth \

• iii ii 1064v FSAR INDEX -T -* T-Quencher Technical Spec. Development Technical Staff Temperature, Reactor Vessel Test Schedule, Pre-operational Testable Check-Isoiation Valves Testing and Surveillance (Reactor) Thermal (Reactor) Characteristics Thermal Shock Effect*s on Core Internals Thermal Shock Effects on Reactor Vessel Components Thermal Sleeves, Feedwater Nozzle TIP Topical Report (CECo)

• Topical Report (GE) Tornadoes Torus Torus Seismic Analysis Torus Water Contamination Total System Conformance Transient Operating Traversing Incore Probe (TIP) Trend Records Turbine Turbine Building i 1052v Section 4.5.2 3.2.4 13 .1.3. 3 7.5.2.1 13.7.2 6.2.3.4 3.4.4 3.2 3.6.3.3 3.6.3.4 6.2.5.3.4 7.4.2 13.2.2 1.1.2.1 2.2.5.l 5.2.2.3 5.2.3:17 6.2.7.8 7.2.3.4 7.2.4.4 3.2.4.3 5.2.2.7 7. 4. 5. 5 7.11.3.3 11. 2. 2 12 .1. 3 .1 r FSAR INDEX -T -Turbine Building Cooling Water System Turbine Building Ventilation Turbine Bypass System Turbine Condenser Turbine Generator Turbine Generator Controls Turbine Generator System Turbine Plant Control Systems Turbine Steam Handling Equipment Turbine Stop and Bypass Valves Turbine Stop Valve Closure Turbine System

• Turbine Tests and Inspection Turbine Trip With Flux Scram Turbine Trip Without Bypass Turnkey Projects Operation Typical Core Lattice Unit 345 KV System 220 Volt and 115 Volt Ac Systems 250 Volt DC Station Battery System ii l052v Section 13.7.3.10 10.9.2 13.7.3.44 11. 2. 2 11. 2. 2 11.2 13.7.3.43 7.8.1 11. 2. 2 7.8 12.2.2.6 11. 2 .4 7.7.1.2 1.2.2.9 11. 2 .4 4.5.3:2a&b 11. 2. 3 3.2.2:10 Appendix E (2.2-1) 3.4.2:2 8.2.1.2 13.7.3.4 13.7.3.7 8.2.2:1

• *

• FSAR INDEX -s -Standby Lighting Standby Liquid Control System Standby Liquid Control System Inspection and Testing Startup and Power Test Program Startup Program, Preoperational Startup Tests Inst and Control Station Access Station Arrangements -station Batteries Station Computer Power Supply Station Distribution System Station Fire Protection System Station Generated Procedures Station Grounding-Construction Tests Station Instrument and Service Air System Station Organization/Management Station Procedure Designations and Steady State Steam Flow Steam Flow Restrictors Steam Handling Equipment, Turbine Steam Jet Air Ejectors Stock System Structures anq Equipment iii 105lv Section . 10.13 .3 6.7 7.3.4 13.7.3.25 6.7.4 13.8 13.7.1 7.2.6.2 13.4.3 1.2.2.2 2.2.1:1 8.2.3.2 8.3.2 8.2.2.4 8.2.2 10.7 & 13.7.3.11 13.3 13.7.3.1 10.8 13 .1. 3 13.3.0:1 3.3.4 7.5.2.5 6.4 12.2.2.6 . 11. 2. 2 9.4.2.1 12 .1.1.1

• *

• FSAR INDEX -s -Section Structural Design and Shielding 12.l Stock Rod Margin 3.3.4:3 Summary Evaluation of Safety

• 1.2.2.13 Summary of Off-Site Doses from Accidents 1.2.2:2 Summary of Pre-operational Test Content & Sequence 13.7.3 Summary of Technical Data 1.2.3 Supplementary Control 3.5.5 Suppression Chamber and Drywell Inspection and Testing 5.2.4 Surveillance and Testing of Control Rods 3.5.4 Surveillance and Testing of Nuclear Instruments 7.4.5.6 Surveillance and Testing of Primary Containment Isolation 7.7.2.4 Surveillance and Testing of Reactor Surveillance and Testing of Reactor Protection System 3.4.4 3.5.4 7.7.1.4 Surveillance and Testing of Reactor Vessel Instrumentation 7.5.9 System Performance Transients 6.2.7.2 iiii 1051v \

FSAR INDEX )* -T -* Technical Spec. Development Technical staff Temperature, Reactor Vessel Test Schedule, Pre-operational Testing and Surveillance (Reactor) Thermal (Reactor) Characteristics Thermal Shock Effects on Core Internals Thermal Shock Effects on Reactor Vessel Components Thermal Sleeves, Feedwater Nozzle TIP Topical Report (CECo) Topical Report (GE)

• Tornadoes Torus Torus Seismic Analysis Torus Water Contamination Total System Conformance Transient Operating Conditions Traversing lncore Probe (TIP) Trend Records Turbine --Turbine Building Turbine Building Cooling Water System

• i 1052v Section 3.2.4 13 .1. 3. 3 7.5.2.1 13.7.2 3 .4 .4. 3.2 3.6.3.3 3.6.3.4 6.2.5.3.4 7.4.2 13.2.2 1.1.2.1 2.2.5.1 5.2.2.3 5.2.3:17 5.2.3:2 6.2.7.8 7.2.3.4 7.2.4.4 3.2.4.3 5.2.2.7 7. 4. 5. 5 7.11.3.3 11.2 .2 . 12. 1. 3 .'1 13.7.3.10 10.9.2 FSAR INDEX -T -Section

• Turbine Building Ventilation 13.7.3.44 Turbine Bypass System 11. 2. 2 Turbine Condenser 11. 2. 2 Turbine Generator 11. 2 13.7.3.43 Turbine Generator Controls 7.8.1 Turbine Generator System 11. 2. 2 Turbine Plant Control Systems 7.8 Turbine Steam Handling Equipment 12.2.2.6 Turbine Stop and Bypass Valves 11. 2 .4 Turbine Stop Valve Closure 7.7.1.2 Turbine System 1.2.2.9 Turbine Tests and Inspection 11. 2 .4

• Turbine Trip With Flux Scram 4.5.3:2a&b 11. 2. 3 Turbine Trip Without Bypass 3.2.2:10 Turnkey Projects Operation Appendix E (2.2-1) Typical Core Lattice Unit 3.4.2:2 345 KV System 8.2.1.2 13.7.3.4 220 Volt and 115 Volt Ac Systems 13.7.3.7 250 Volt DC Station Battery System 8.2.2:1

• ii 1052v FSAR INDEX -u -Ultimate Performance Limit Criteria Ultrasonic Resin Cleaners Unit Auxiliary Power Supplies Unit Control and Instrumentation Unit-1 Spent Fuel Updated FSAR i 1053v Section 7.2.3 9.3.2.4 1.2.4.3 1.2.2.6 10.1. 2. 2 .1 1.1.1.3 1.1.1.4 FSAR INDEX -v -* Vacuum* Pump System Vacuum Relief Velocity Limiter, CRD Vent Pipes Vent, Purge, and Inerting Systems Venting and Cooling System Ventilating Ventilation and Off-Gas Inspection and Testing Ventilation, Control Room Ventilation, Drywell Ventilation, Emergency Ventilation, Reactor, Radwaste, and Turbine Bldgs Ventilation Stack Monitoring, Reactor Bldg

• Ventilation System Containment Venturis, Hain Steam Line Vessel Components, Reactor Vessel Head Cooling System Vessel Instrumentation Vibration of Components (Rx Internals) Visual Weld Inspection Vulkene Insulation i 1054v Section 11. 2. 2 5.2.2.9 6.2.5 5.2.2.2 6.8.3.2 5.2.2.8 10.11 9.2.4 12.2.2.5 13.7.3.40 13.7.3.41 13.7.3.44 7.6.2.6 9.2.2.1 5.2.4.4 6.4.2 13.7.3.27 10.5 13.7.3.28 3.6.3.1 12.1.2.4.4.1.3 8.2.2.3 Waste Concentrator System Water Level, Reactor Vessel Water System (Clased Cooling) Water System (Service) Weather, Wind Weld Inspection, Visual Well Water System Wind WINDOW 1065v FSAR INDEX -w -i Section 9.3.2.3 7.5.2.3 10.10 10.9 Appendix G 12.1.2.4.4.1.3 10.12,2:1 Appendix G 6.8.3.3.4

• Xenon Equilibrium Xeno.n Stability X-Area Coolers

• lOSSv FSAR INDEX -x -i Section 6.7.1 3.3.5.2 7.2.4.S 10.9.2 10.9.3

• *

• 1066v FSAR INDEX -y -i Section

• 1056v FSAR INDEX -z -. i Section -6. 8 .1.1 v TABLE OF CONTENTS -DRESDEN UNITS 2 & 3 UPDATED FINAL SAFETY ANALYSIS REPORT SECTION 1 INTRODUCTION AND SUMMARY 2 SITE 3 REACTOR CORE AND INTERNALS 4 REACTOR COOLANT SYSTEM 5 CONTAINMENT SYSTEMS 6 ENGINEERED SAFEGUARDS 7 CONTROL AND INSTRUMENTATION 8 ELECTRICAL SYSTEM 9 RADWASTE SYSTEM 10 REACTOR AUXILIARIES 11 TURBINE AND CONDENSATE SYSTEMS 12 STRUCTURES AND SHIELDING e. 13 CONDUCT OF OPERATION 14 SAFETY ANALYSIS APPENDIX A CHIMNEY RELEASE RATE CALCULATION B PLANT COMPARATIVE EVALUATION WITH DESIGN CRITERIA c CONTAINMENT CERTIFICATIONS D UNIT 2 REACTOR PRESSURE VESSEL DESIGN E QUALITY CONTROL F REACTOR VESSEL ELECTROSLAG WELD REPORT G METEOROLOGICAL DATA H RESUMES FOR STARTUP PERSONNEL -e

• . *'

• 1.1 1.1.1 1.1.1.1 1.1.1.2 1.1.1.3 1.1.1.4 1.1.2 1.1.2.1 1.1.2.2 1.2 1.2 .1 1.2.1.1 1.2.1.2 1.2.1.3 1.2.1.4 1.2.1.5 1.2.1.6 1.2.1.7 1.2.1.8 1.2 .2 1.2.2.1 1.2.2.2 1.2.2.3 1.2.2.4 1.2.2.5 1.2.2.6 1.2.2.7 1.2.2.8 1.2.2.9 1. 2. 2 .10 1.2.2.11 1. 2. 2 .12 1. 2. 2 .13 1.2 .3 1. 2 .4 1.2.4.1 1.2.4.2 1.2.4.3 1.2.4.4 e 1.2.4.5 0013f OOOlf TABLE OF CONTENTS SECTION 1 --INTRODUCTION*AND SUMMARY PURPOSE AND ORGANIZATION OF REPORT

• PURPOSE OF REPORT Introduction Purpose and Scope of Safety Analysis Report Updating of Original FSAR FSAR Controlled Copy Recipient ORGANIZATION OF REPORT General Format Revisions PLANT DESCRIPTION PRINCIPAL DESIGN CRITERIA Reactor Core Reactor Core Cooling Systems Containment Control and Instrumentation Electrical Power Radioactive Waste Disposal Shielding and Access Control Fuel Handling and Storage SUMMARY DESIGN DESCRIPTION AND SAFETY ANALYSIS Design Bases Dependent On Site Characteristics Station Arrangements Reactor Systems Containment Systems Shutdown Cooling System and ECCS Unit.Control and Instrumentation Radiation Monitoring Systems Fuel Handling and Storage Turbine System Electrical System Shielding, Access Control, and Radiation Protection Procedures Radioactive Waste Control Summary Evaluation of Safety SUMMARY OF TECHNICAL DATA INTERACTION OF UNITS 1, 2, & 3 Gaseous Waste Effluents Liquid Waste Effluents Unit' Auxiliary Power Supplies Common Auxiliary Systems Inter-Plant Effects of Accidents Rev. 4 June 1986 1i l.Ll-1 1.1.1:-1 1.1.1-1 1.1.1-1 1.1.1-2 1.1.1-2 1.1. 2-1 1.1. 2-1 1.1.2-1 1. 2 .1-1 1.2.1-1 1.2.1-1 1. 2 .1-2 1. 2 .1-2 1. 2 .1-3 1. 2 .1-3 1. 2 .1-3 1. 2 .1-3 1. 2 .1-4 1.2.2-1 1.2.2-1 1. 2. 2-3 1.2.2-3 1. 2. 2-4 1. 2. 2.,... 7 1. 2. 2-8 1.2.2-9 1. 2. 2-9 1. 2 .2-10 1. 2. 2-10 1. 2. 2-10 1.2.2-11 1.2.2-11 1. 2. 3-1 1.2.4-1 1.2.4-1 1.2.4-1 1. 2 .4-2 1. 2 .4-2 1. 2. 4-4 TABLE OF CONTENTS (Contd.) SECTION 1 --INTRODUCTION AND SUMMARY 1.2.5 NEW FEATURES 1.2.5.1 Features Reduce the Probability and Magnitude of Potential Reactivity Insertion Accidents 1.2.5.2 Features Which Mitigate Effects of Postulated LOCA1s 1.2.5.3 Features Which Improve Operability of the Units 1.3 IDENTIFICATION OF CONTRACTORS 1.4 GENERAL CONCLUSIONS 1 ii 1.2.5-1 1. 2. 5-1 1.2.5-1 1.2.5-2 1. 3. 0-1 1.4 .0-1

• *

• 1.1.2:1 1.1.2:2 1.2.2:1 1.2.2:2 -1.2.3:1 LIST OF TABLES --SECTION INTRODUCTION General Electric Company Topical Reports Acronyms and Initialisms Design Bases For Shielding Rev. 2 June 1984 liv S1.DT1mary of Maximum Off-site Doses From Postulated Accidents Principal Features of Plant Design

• • • ... e . . ... : .. . *: r . : . I :.*; .. . *. . . . . . . . . . * .. LL2:1 1.2.2:2 .* 1.2.3:1 \ '* .') *! .. .*LIST OF TABLES SECTION INTRODUCTION Rev. 1 June 1983 liv General Electric Company Jbpical Reports I Design Bases For Shfelding

• Summary of Maximum Off-site Doses From Postulated AcCidents Principal Features of Design* *. *' *. ' . . *'. . ... .*.-.* .. **-* *! . .* .. :. .... :-......... -. '* *'. . _: . ..,. t * * *

• l

,olJ ** -, ..... **-*------** *------:. *

• 1.1.1.3 Updating of original FSAR Rev. 4 June 1986 1.1.1-2 This documertt is the Updated Final Safety Analysis Report {UFSAR), a report separate and distinct from the original Final Safety Analysis Report. The original FSAR and the associated docket files {50-237 and 50-249) are the basis for the licensing of the plant. In the event that a discrepancy exists between the original FSAR and the UFSAR, the original FSAR will be the final authority. The Technical Specifications may reference the UFSAR. The UFSAR is revised annually as required in 10 CFR 50.7le. The UFSAR is designe*d to serve as a reference document, reflecting the current configuration of the plant, including information and analyses required by and submitted to the NRC since submission of the original FSAR, and containing the information in a contiguous format. 1.1.1.4 FSAR Controlled Copy Recipient

Subject:

FSAR Update Dresden Station has reviewed the FSAR for rev1s1ons, corrections, *and material information additions. The changes contained herein will become Revision 4 {June, 1986) to the FSAR . The changes are in compliance with the 10 CFR 50.71{e) requirement to identify changes and which references the requirements defined in 10 CFR 50.59. The 50.59 report refers to changes in the facility as described in the FSAR, changes in procedures described in' the FSAR, and tests or experiments not described in the {original) FSAR. All changes which have been implemented were previously reviewed to the 50.59 criteria and in our opinion do not constitute any additional unreviewed safety questions. Dated 0013f OOOlf n Manager Dresden Nuclear Power Station

• *

• APR APRM ASME BTP BWR CE Co CFR CSE CST CVTR DBE DER DG ECCS EHC EI&C FSAR FTOL FWCI GDC GE gpm HEPB hp HPCI IE IEEE IP SAR IREP IRK LCO LER LOCA LPCI LPRM LWR MCC MCPR MDC MOV mph MSIV MSL MWe MWt NRC ORNL PMF PMP POL 0013f OOOlf TABLE 1. 1. 2: 2 ACRONYMS AND INITIALISMS Automatic Pressure Relief average power range monitor American Society of Mechanical Engineers Branch Technical Position boiling-water reactor Commonwealth Edison Company Code of Federal Regulations Containment Systems Experiments condensate storage tank Carolina Virginia Tube Reactor design-basis event design electrical rating diesel generator emergency core cooling system electrohydraulic control electrical instrumentation and control Final Safety Analysis Report full-term operating license feedwater coolant injection General Design Criterion(a) General Electric Company gallons per minute energy pipe break horsepower high-pressure coolant *injection Off ice of Inspection and Enforcement of Electrical and Electronics Engineers Integrated Plant Safety Assessment Report Integrated Reliability Evaluation Program intermediate range monitor limiting condition for operation licensee event report loss-of-coolant accident low-pressure coolant injection low power range monitor light-water reactor motor control center. minimum critical power ratio maximum dependable capacity motor-operated valve miles per hour main steam isolation valve

• mean sea level megawatt-electric megawatt-thermal U.S. Nuclear Regulatory Commission Oak Ridge National Laboratory probable maximum flood probable maximum precipitation provisional.operating license Rev. 3 June 1985

• *

• PRA psi psig PWR RBCCW RCPB RPS RSCS RWCU SALP SAR SBGTS SEP SER -SOAD SRP STS **sws TMI UHS USI 0013f OOOlf TABLE 1.1.2:2 .(Cont'd) probabilistic risk assessment pounds per square inch pounds per square inch gage pressurized-water reactor reactor building closed cooling water reactor coolant pressure boundary reactor protection system reactor shutdown cooling system reactor water cleanup Systematic Appraisal of Licensee Performance safety analysis report standby gas treatment system Systematic Evaluation Program safety evaluation report Station Operational Analysis Department Standard Review Plan Standard Technical Specification service water system Three Mile Island ultimate heat sink unresolved safety issue Rev. 3 June 1985 1.2.2-4 The core is assembled in modules of four fuel assemblies set in the interstices of a crucifonn control rod. This modular core fonn, common to all General Electric boiling water reactors, permits substantial increase in thennal power with a small increase in core diameter and at the same time preserves the reactivity control characteristics demonstrated in the *operation of Dresden Unit 1 and other General Electric power reactors. The reactor pressure vessel contains the reactor core and structure, steam separators and dryers, jet pumps, control rod guide tubes, and water, emergency core cooling system (ECCS), and standby liquid control spargers and other components as shown in Figure 3.6.2:1. The inside diameter of the vessel is approximately 21 feet and the inside height between heads is approximately 68 feet. The main connections to the reactor vessel include the steam lines, jet pump lines, feedwater lines, and control rod drive thimbles. Other connections are provided for the isolation condenser system, standby liquid control system, ECCS, and The fuel for the reactor core consists of uranium dioxide pellets contained in sealed Zircaloy-2 tubes. These fuel rods are assembled into square arrays in individual blies. The original assemblies were of a 7x7 configuration; later designs introduced in subsequent fuel cycles were of an 8x8 configuration. The fuel enrichment is varied from rod to rod within an assembly to achieve desired neutron flux characteristics. Some water rods may be included, and gadolinium is used in some rods as a burnable poison, in the fonn of Gd203-U02* Each fuel assembly is surrounded by a Zircaloy-4 flow channel. Water serves as both the moderator and coolant for the core. The control rods consist of assemblies of 3/16-inch diameter, sealed, stainless steel tubes filled with compacted boron carbide powder and held in a crucifonn array by a stainless steel sheath of 1/16 inch wall thickness fitted with castings at each end. The design of such control rods is almost identical with those which have been used successfully in Unit 1 for more than six years except that control rods of current design are longer *due to the use of longer fuel assemblies. The control rods are of the bottom entry type and are moved vertically within the core by individual, hydraulically operated, locking piston type control rod drives . The control rod drive hydraulic system is designed to allow control rod withdrawal or insertion at a limited rate, one rod at a time , for power level control and flux shaping during reactor operation. Stored energy available fran gas charged accumulators and from reactor pressure provides hydraulic power for rapid simultaneous insertion of all control rods for reactor shutdown. Each drive has its own separate control and scram devices. The systems for reactivity control are of the same design as those used in the Oyster Creek and Nine Mile Point Plants, including two features which provide improved plant safeguards.

1.2.3-1 1.2 .3 SUMMARY OF TECHNICAL DATA Design features and data appropriate to achieve a reactor thermal output of 2527MW are summarized in Table 1.2.3:1. TABLE 1.2.3:1 PRINCIPAL FEATURES OF PLANT DESIGN Site Location Size of Site Site and Plant Ownership Plant Net Electrical Output Gross Electrical Output Net Heat Rate Feedwater Temperature Thermal and Hydraulic Design Design Thennal Output Reactor Pressure (dome) Steam Fl ow Rate Recirculation Flow Rate Fraction of Power Appear-ing as Heat Flux Power Density Heat Transfer Surface Area/ Assembly Average Heat Flux Maximum Heat Flux Maximum U02 Temperature Average Volumetric Fuel Temp. Core Subcool i ng Core Average Void Fraction, Active Coolant Core Average Exit Quality Minimum Critical Power Ratio Safety Limit GE 7x7 41.08 i ter 86.52 ft 2 131,200 Btu/(hr-ft2) 405,000 Btu/(hr-ft ) 3470°F 1050°F 22.4 Btu/lb 0.299 0.101 1.06 Dresden Site, County of Grundy, State of Illinois 953 Acres plus 1275 acre cooling lake Commonwealth Edison Company 809 MW 850 mi 10,648 Btu/kw-hr 340.1 F 2527

• 1020 psia 6 9.765 x610 lb/hr 98 x 10 lb/hr 0.965 GE 8x8 41.09 97.6 117 ,100 354,400 1.06 GE 8x8R/P8x8R 40.74 94.9 120,400 362,000 1.07 1.2.3-2 TABLE 1.2.3:1 (Contd.) PRINCIPAL FEATURES OF PLANT DESIGN Nuclear Design Initial Fuel Enrichment: ( 7x7 assembly) Typical Reload Fuel Enrichment: (8DRB265H 8x8 assembly) Water/U02 Volume Ratio Core Average Neutron Flux Thenna 1 1 Mev GE 7x7 2.41 Burnup target (average assembly) Power Coefficient for xenon stability Heat flux peaking factors: Relative Assembly Axial Local Overpower Gross . Reactivity Control: Cold shutdown keff all rods inserted Cold shutdown k ff rod of maximum worth stuck fO out Enrichment No. of rods Wt % U-235 per assembly 2.44 30 1.69 16 1.20 3 3.8 14 3.0 27 2.4 2 2.0 14 1. 7 4 1.3 1 water rods 2 GE GE 8x8 8x8R 2.60 2.76 13 2 3.50 x 1013n/cm2-sec 3.67 x 10 n/cm -sec 28 ,ooo MvJD/ton More negative than -.Ol(dK/K)/(dP/P) Design Operating 1.47 1.47 1.57 1.57 1.30 1. 30 1.20 3.60 3.00 0.96 0.96 0.99 0.99 TABLE 1.2.3:1 (Contd.) PRINCIPAL FEATURES OF PLANT DESIGN Standby liquid control shutdown, dkeff Minimum Critical Power Ratio: Linear Heat Generation Rate (kw/ft): 7x7 fuel GE 8x8 fuel ENC fuel Approximate Coefficients: Moderator Coefficient [ ( d k/ k ) I ° F J Moderator Void Coefficient [ ( dk/k) /% Void] Fuel Temp. (Doppler) Coefficient [(dk/k)/°F] Excursion Parameters: Design 0.16 1.07 17.5 13.4 14.9 Hot Cold (no voids) -8.9xl0-5 -17.0xl0-5 -3 less than_3 -1.0xlO -1.2xl0-5 1. 2. 3-3 Operating 1.39 17.5 13.4 14.9 Operating -1.4x10-3 -1.2x10-5 1* Prompt Neutron Lifetime .B Effective Delayed Neutron Fraction 48.9 microseconds .0058 Core Equivalent Core Dia. Circumscribed Core Diameter Core Lattice Pitch Number of Fue 1 ,l\ssemb 1 i es Fuel Assembly Fuel Rod Array Fue 1 Rod Pitch Weight of U02 per Fuel Assembly Channel Material Total Assbly plus Channel Weight Fuel Rods Water Rods 182. 2 inches 189.7 inches 12 inches (4 assemblies/unit cell) GE 7x7 7x7 724 0.738 in. 492.5 lbs. Zircaloy-4 678.9 lbs. 49 0 GE 8x8 8x8 0.640 458.6 Zircaloy-4 650 63 1 GE 8x8R/Px8x8R 8x8R/P8x8R 0.640 441.6 Zircaloy-4 650 62 2 ENC 8x8 P8x8 0.641 434.4 Zircaloy-4 580 63 1 Fuel Rod, Cold Fuel Pellet Dia. Cladding Thickness Cladding O.D. Active Fuel Length Lgth of Gas Plenum Fuel Material Cladding Material Fi 11 Gas Fill Gas Pressure TABLE 1.2.3:1 (Contd.) PRINCIPAL FEATURES OF PLANT DESIGN GE GE GE 7x7 8x8 8x8R/Px8x8R 0.488 in. 0.416 0.410 0.032 in. 0.034 0.034 0.563 in. 0.493 0.483 144 in. 144 145.24 11.22 in. 11.24 9.48 U02 U02 U02 Zircaloy-2 Zircaloy-2 Zircaloy-2 He He He 1 atm 1 atm 1 atm/3 atm Movable Control Rods Number Shape Pitch Stroke \4 i dth 177 Cruciform 12.0 in. 144 in. 9.75 in. 143 in. 1. 2 .3-4 ENC 8x8 0.405 0.035 0.484 145.24 10.06 U02 Zircaloy-2 He 3 atm Control Length Control Material Number of Cntrl Mtrl c granules in stainless steel tubes and sheath Tubes per Rod Tube Di mens i ans 0.188 in. o.d. x 0.025 in. wall Temporary Control Curtains Number Shape Width Thickness Control Length Control Material Curtain Locations Burnable Neutron Absorber Control Material

• Location Concentration Reactor Vessel Inside Diameter Overall Length Inside Design Pressure 340 Flat sheet 9.20 inches 0.0625 inches 141.25 inches Stainless steel containing 5400 ppm natural boron Between fuel assemblies in water gaps without control rods. Gd203 Mixed with U02 in several fuel rods per fuel assbly Location and reload dependent. 20 ft.-11 in. 68 ft.-7-5/8in. 1250 psig

• • • -.. ***e I*.'. *-'. ' . , ... 2.1 . 2.2 2;2;1 2 .. 2 .1.1 2:. 2.1.2 2.2.1.4' 2.2.1.5 2.2.1.6 2. 2. 2 . 2.2.2.1 2:2.

2.2 INTRODUCTION

TABLE OF CONTENTS SECTION 2 --SITE DESCRIPTION Of SITE AND ADJACENT *AREAS SITE Site Size and Location , : . Site Ownership .

• Location of the Units on the.Site Activities on. the Site Access to the Site

• Exel us ion Area . POPULATION AND LAND USAGE IN ADJACENT AREAS Popu 1 at ion Data Land Use . . * '2.2.*2.J POTENTIAL. HAZARn°S DUE TO)IEARBY FACilITIES .* 2 2 '2 3 1 :INTRODUCTION: ** .. * * .. . * .. * .* ' *. "* , .. ' . .. 2*:i:*2 .. : f 2 * .. HAZARDS FROM* EXPLOSIONS. *. .* ..

• i . * .. 2:2.2.3.2.1 *

• industrial Facilities* .2.2.2.J .. 2.2

• 2. 2'. 2. 3. 2. 3 Rail way Transportatfori . vJaterway Transportation 2.2.2.3.2.S: Military Facilities 2.2.2.3.2.6

• Pipelines 2. 2 .. 2. 3. 3

• HAZARDS FROM. VAPOR CLOUDS AND FIRES *.* HAZARDS FROM TOXIC CHEMICALS. 2.2.2.3.5 HAZARDS FROM COLLISION WITH THE INTAKE STRUCTURE . HAZARDS FROM LIQUID SPILLS. . . . ' 2*. 2 *. 2 *. 3. 7 HAZARDS FROM AIRCRAFT. * *2. 2. 2. J. 7 .1 Airports 2.2 * .Z.3.T.2 Airways* 2 .. 2:.2*.

3.8 CONCLUSION

S

• REFERENCES Rev. 1 June 1983 2i Page . 2.1. 0-1 2. 2.1-1. 2.2.-1-1*. 2. 2.1-1 ... * '*2 .. 2.1-1 2.2.1-1:' . 2.2.1-2 2.2.1-2 2.2.1-3; 2.2.2-1 ' *,* 2. 2::.6: *_<;, *: :. ': 2*; 2'. 2:;.6 .* '.**,*. *: :. ;*; ' ': << 2.2.2..:.6'. *: ; .2.2*. 2-8 2 ... 2. . 2. 2. 2;..10 2.2.2-10.*. 2. 2 .. 2-1 L 2.2. 2-n.,. * .. 2 .. 2.2-11' 2. 2.2-n z. 2. 2:.12 2.2.2-12 2. 2. 2-1.4 2.2.2;..15 2.2.2-16

• *

• 2.2.1:1 2.2.1:2 2.2.2.3:1 2.2.2.3:2 2.2.4:1 2.2.4:2 2.2.6:1 2.2.6:2 LIST OF FIGURES --SECTION 2, SITE Station Property Plan Cooling Lake General Arrangement Dresden Nuclear Power Station Area Map Rev. 2 June 1984 2i ii Pipelines Considered in the Evaluation of Hazard From Explosion Cooling Water Flow Diagram --Unit 2/3 Dresden Cooling Lake Dam Site Flow Diagram at Illinois River Above Dresden Island Lock and Dam General Arrangement .--Crib House 2.2.2:1 2.2.2:2 2.2.2:3 2.2.5:1 LIST OF TABLES --SECTION 2, SITE Population Centers .Surrounding Station Industrial Facilities Near Station 2iii Recreational and Institutional Facilities Near Station Distances From Release Points To Various Points Near Site

.I.: **e* ' ....... ' -' ' ' **Table Assessment Summary HAZARD *NUMBER l* 2 3 4 5 6 8 9 10 11 12 SOURCE OF HAZARD ' ' Explosion from:*. Industrial facilities* Highway transportation Railway transportation Watt;!rway transportation Military facilities Pipelines Vapor cloud expiqsion & fire from waterway transporation Toxic chemicals Collis{on with intake structure Li quid spi 11 s Aircraft .impact from: Airports Airways* ** .. * **.: ... * .' '1 . . . . '. . . . .*. REPORT . * ... SECTION*. ,2.1 .. *2 .. 2 ' *2:3' ' '2.4' ,* 2. 5 2.6 3.t '*, '. 4 5 '.< . *; 7:*1: T:f DESIGN BASIS EVENT? . .:*. No, based Qh adequatE? separation distance** No ? based on separation distance No, .based on. adequate separation distance No, based on adequate separation No, based .on adequate separatigg distance No, based on frequency of 6x10 /yr using conservative -7 No, pased on of 4xl0 /yr Not part of SEP U-1.C Nq, based on physii:al considerations *.No, based on physical considerations ' .** *.. ; -7 ' No, based on frequency of 3.24x10 l¥ear based on of 0.93 x 10 /year 'l'rData for facilities which responded to. the * . **There is one exception to this conclusion .:. .storage tank pn the Reichhold Chemical site.* :* * .. ; ** <r . ... * .*' , *:*.;

Table 2.2.2.3:2 Industries Within 5 Miles I **, e. Dresden Station (Ref. 18) . *,' . *. ;,* DISTANCE (MILES) INDUSTRY & DIRECTION . GE BWR Training Center & Spent Fuel Storage Reichhold Chemicals A .. P. Green-* Atrco 1ndustrial Gases Northern Illinois Gas Alumax Mill Products

• Northern Petrochemicals Northern Petrochemical Dock

• ARMAK Chemicals * *

• Dur.kee Chemicals . . . , Truck Tennina*i Dow Chemicals Dow Chemical *Dock 'Exxo_n (chemical plant) Hydrocarbon Transportation, Inc. Streator Industrial Supply Mobil Chemical Jal iet Livestock Market

• Mo_bn O:il Refi-nery Commonweal th Edison Co * . Collins 0. 7 -: 1. 6. -w .

• 2. 1 "". SSW-*

• 2.S NW

• 2*,5_,.. NW 2.8 -MW 3. 3 -* MW . 1 -* W* 3.6 -WNW :. *.*, 2>-.EN{ J.6*;,. ENE . 3. 7 -E . 2. 7 -* E J.9 -ENE 4.0 ..: NW 4. 0 -.s -4.1 -NE 4.2 -ESE .. 4. 5 -NE 5 *. 0 -WSW PRODUCT Spent nuclear fuel storage Resins and chemicals Br.iCk and clay co . 2 . ! I. Natl,J na 1 gas Aluminum sheet and co.il ethyl oxide glycol* Fatty nitrogen chemi.CaJs .; Ed.i b le. oi-l -* . .

• Under construction *. * . . ., .. Polystyrene** pla-stic.

• Under construction .Propane Industrial supplies . Po.lystyrene sheets. & crystal Livestock Petroleum** products Electricity* ;... ..

. , a.,, .* t *** '.:.:* *. *.; .. . *' l'.:. .. i ..*. ; ' ' ' ' :.:: .. ; *.:'".{ . . . . .\* ... . -. ' ' ., :. ,,,._-.. . ._ ..... ** ....... . e *.** Table2.2.2.3:3 Dresden Island Traffic Statistics Fiscal Years l9j8 (Refs. 6, 11) . FISCAL YEAR Total commodities, tons x 10 Hazardous mate5ials,*

• tons x 10 . Liquefied Gases,** tons 1973 28.476 5.653 . o.o* 1974 1975 1976 30.853 27.808 25.882 6.073 5.358 5.059 0.0 O*.O *

• 17 ,992 *Hazardous materials are defined as all materials listed under the. category of petroleum products in the lock statistics. **Liquefied gases shown are the amounts transported on the entire navigable length of *the lliinois River. . .. ' '.*. 1977 23.452 4.093 0.0 1978 19. 521 . 3.658 0.0 Average . 26.0 5.0 . 3000.0 Table 2.2.2.3:4 Casualty and Spill Statistics -Fiscal Years 1969 thru 1972 (Ref. 10) CASUALTY/SPILL ILLINOIS RIVERS WESTERN *RIVERS Casualties** -all type barges Casualties of hazardous material barges***. Spills from hazardous * .mat.erial barges Casual ti es* of Liquefied gas barges Spills from double-skinned vessels ... Total length of waterway (miles) 178 40 1 ._.;._ 333 *Lower Mississippi, Upper Mi-ssiss.ippi, Ohio, and Illinois Rivers; casualties from these rivers* 97% of the casualties on western rfvers. **Casualtie.s whfch result in any of the following: loss of life,. damage to cargo-irr excess of $1;500, or release of cargo. 2831 508 69 9 7 3137 ***Hazardous material barges are generic type 17, 18, and 29 vessels. See Reference 10 for description. .,

• • • • • • -* -* **-****: ... TABLE '2.2.2.3:5 DATA ON AIRPORTS WITHIN 10 MILES OF DRESDEN STATION.(REFS .. 23, 27) APPROX. DIST. DIRECTION NO. LENGTH OF TYPE FROM STATION FROM STAT ION OPERATIONS. RUNWAY '> FROMM PVT. 4.5miles E 50* 2,773 ft. MORRIS PVT. 8 WNW 2,400 ft. ?,987 ft. ROSSI PVT. 9 miles N 50** 2,400 ft. BUSHBY PVT. 9.9 miles NNE 45** 1,800 ft. . JOLIET . Pub. 10 miles NNE 10,000* 3,452 ft. ., ._ . ...:.. ADELMANN*** PVT. 1 mile NE *Total peak month from FAA supplied documents. **Number per month as supplied by owner of airport ***Recent1y approved airstrip ft. 20*'11' 1,600 ft. WIDTH OF TYPE ORIENTATION RUNWAY OF RUNWAY OF RUNWAY 100 ft. TURF. NNE-SSW 135 ft. TURF. E-W 60 ft. ASPH. N-S 70 ft. TURF. E-W 100 ft. TURF . N-S 125 ft.* TURF. NE-SW 100 ft. ASPH. NW-SE 70 ft. TURF. SE-NW

' .. ' . ' . ;-. ..

• e e -' . ; *. Table 2.2.2.3:6 Data for Aircraft Crash and Probability Analysis R 106 N NARDx1C'i7 OPERATING r 0 D(r,O) .x (OPERATIONS/ A . AIRPORT MODE (MILES) (DEG) (/MILES2) (/OPERATION) YEAR) (MILES2) (/YEAR) FROMM Landing 4.5 90 0.0014 150 0.0056167 0.02833 4.5 90 0.0014 2.4 150 q.0056167 0.02833 Take-Off 4.5 90 0.00167 0.9 150 0.0056167 0.01267 4.5 90 0.00167 9 150 0.0056167 0.01267 MORRIS Lanqing 8.0 25 . 0. 000883 :. . 2.4 1456 0.0056167 0.17333 8.0. 155 0.000043 2A 1456 0.0056167 0.00833 8.0 65 0.00035 2.4 4370 0.0056167 0.206 8.0 115 o. 00011 2.4 4370 0.0056167 0.06467 Take-Off 8.0 25 . 0.000369 0.9 1456 . 0.0056167 0.027 8.o 0.000073 .. ' 155 0.9 1456 0.0056167 0.00533 8.0 65 *._ o. 00022 0. 9* 4370 0.0056167' 0.04867 8.0 115 0.00012 0.9 4370 0.0056167 0.02633 JOLIET Landing 10.0 . 10 0.00045 , *2A .* 6000. 0.0056167 0.364 10.0 170 0. 000011 . *. 2. 4 9000 . 0. 0056167 0.01333 10.0 80 0.000088 2.4 22500 0.005p167 0.26667 10.0 100 o. 000056 . 2.4 22500 0.0056167 0.17 Take-Off 10.0 : 10 0.00013 . 0.9 *. 7500 0.0056167 0.04933 10.0 170 0.000018 b. 9' 7500 .. 0.0056167 0.00667 10.0 80 0.000055 22500 0.0056167 0.06267 10.0. 100 0.000043 0.9 22500 0.0056167 0.049 Landing 1.0 115 0.01433 2.4 60 0.0056167 0.116 0.9 80 0.0374 . 2. 4 60 0.0056167 0.30233 Take-Off LO 115 0.0317 0.9 60 0.0056167 0.0906 0.9 80 0. 05734 .. 0 *. 9 60 0.0056167 0.174

,*.JO. --Table 2.2.2.3:7 with 5. Miles of the Site PIPE SIZE OPERATING CLOSEST DISTANCE PIPELINE COMPANY (in) MATERIAL CARRIED PRESSURE (PSI) TO THE PLANT (MILtS) Natural Gas 36 Natura 1 Gas 858 \ 1. 75 Pipeline Co. 36 **Natural Gas 858 1. 70 30 Natura 1 Gas 858 1.60 36 Natural Gas 650 1. 25 30 Natura 1 Gas 858 1. 70. 30 Natural Gas 858 1.60 Hydrocarbon 10 Propane, Natural. Gas 2100 4.0 Transportation, Inc. 10 Propane, Natural Gas 2100 4.0 6 Propane, Butane 500 2.0 Northern lllinois Gas 36 Natural Gas 740 2.5 10 Out of Operation 4 Natural Gas Unknown 3.0 Amoco 10 Crude Oi 1 3.0 12 Crude Oi 1 3.0 22 Crude Oi 1 3.0 *.'l-_

INDUSTRIAL SITES IN VICINlfY 1 MlOWEST FUELS REPROCESSING PLANT {GE). 2 PETROCHEMICAL CO. 3 /\LUMAX 4 REICllHOLO CHEMICAL CO 5. A. P. GREEN* 6 GENERAL ELECTRIC CO.TRAINING SCHOOL 1 MOBIL CHEMICAL 8

• MOBIL OIL 9 DURKEE SCH .** t I

• I ... PeN .1 e *,,' FIGURE 2.2.2.3:1 DRESOEN .. NUCLEAR POWER STATION AREA MAP . .. .-,. ... , . ' . ' > . .. * . . . =** *1 ,. : :saL\ET t\MMUN ITION

.1.* *. -. *' LEGEND 36" 36" .. JO" --36"' ' ...... 3011 **

• JO" '.**. . * .... . *' . *

• i*' "*. . *'* .**.* , -*. ',' ., *SITB :"" .. "'.**0*':** ... *' .. . * .. **: :. . . .. ** .... -, *. : .. *_ ',, .. : .... * '*. , '. : . :-. *' ........ radius FIGURE 2.2.2.3:2 PIPELINES CONSIDERED IN :rnE EVALIJAnoN OF HAZARO FROM EXPLOSION' . '" . . . / . . '. ' .... . . '.. . . ' . ' . ;. .. . ...